Text

Draft License Amendment Request Regarding Alternative Source Team
	ML16260A097
Person / Time
Site:	Farley
Issue date:	10/15/2016
From:	Pierce C R; Southern Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML16260A083	List: ML16260A097; ML16260A101; ML16260A109; ML16260A116; NL-16-0388, Draft Fuel Handling Accident Dose Consequences Using AST Methods; NL-16-0388, Draft LOCA Dose Consequences Using AST Methods; NL-16-0388, Draft License Amendment Request Regarding Alternative Source Team; NL-16-0388, September 22, 2016, Public Meeting Presentation Slides;
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	NL-16-0388
	Download: ML16260A097 (181)
	v • d • e

Charles R. Pierce Southern Nuclear Regulatory Affairs Director Operating Company, Inc.

40 Inverness Center Parkway Post Office Box 1295 Birmingham, AL 35242 Tel 205.992.7872 Fax 205.992.7601 October 15, 2016 Docket Nos.:

50-348 NL-16-0388 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555

-0001 Joseph M. Farley Nuclear Plant - Units 1 and 2 Alternative Source Term License Amendment Request

Ladies and Gentlemen:

Pursuant to 10 CFR 50.90, Southern Nuclear Operating Company (SNC) hereby requests an amendment to the Technical Specifications (TS) for Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2.

SNC requests Nuclear Regulatory Commission (NRC) review and approval of proposed revisions to the licensing basis of FNP that supports a full scope application of an Alternative Source Term (AST) methodology. Proposed TS changes, which are supported by the AST Design Basis Accident (DBA) radiological consequence analyses, are included in this license amendment request (LAR). In addition, the proposed amendment incorporates Technical Specification Task Force (TSTF) Traveler, TSTF-448, "Control Room Habitability," Revision 3. Enclosure 1 to this letter contains SNC's evaluation of the proposed changes. Enclosures 2 and 3 provide the markup changes to the Operating License, TS, and the TS Bases (for information). Enclosure 4 provides the clean, retyped Operating License and TS pages. Enclosures 5 through 12 provide additional information in support of this LAR.

SNC requests approval of the proposed LAR by October 31, 201

7. The proposed changes would be implemented within 120 days of issuance of the amendment.

In accordance with 10 CFR 50.91, a copy of this LAR with enclosures is being provided to the designated Alabama state officials.

This letter contains no NRC commitments.

If you have any questions, please contact Ken McElroy a t 205.992.7369.

U.S. Nuclear Regulatory Commission NL-16-0388 Page 2 Mr. C. R. Pierce states he is the Regulatory Affairs Director for Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and, to the best of his knowledge and belief, the facts set forth in this letter are true. Respectfully submitted, C. R. Pierce Regulatory Affairs Director Sworn to and subscribed before me this _____ day of ________________, 2016.

_____________________

Notary Public

My commission expires: __________

_________ CRP/wrv

Enclosures:

1. Basis for Proposed Change

2. Operating License and Technical Specification Pages (Markup)

3. Bases Pages (Markup) (For information only)

4. Operating License and Technical Specification Pages (Retyped)

5. Regulatory Guide 1.183 Conformance Tables

6. Loss of Coolant Accident Analysis

7. Fuel Handling Accident Analysis

8. Main Steam Line Break Analysis

9. Steam Generator Tube Rupture Analysis

10. Control Rod Ejection Analysis

11. Locked Rotor Analysis

12. FNP AST Accident Analysis Input Values Comparison Tables cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Ms. C. A. Gayheart, Vice President - Farley Mr. M. D. Meier, Vice President - Regulatory Affairs Mr. D. R. Madison, Vice President - Fleet Operations Mr. B. J. Adams, Vice President - Engineering Ms. B. L. Taylor, Regulatory Affairs Manager - Farley RTYPE: CFA04.054

U.S. Nuclear Regulatory Commission NL-16-0388 Page 3 U. S. Nuclear Regulatory Commission Ms. C. Haney, Regional Administrator Mr. S. A. Williams, NRR Project Manager - Farley Mr. P. K. Niebaum, Senior Resident Inspector - Farley Alabama Department of Public Health Dr. T. M. Miller, MD , State Health Officer

Joseph M. Farley Nuclear Plant - Units 1 and 2 Alternative Source Term License Amendment Request Enclosure 1 Basis for Proposed Change

Enclosure 1 Basis for Proposed Change Table of Contents 1.0 Summary Description 2.0 Detailed Description 2.1 Background 2.2 TSTF-448 3.0 Technical Evaluation

3.1 Meteorology and Atmosp heric Dispersion 3.2 Analytical Models 3.3 Loss of Coolant Accident 3.4 Fuel Handling Accident 3.5 Main Steam Line Break

3.6 Steam Generator Tube Rupture 3.7 Control Rod Ejection

3.8 Locked Rotor

3.9 Conclusions 3.10 TS Discussion

4.0 Regulatory Safety Analysis

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent

4.3 Significant Hazards Consideration

4.4 Conclusions 5.0 Environmental Consideration

6.0 References

to NL-16-0388 Basis for Proposed Change E1 - 1 1.0 Summary Description This evaluation supports a request to revise Operating License (OL) NPF

-2 and N FP-8 for Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2, respectively.

Southern Nuclear Operating Company (SNC) requests Nuclear Regulatory Commission (NRC) review and approval of proposed revisions to the licensing basis of FNP that supports a selected scope application of an Alternative Source Term (AST) methodology.

The proposed amendment also incorporates Revision 3 of TSTF

-448, "Control Room Habitability" into the FNP Technical Specifications (TS). 2.0 Detailed Description 2.1 Background In December 1999, the NRC issued a new regulation, 10 CFR 50.67, "Accident Source Term," which provided a mechanism for licensed power reactors to voluntarily replace the traditional accident source term used in their Design Basis Accident (DBA) analyses with an AST. Regulatory guidance for the implementation of the AST is provided in Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" (Reference 1). 10 CFR 50.67 requires a licensee seeking to use an AST to submit a license amendment request (LAR) and requires that the application contain an evaluation of the consequences of DBAs.

This LAR addresses the applicable requirements and guidance in proposing to use an AST in evaluating the offsite and Control R oo m (CR) radiological consequences of the FNP design basis accidents. This reanalysis involves several changes in selected analysis assumptions including different atmospheric dispersion values for the CR outside air intake. As part of the implementation of the AST, the Total Effective Dose Equivalent (TEDE) acceptance criterion of 10 CFR 50.67(b) replaces the previous whole body and

thyroid dose guidelines of 10 CFR 100.11. This will also replace the whole body (and its equivalent to any part of the body) dose criteria of 10 CFR 50, Appendix A, GDC 19.

2.2 TSTF-448 The proposed amendment would modify TS requirements related to CR envelope habitability in TS 3.7.

10 , Control Room" and TS Section 5.5, Administrative Controls

-Programs and Manuals.

The changes are consistent with NRC approved Industry/Technical Specification Task Force (TSTF) TS change TSTF

-448 Revision 3. The availability of this TS improvement was published in the Federal Register on January 17, 2007 as part of the consolidated line item improvement process (CLIIP) (Reference 2).

Adoption of TSTF

-448 supersedes in its entirety the current licensing basis for the Control Room Integrity Program (CRIP), as established for FNP Units 1 and 2 by License Amendments 166 and 158 respectively (Reference 3). These to NL-16-0388 Basis for Proposed Change E1 - 2 amendments for issued on September 30, 2004, well before NRC acceptance of TSTF-448 Revision 3 as a basis for CR habitability.

3.0 TECHNICAL EVALUATION

3.1 Meteorology and Atmospheric Dispersion The AST application uses atmospheric dispersion (X/Q) values for the Exclusion Area Boundary (EAB), the Low Population Zone (LPZ), and the CR receptors. As described below, the EAB and LPZ X/Q values are consistent with current licensing basis, as described in FNP Final Safety Analysis Report (FSAR) Table 2.3-12. In addition, for the CR, new and revised X/Q values have been developed. The values resulting at the CR intake s are calculated using the NRC-sponsored computer code ARCON96 consistent with the procedures in RG 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," (Reference 4). 3.1.1 Meteorological Data For the AST analyses, new and revised X/Q values were determined for the CR. A continuous temporally representative four year period of hourly average data from the FNP meteorological tower (i.e., January 1, 2000, through December 31, 2003) is used in these analyses. The data for these four years was selected as previous reviews have indicated that the data is of good quality.

Furthermore, this set of data has been previously reviewed by the NRC as part of a previous LAR (Reference 5). In that review an additional six months of derived data from 1999 was also developed , and has been used, as needed, in the atmospheric dispersion factors developed for this LAR.

As shown on Tables 3.4 and 3.5, the four year data set generally provided more conservative X/Q results. 3.1.2 EAB and LPZ Atmospheric Dispersion Factors

RG 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Plants," Section 5.3, "Meteorology Assumptions," states: Atmospheric dispersion values (X/Q) for the EAB, the LPZ, and the CR that were approved by the staff during the initial facility licensing or in subsequent licensing proceedings may be used in performing the radiological analyses identified by this guide. For the AST analyses, X/Q values for the EAB and the LPZ are consistent with the current licensing basis. Usage of the current licensing basis X/Q values for the EAB and the LPZ were approved by the NRC in Reference

5. The X/Q values for the EAB and the LPZ used in the radiological consequence analyses are shown in Table 3.1. to NL-16-0388 Basis for Proposed Change E1 - 3 Table 3.1 - EAB and LPZ X/Q values (sec/m

3) Location Time Period X/Q Value EAB 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 7.6x10-4 LPZ 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2.8x10-4 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.1x10-4 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.0x10-5 24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 5.4x10-6 96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 2.9x10-6 3.1.3 Control Room Atmospheric Dispersion Factors X/Q factors for onsite release

-receptor combinations were developed using the ARCON96 computer code. A number of various release

-receptor combinations were considered for the onsite CR atmospheric dispersion factors. These different cases are considered to determine the limiting release

-receptor combination for the events.

The X/Q factors from the existing calculations of record were used for the Containment Hatch, Reactor, and Plant Vent release points, based on the 4

-year and 4 1/2 year meteorological data described previously. New X/Q values were developed for the Reactor Water Storage Tank (RWST) release point for the LOCA.

Figure 3.1 provides a sketch of the general layout of FNP that has been annotated to highlight the onsite release and receptor point locations for the Loss of Coolant Accident (LOCA) (the LOCA provided the most limiting receptor-release locations and so were used by the other DBAs for conservatism). All releases are taken as ground level releases per the guidance of Position 3.2.1 of Reg. Guide 1.194.

Table s 3.2 and 3.3 provide 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 new RWST release point. Angles are calculated based on trigonometric layout of release and receptor points in relation to the North

-South and East

-West axes.

Table 3.4a provides the ARCON96 modeling outputs for releases originating at the reactor buildings, reactor vents, hatch doors, and RWST. These CR X/Q factors for LOCA, Main Steam Line Break (MSLB), Steam Generator Tube Rupture (SGTR), Control Room Envelope (CRE), and Locked Rotor Accidents are based on 4 years of meteorological data from years 2000 to 2003. Table 3.4b provides the X/Q values based on 4 1/2 years of data from 1999 to 2003 which was used in the Fuel Handling Accident (FHA). to NL-16-0388 Basis for Proposed Change E1 - 4 Figure 3.1 - Air Intake Locations and Release Points H

Location Description 1 Plant/Reactor Centerline Intersection (Coordinate Origin) 2 Unit 1 RWST Release 3 Unit 2 RWST Release 4 Unit 1 Reactor Release 5 Unit 2 Reactor Release 6 Unit 1 Vent Release 7 Unit 2 Vent Release 8 Unit 1 CR Emergency Intake 9 Unit 2 CR Emergency Intake 10 Normal CR Intake 11 Unit 2 Hatch Door 12 Unit 1 Hatch Door 1111 1112 to NL-16-0388 Basis for Proposed Change E1 - 5 Table 3.2 - Distance and Geometry of Release and Receptor Locations Receptor Release Point Horizontal Distance (ft) Vertical Distance (ft) Horizontal Distance (m) Vertical Distance (m) Direction to Source Unit 1 CR Emergency Air Intake Unit 1 RWST 3 72 3.5 1 13 1.07 101 Unit 1 CR Emergency Air Intake Unit 2 RWST 3 73 3.5 1 14 1.07 78 Unit 2 CR Emergency Air Intake Unit 1 RWST 373.27 3.5 1 14 1.07 102 Unit 2 CR Emergency Air Intake Unit 2 RWST 372.38 3.5 113.5 1.07 79 CR Normal Air Intake Unit 1 Vent Reactor* 154.04 111.46 121.5 0.0 46.9 33.9 37.0 0.0 135 1 59 CR Normal Air Intake Unit 2 Vent Reactor* 121.73 61.61 121.5 0.0 37.1 18.7 37.0 0.0 64 30

Reactor release heights are conservatively made the same as the normal and emergency CR intakes.

Table 3.3 - Elevations of A ir Intakes and Release Points Location Elevation (ft) Elevation (m) Ht above grade (ft) Ht. above grade (m) CR Emergency Intake 192 58.5 37.5 11.4 CR Normal Intake 178.5 54.4 24.0 7.3 Vent Stack Release 300 91.4 145.5 44.3 RWST Release 195.5 59.6 41 12.5 Table 3.4a - X/Q Values at the Control Room Air Intakes (4 years meteorological data)

Release Point Receptor 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1- 4 days 4 - 30 days U1 Vent U1 X CR 1.61E-03 1.26E-03 5.65E-04 3.43E-04 2.32E-04 U1 Reactor U1 X CR 1.51E-03 9.01E-04 3.95E-04 2.75E-04 1.91E-04 U1 Hatch Door U1 X CR 8.39E-04 5.11E-04 2.09E-04 1.47E-04 9.35E-05 U1 RWST U1 X CR 4.97E-04 3.69E-04 1.53E-04 1.15E-04 7.98E-05 U2 Vent U1 X CR 1.64E-03 1.37E-03 7.17E-04 5.41E-04 3.60E-04 U2 Reactor U1 X CR 1.66E-03 1.36E-03 6.81E-04 5.60E-04 4.21E-04 U2 Hatch Door U1 X CR 7.95E-04 6.73E-04 3.35E-04 2.48E-04 1.87E-04 U2 RWST U1 X CR 4.80E-04 3.82E-04 1.70E-04 1.28E-04 9.98E-05 to NL-16-0388 Basis for Proposed Change E1 - 6 Release Point Receptor 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1- 4 days 4 - 30 days U1 Vent U2 X CR 1.59E-03 1.25E-03 5.54E-04 3.34E-04 2.27E-04 U1 Reactor U2 X CR 1.51E-03 8.91E-04 3.91E-04 2.71E-04 1.87E-04 U1 Hatch Door U2 X CR 8.04E-04 4.90E-04 2.01E-04 1.39E-04 9.04E-05 U1 RWST U2 X CR 4.95E-04 3.66E-04 1.52E-04 1.14E-04 7.90E-05 U2 Vent U2 X CR 1.65E-03 1.38E-03 7.20E-04 5.47E-04 3.63E-04 U2 Reactor U2 X CR 1.65E-03 1.34E-03 6.75E-04 5.40E-04 4.03E-04 U2 Hatch Door U2 X CR 8.33E-04 6.98E-04 3.43E-04 2.57E-04 1.91E-04 U2 RWST U2 X CR 4.82E-04 3.82E-04 1.70E-04 1.28E-04 1.00E-04 U1 Vent Normal CR 2.01E-03 1.46E-03 6.07E-04 3.77E-04 2.59E-04 U1 Reactor Normal CR 1.56E-03 8.89E-04 3.62E-04 2.72E-04 1.85E-04 U2 Vent Normal CR 2.79E-03 2.36E-03 1.23E-03 9.18E-04 6.22E-04 U2 Reactor Normal CR 3.88E-03 3.11E-03 1.38E-03 1.29E-03 1.04E-03 Table 3.4b - X/Q Values at the Control Room Air Intakes (4 1/2 years meteorological data)

Release Point Receptor 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1- 4 days 4 - 30 days U1 Vent U1 X CR 1.62E-03 1.21E-03 5.37E-04 3.35E-04 2.32E-04 U1 Reactor U1 X CR 1.54 E-03 9.62 E-04 4.45 E-04 2.61 E-04 3.09 E-04 U1 Hatch Door U1 X CR 8.79 E-04 5.89 E-04 2.63 E-04 1.57 E-04 1.93 E-0 4 U2 Vent U1 X CR 1.59 E-03 1.35 E-03 7.08 E-04 5.16 E-04 3.59 E-04 U2 Reactor U1 X CR 1.64 E-03 1.34 E-03 6.65 E-04 5.45 E-04 4.18 E-04 U2 Hatch Door U1 X CR 7.83 E-04 6.52 E-04 3.22 E-04 2.43 E-04 1.85 E-04 U1 Vent U2 X CR 1.60E-03 1.19E-03 5.23E-04 3.24E-04 2.28E-04 U1 Reactor U2 X CR 1.54 E-03 9.50 E-04 4.43 E-04 2.70 E-04 3.11 E-04 U1 Hatch Door U2 X CR 8.56 E-04 5.67 E-04 2.50 E-04 1.52 E-04 1.92 E-0 4 U2 Vent U2 X CR 1.60 E-03 1.37 E-03 7.10 E-04 5.21 E-04 3.60 E-04 U2 Reactor U2 X CR 1.64 E-03 1.32 E-03 6.60 E-04 5.27 E-04 4.01 E-04 U2 Hatch Door U2 X CR 8.18 E-04 6.77 E-04 3.32 E-04 2.49 E-04 1.89 E-04 Notes: U1 - Unit 1 U2 - Unit 2 X CR - Emergency CR Intake Normal CR - Non-emergency CR Intake 3.2 Analytical Models The following computer codes are used in performing the FNP radiological dose analyses:

RADTRAD is used to determine the CR and offsite doses for the LOCA and FHA using the source term and X/Q inputs. The code considers the release timing, filtration, hold

-up, and chemical form of the nuclides released into the environment.

LocaDose is used to determine the CR and offsite doses for the MSLB, SGTR, CRE, and Locked Rotor Accidents using source term and X/Q inputs. This proprietary Bechtel software calculates radioactive isotope activities within to NL-16-0388 Basis for Proposed Change E1 - 7 regions, radioactive releases from regions, doses and dose rates within regions for humans and equipment, and inhalation and immersion doses to plant personnel.

ARCON96 (NUREG/CR

-6331) was used to determine the X/Qs at the CR intakes for selected release locations from plant meteorological data.

ORIGEN2 was used to calculate plant

-specific fission product inventories for use in the LOCA dose calculation.

3.3 Loss of Coolant Accident The LOCA is a postulated rupture in the reactor coolant system that results in expulsion of the coolant to containment. Even though the Emergency Core Cooling System (ECCS) is designed to maintain cooling of the fuel assemblies in this event, the dose consequence analysis is performed assuming a significant release of the radionuclides from the fuel assemblies.

3.3.1 Methodology Overview The LOCA is modeled as a release of nuclides from the reactor core into the containment building.

The Containment release paths modeled are: 1) the Containment Mini

-Purge System, 2) Containment leakage, 3) Engineered Safety Features (ESF) leakage, and 4) RWST backleakage.

The radiological source term characteristics and release timing are based on the AST methodology in RG 1.183.

Atmospheric dispersion factors from Section 3.1, above, are used in this analysis.

Doses to the public at the EAB and the LPZ, and occupants in the CR are determined.

3.3.2 Radiological Dose Models The RADTRAD (Versions 3.03 and 3.10) code was used to calculate the immersion and inhalation dose contributions to both the onsite and offsite radiological dose consequences. Models were developed for both the containment leakage and ECCS leakage cases.

The analysis used assumptions and inputs that follow the guidance in RG 1.183. The key parameters and assumptions are listed in Table 3.5 a. The calculated dose results are given in Table 3.5b. The calculated doses are within the RG 1.183 radiological dose acceptance criteria for a LOCA. These TEDE criteria are 25 rem at the EAB and LPZ for the duration of the accident, and 5 rem in the CR for the duration of the accident. to NL-16-0388 Basis for Proposed Change E1 - 8 Table 3.5a - Parameters and Assumptions for the LOCA Parameter Value ESF Leakage Initiation Time 20 minutes ESF Leakage Iodine Flashing Factor

10% Iodine Species ECCS Leakage Released to the Atmosphere Elemental 100% Organic 0% ECCS Leakage Rate to the RWST 1 gal/min RWST Leakage Iodine Flashing Factors: See Table 43 to Enclosure 6 RWST Capacity 505,562 gallons RWST Volume at Transfer to Recirculation 29,002 gallons Atmospheric Dispersion Factors (sec/m 3) Containment: Time (hr) EAB LPZ CR 0 - 2 7.6E-4 2.80E-4 1.66E-03 2 - 8 - 1.10E-4 1.36E-03 8 - 24 - 1.00E-5 6.81E-04 24 - 96 - 5.40E-6 5.60E-04 96 - 720 - 2.90E-6 4.21E-04 Plant Vent:

Time (hr) EAB LPZ CR 0 - 2 7.6E-4 2.80E-4 1.65E-03 2 - 8 - 1.10E-4 1.38E-03 8 - 24 - 1.00E-5 7.20E-04 24 - 96 - 5.40E-6 5.47E-04 96 - 720 - 2.90E-6 3.63E-04 RWST: Time (hr) EAB LPZ CR 0 - 2 7.6E-4 2.80E-4 4.97E-04 2 - 8 - 1.10E-4 3.82E-04 8 - 24 - 1.00E-5 1.70E-04 24 - 96 - 5.40E-6 1.28E-04 96 - 720 - 2.90E-6 1.00E-04

to NL-16-0388 Basis for Proposed Change E1 - 9 CR Parameters Parameter Value CR Volume 114,000 ft3 CR Ventilation System Normal Flow Rate 2340 cfm < 60 seconds CR Ventilation System Makeup Rate 375 cfm > 60 seconds CR Ventilation System Recirculation Flow Rate 2700 cfm > 60 seconds CR Ventilation System Charcoal Filter Efficiencies Pressurization Filters 98.5% all iodines Recirculation Filters 94.5% elemental and organic 98.5% particulate CR Unfiltered Inleakage 325 cfm CR Breathing Rate 3.5E-4 m 3/sec Occupancy Factors 0-24 hours 1.0 1 - 4 days 0.6 4 -30 days 0.4 Table 3.5b - Calculated LOCA Radiological Consequences TEDE (rem)

EAB LPZ CR Calculated results 13.24 6.01 4.68 Dose acceptance criteria 25 25 5 3.4 Fuel Handling Accident

Two cases are analyzed for the FHA: an accident in Containment and an accident in the spent fuel pool are of the Auxiliary Building.

For the FHA in Containment, the accident occurs in the Refueling Cavity. All gap activity of the damaged fuel rods is assumed to be released instantly into the overlying water. That activity that escapes the overlying water is then assumed to be uniformly distributed throughout the free volume of the Containment above the operating deck (EL 155'

-0"). There are two unfiltered release paths: through the open Equipment Hatch directly and one through the open Personnel Airlock to the Auxiliary Building and the Vent Stack.

The release through the open Equipment Hatch is directly to the environment. The release through the open Personnel Airlock credits neither filtration by the Auxiliary Building Ventilation System nor the holdup and dilution in the Exhaust Plenum or Vent Stack. This release path is effectively a direct release to the environment. It also establishes a pathway for the 10 CFM unfiltered ingress/egress CR inleakage.

Releases which pass through the personnel air lock can contaminate the Auxiliary Building at the level of the CR. Doses to operators from the to NL-16-0388 Basis for Proposed Change E1 - 10 ingress/egress of the CR through the doors into the contaminated area have been evaluated and are included in the dose results.

The FHA in the spent fuel pool area has also been evaluated. The results from this accident are bounded by the accident in Containment.

The analysis used assumptions and inputs that follow the guidance in RG 1.183. The key parameters and assumptions are listed in Table 3.6a. The calculated d ose results are given in Table 3.6b. The calculated doses are within the RG 1.183 radiological dose acceptance criteria for an FHA. These TEDE criteria are 6.3 rem at the EAB for the worst two hours, 6.3 rem at the LPZ for the duration of the accident and 5 rem in the CR for the duration of the accident.

Table 3.6a - Parameters and Assumptions for the FHA Parameter Value Reactor power 2,830.5 MWt Fraction of Fission Product Inventory in Gap I-131 0.08 Kr -85 0.10 Other Noble Gases 0.05 Other Halogens 0.05 Alkali Metals 0.12 Number of Damaged Fuel Assembly 1 Irradiated Fuel Decay 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months Radial Peaking Factor 1.7 Iodine Chemical Form Release from Fuel to Water Elemental 99.85% Organic 0.15% Minimum Refueling Cavity and Pool Water Depths 23 feet Overall Effective Decontamination Factor (DF) for Iodine 200 Chemical Form of Iodine Released from Pool Water Elemental 57% Organic 43% DF of Noble Gas 1 Duration of Rel ease 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Activity Release Rate 55,000 cfm Atmospheric Dispersion Factors (sec/m 3) Time (hr) EAB LPZ CR Vent Hatch 0 - 2 7.6E-4 2.80E-4 1.62E-3 8.79E-4 2 - 8 - 1.10E-4 1.37E-3 6.77E-4

to NL-16-0388 Basis for Proposed Change E1 - 11 CR Parameters Parameter Value CR Volume 114,000 ft3 CR Ventilation System Makeup Rate 375 cfm CR Ventilation System Recirculation Flow Rate 2700 cfm CR Ventilation System Charcoal Filter Efficiencies Pressurization Filters 98.5% all iodines Recirculation Filters 94.5% elemental and organic 98.5% particulate CR Unfiltered Inleakage 600 cfm (Isolation Mode) 325 cfm (Pressurization Mode)

CR Breathing Rate 3.5E-4 m 3/sec Occupancy Factors 0-24 hours 1.0 Table 3.6b - Calculated FHA Radiological Consequences TEDE (rem)

EAB LPZ CR Calculated results 2.4 0.9 1.0 Dose acceptance criteria 6.3 6.3 5 3.5 Main Steam Line Break

This event consists of a break in one main steam line outside of containment in which the faulted steam generator (SG) completely depressurizes and instantly releases the initial contents of the SG secondary side to the environment. The plant cooldown continues by dumping steam with the intact SGs. In addition to the release of nuclides that are initially present in the SG secondary side, leakage of primary coolant into the SG secondary side occurs at a rate equal to 0.35 gpm to the faulted SG, and 0.65 gpm to the intact SGs.

This is conservative relative to the TS limit of 150 gallons per day per SG.

Two iodine spike cases are considered. In the pre

-accident iodine spike case, a reactor transient is assumed to occur prior to the event in which the primary coolant iodine concentration has increased to the maximum TS value of 30 ty concentration is at the equilibrium TS iodine spike is assumed to have a duration of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

In both cases, a 1% failed fuel term is assumed, consistent with the F NP current licensing basis.

Leakage from the Reactor Coolant System (RCS) into all of the SGs, and steam release from the intact SGs, continues until the RCS is cooled to 200 F after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The leakage to the faulted SG is modeled as a direct flow from the RCS to the environment without partitioning. In the leakage to the intact SGs, noble gases are assumed to leak directly to the environment.

A partition to NL-16-0388 Basis for Proposed Change E1 - 12 coefficient of 100 is applied to the iodine nuclides in the intact SGs. Flows out of the faulted SG are assumed to be released to the environment without partitioning.

The release locations from the faulted and intact SGs are conservatively taken as the most limiting release locations from the LOCA.

The CR is automatically realigned into the emergency ventilation mode upon receipt of a safety injection signal.

The analysis used assumptions and inputs that follow the guidance in RG 1.183. The key parameters and assumptions are listed in Table 3.7a. The calculated dose results are given in Table 3.7b. The calculated doses are within the RG 1.183 radiological dose acceptance criteria for a MSLB. These TEDE criteria are 25 rem at the EAB and LPZ for the fuel damage or pre

-incident spike case, and 2.5 rem at the EAB and LPZ for the concurrent iodine spike case. The TEDE criteria is 5 rem for the CR occupant in both cases, and the duration is until cold shutdown is established.

Table 3.7a - Parameters and Assumptions for the MSLB Parameter Value Steam Releases from Intact SG to Environment 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 316,715 Ibm 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 703,687 Ibm 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 948,000 Ibm Intact SG Liquid Iodine Partition Coefficient 100 Steam mass released from faulted SG to the Environment 439,145 Ibm Faulted SG Dryout Time 322.8 seconds Atmospheric Dispersion Factors (sec/m 3) Time (hr) EAB LPZ CR 0 - 2 7.6E-4 2.80E-4 1.66E-3 2 - 8 - 1.10E-4 1.38E-3 8 - 24 - 1.10E-5 7.20E-4 CR Parameters Parameter Value CR Volume 114,000 ft3 CR Ventilation System Makeup Rate 375 cfm CR Ventilation System Recirculation Flow Rate 2700 cfm CR Ventilation System Charcoal Filter Efficiencies Pressurization Filters 98.5% all iodines Recirculation Filters 94.5% elemental and organic 98.5% particulate CR Unfiltered Inleakage 3 10 cfm to NL-16-0388 Basis for Proposed Change E1 - 13 CR Breathing Rate 3.5E-4 m 3/sec CR Parameters (continued)

Occupancy Factors 0-24 hours 1.0 1 - 4 days 0.6 4 -30 days 0.4 Table 3.7b - Calculated MSLBA Radiological Consequences TEDE (rem)

EAB LPZ CR Calculated results Pre-Incident Spike 0.94 0.37 0.23 Concurrent Iodine Spike 0.95 0.45 0.45 Dose acceptance criteria Fuel Damage or Pre- Incident Spike 25 25 5 Concurrent Iodine

Spike 2.5 2.5 5 3.6 Steam Generator Tube Rupture

The SGTR event represents an instantaneous rupture of a SG tube that releases primary coolant into the lower pressure secondary system. In addition to the break flow rate, primary

-to-secondary leakage occurs at a rate equal to 0.35 gpm to the ruptured and 0.65 gpm for both of the intact SGs. All leakage flow into the ruptured SG is secured after 30 minutes. Leakage into the intact SGs continues until the RCS is cooled to cold shutdown conditions after 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . A portion of the break and leakage flow to the ruptured SG flashes to vapor based upon the thermodynamic conditions in the reactor and secondary coolant. The portion of the primary coolant that does flash in the SG secondary is released directly to the environment without mitigation. The break and leakage flow that does not flash mixes with the bulk water in the SG where the activity is released based upon the steaming rate and a partition coefficient.

A SG partition coefficient of 100 is applied to the iodine nuclides. Two iodine spike cases are considered. In the pre

-accident iodine spike case, a reactor transient is assumed to occur prior to the event in which the primary coolant iodine concentration has increased to the maximum TS value of 30 Ci/gm. For the concurrent spike case, the initial iodine activity concentration is at the equilibrium TS This concurrent to NL-16-0388 Basis for Proposed Change E1 - 14 iodine spike is assumed to have a duration of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

In both cases, a 1% failed fuel term is assumed, consistent with the F NP current licensing basis. The release locations from the faulted and intact SGs are conservatively taken as the most limiting release locations from the LOCA.

The CR is automatically realigned into the emergency ventilation mode upon receipt of a safety injection signal. The analysis used assumptions and inputs that follow the guidance in RG 1.183. The key parameters and assumptions are listed in Table 3.8a. The calculated dose results are given in Table 3.8b. The calculated doses are within the RG 1.183 radiological dose acceptance criteria for a SGTR. These TEDE criteria are 25 rem at the EAB and LPZ for the fuel damage or pre-incident spike case, and 2.5 rem at the EAB and LPZ for the concurrent iodine spike case. The TEDE criteria is 5 rem for the CR occupant in both cases, and the duration is until cold shutdown is established.

Table 3.8a - Parameters and Assumptions for the SGTR Accident Parameter Value Reactor Coolant Activity (Initial)

Pre-Accident Iodine Spike -131 Accident-Initiated Iodine Spike -131 Noble Gas 1% failed fuel Concurrent Iodine Spiking Factor 335 Duration of Intact SG Flow 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Secondary Coolant Iodine Specific Activity -131 Atmospheric Dispersion Factors (sec/m 3) Time (hr) EAB LPZ CR 0 - 2 7.6E-4 2.80E-4 1.66E-3 2 - 8 - 1.10E-4 1.38E-3 CR Parameters Parameter Value CR Volume 114,000 ft3 CR Ventilation System Makeup Rate 375 cfm CR Ventilation System Recirculation Flow Rate 2700 cfm CR Ventilation System Charcoal Filter Efficiencies Pressurization Filters 98.5% all iodines Recirculation Filters 94.5% elemental and organic 98.5% particulate CR Unfiltered Inleakage 310 cfm CR Breathing Rate 3.5E-4 m 3/sec to NL-16-0388 Basis for Proposed Change E1 - 15 CR Parameters (continued)

Occupancy Factors 0-24 hours 1.0 1 - 4 days 0.6 4 -30 days 0.4 Table 3.8b - Calculated SGTR Accident Radiological Consequences TEDE (rem)

EAB LPZ CR Calculated results Pre- Incident Spike 2.4 0.92 0.48 Concurrent Iodine Spike 0.82 0.34 0.17 Dose acceptance criteria Fuel Damage or Pre- Incident Spike 25 25 5 Concurrent Iodine Spike 2.5 2.5 5 3.7 Control Rod Ejection The Control Rod Ejection event involves a reactivity insertion that produces a short, rapid core power level increase which results in fuel rod damage and localized melting. Two separate release pathways are evaluated: a release from containment and a release from the secondary system. In both cases, 10% of the noble gases and 10% of the iodines in the core are available for release from the fuel gap of the damaged fuel rods.

In addition, 12% of the alkali metals are also assumed to be located in the fuel rod gap.

For releases from containment, 10% of the fuel rods in the core experience cladding failure and 0.25% of the fuel experiences melting. The activity in the fuel rod gap of the damaged fuel is instantaneously and uniformly mixed throughout the containment atmosphere. Moreover, 100% of the noble gases and 50% of the iodines in the melted fuel are also added to the fission product inventory in containment.

No credit is taken for removal by containment sprays or for deposition of elemental iodine on containment surfaces. Credit is taken for natural deposition of aerosols in containment. Activity is released from containment at the TS leak rate limit for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at half that rate after that. The release of iodine initially present in the SG secondary side is also included. For releases from the secondary system, 10% of the fuel rods in the core are breached and 0.25% of the fuel experiences melting. Activity released from the to NL-16-0388 Basis for Proposed Change E1 - 16 fuel is completely dissolved in the primary coolant and is available for release to the secondary system. In this case, 100% of the noble gases and 50% of the iodines in the melted fuel is also released into the reactor coolant. The noble gases are assumed to be released directly to the environment, and the remaining fission products are transported to the SGs at 1 gpm which is conservative relative to the TS limit of 150 gallons per day per SG

. The leakage duration is 2500 seconds. In keeping with previous evaluations of the CRE accident, the Secondary System mass releases to the environment last for 98

seconds. With the large amount of fission products introduced into the reactor coolant by failed fuel, the initial activity of the RCS prior to the event is not considered. However, the dose contribution from the iodine activity initially present in the SG secondary is included in the analysis. The release locations are conservatively taken as the most limiting release locations from the LOCA.

The CR ventilation system is automatically realigned into the emergency ventilation mode following receipt of a safety injection signal.

The analysis used assumptions and inputs that follow the guidance in RG 1.183. The key parameters and assumptions are listed in Table 3.9a. The calculated dose results are given in Table 3.9b. The calculated doses are within the RG 1.183 radiological dose acceptance criteria for a CRE. These TEDE criteria are 6.3 rem at the EAB and LPZ, and 5 rem for the CR occupant. The duration is 30 days for the Containment pathway, and until cold shutdown is established for the secondary pathway.

Table 3.9a - Parameters and Assumptions for the CRE Accident Parameter Value Reactor power 2831 MWt Post-CREA failed fuel 10% Percentage of Melted Fuel Release Containment Leakage Iodine 50% Noble Gases 100% Primary-to-Secondary Leakage Iodine 50% Noble Gases 100% Iodine Chemical Form Release to Containment Aerosol (cesium iodide) 95% Elemental 4.85% Organic 0.15% Containment Leak Rates 0-24 hours 0.15 weight %/day

> 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.075 weight %/day Primary-to-Secondary Leak Duration 2500 seconds RCS Leakage 1 gpm SG Liquid Iodine Partition Coefficient 100 Steam Releases from Intact SG to Environment 426,000 lbm to NL-16-0388 Basis for Proposed Change E1 - 17 Atmospheric Dispersion Factors (sec/m 3) Time (hr) EAB LPZ CR 0 - 2 7.6E-4 2.80E-4 1.66E-3 2 - 8 - 1.10E-4 1.38E-3 8 - 24 - 1.00E-5 7.20E-4 24 - 96 - 5.40E-6 5.60E-4 96 - 720 - 2.90E-6 4.21E-4 CR Parameters Parameter Value CR Volume 114,000 ft3 CR Ventilation System Makeup Rate 375 cfm CR Ventilation System Recirculation Flow Rate 2700 cfm CR Ventilation System Charcoal Filter Efficiencies Pressurization Filters 98.5% all iodines Recirculation Filters 94.5% elemental and organic 98.5% particulate CR Unfiltered Inleakage 310 cfm CR Breathing Rate 3.5E-4 m 3/sec Occupancy Factors 0-24 hours 1.0 1 - 4 days 0.6 4 -30 days 0.4 Table 3.9b - Calculated CRE Accident Radiological Consequences TEDE (rem)

EAB LPZ CR Calculated results 3.8 2.7 3.7 Dose acceptance criteria 6.3 6.3 5 3.8 Locked Rotor The Locked Rotor dose analysis is defined by the 20% of the fuel rods which become damaged by the event. A radial peaking factor of 1.7 is assumed. Radionuclides released from the fuel are instantaneously and uniformly mixed throughout the primary coolant. Noble gases are released directly to the environment, and the remaining isotopes are transported to the SGs at a rate of 1 gpm. This continues for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , by which time the RCS temperature is cooled to cold shutdown conditions. Since the quantity of the fission products released from the failed fuel dominates the RCS activity during the event, the initial nuclide concentration in the RCS prior to the event is not considered. However, the analysis does include the to NL-16-0388 Basis for Proposed Change E1 - 18 dose contribution from the release of iodine initially present in the SG secondary side. The release locations are conservatively taken as the most limiting release locations from the LOCA.

The analysis assumes that the CR isolates and enters the emergency ventilation mode at the onset of the accident.

The analysis used assumptions and inputs that follow the guidance in RG 1.183. The key parameters and assumptions are listed in Table 3.10a. The calculated dose results are given in Table 3.10b. The calculated doses are within the RG 1.183 radiological dose acceptance criteria for a Locked Rotor Accident. These TEDE criteria are 2.5 rem at the EAB and LPZ, and 5 rem for the CR occupant. The duration is 30 days for the Containment pathway, and until cold shutdown is established for the secondary pathway.

Table 3.10a - Parameters and Assumptions for the L ocked Rotor Accident Parameter Value Reactor power 2831 MWt Post-Locked Rotor Accident 2 0% Secondary Coolant Iodine Specific Activity Ci/gm DE 1

-131 Fraction of Fission Product Inventory in Gap 1-131 0.08 Kr-85 0.10 Other Halogens and Noble Gases 0.05 Alkali Metals 0.12 RCS Leakage 1 gpm SG Liquid Iodine Partition Coefficient 100 Iodine Release from SG Elemental 97% Organic 3% Steam Releases from SG to Environment 0-2 hours 512,325 Ibm 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 833,221 Ibm Atmospheric Dispersion Factors (sec/m 3) Time (hr) EAB LPZ CR 0 - 2 7.6E-4 2.80E-4 1.66E-3 2 - 8 - 1.10E-4 1.38E-3 CR Parameters Parameter Value CR Volume 114,000 ft3 CR Ventilation System Makeup Rate 375 cfm CR Ventilation System Recirculation Flow Rate 2700 cfm CR Parameters (continued) to NL-16-0388 Basis for Proposed Change E1 - 19 CR Ventilation System Charcoal Filter Efficiencies Pressurization Filters 98.5% all iodines Recirculation Filters 94.5% elemental and organic 98.5% particulate CR Unfiltered Inleakage 310 cfm CR Breathing Rate 3.5E-4 m 3/sec Occupancy Factors 0-24 hours 1.0 1 - 4 days 0.6 4 -30 days 0.4 Table 3.10b - Calculated Locked Rotor Accident Radiological Consequences TEDE (rem)

EAB LPZ CR Calculated results 1.2 0.83 <5* Dose acceptance criteria 2.5 2.5 5

The actual CR dose is not reported in the FNP FSAR for this event.

3.9 Conclusions

The proposed changes provide a source term for FNP that will result in a more accurate assessment of the DBA radiological doses. The results from all of the dose analyses show that the predicted dose consequence results are within the allowable regulatory limits. The revised radiological dose to the CR occupants allows for a revised unfiltered air in

-leakage assumption that provides a conservative margin over that determined by air in

-leakage testing.

3.10 TS Discussion 3.10.1 TSTF-448 With License Amendments 166/158 for FNP Units 1/2, the NRC approved a new section TS 5.5.18, "Control Room Integrity Program," to the Programs and Manuals Section of the TS. As stated in the Safety Evaluation for this LAR: The CRIP represents the manner in which the licensee will demonstrate CRE integrity. SNC has proposed a CRIP which incorporates the FNP design and licensing

-basis details. It has been formulated following numerous discussions with the NRC staff. The proposal reflects the evolution of the NRC staff's guidance on the CRIP from that which was presented in RG 1.196. This guidance is current as of the date this SE was issued. However, the issuance date of these License Amendments (September 30, 2004) predated NRC approval of TSTF

-448 Revision 3 (January 17, 2007). to NL-16-0388 Basis for Proposed Change E1 - 20 Accordingly, the proposed adoption of TSTF

-448 will supersede the current licensing basis for the CRIP. 3.1 0.1.1 Applicability of Published Safety Evaluation SNC has reviewed the safety evaluation dated January 17, 2007 as part of the Consolidated Line Item Improvement Process. This review included a review of the NRC staff's evaluation, as well as the supporting information provided to support TSTF

-448. SNC has concluded that the justifications presented in the TSTF proposal and the safety evaluation prepared by the NRC staff are applicable to FNP Units 1 and 2 and justify this amendment for the incorporation of the changes to the FNP TS.

3.1 0.1.2 Optional Changes and Variations The model Safety Evaluation and model application provided optional statements and evaluations to accommodate variations in plant design and licensing basis. For the purposes of the FNP Unit 1 and 2 TSs, the following optional statements and evaluations are applicable:

1. Throughout SE - The name of the CR ventilation system at FNP is "Control Room Emergency Filtration/

Pressurization System (CREFS)" rather than "CREEVS."

2. Throughout SE - FNP refers to CREFS as "trains" versus "subsystems."

3. Throughout SE - The facility name is "Farley Nuclear Plant, Unit s 1 and 2." 4. Throughout SE - The term "TS 3.7.10, 'CREEVS'" should be "TS 3.7.10, "CREFS'."

5. Throughout SE - SNC confirms that the CRE Habitability Program is contained in TS 5.5.18.

6. Section 1.0, last paragraph, and Section 3.1, first paragraph - SNC is not adding a new administrative controls program TS 5.5.18; it is revising an existing TS 5.5.18 program on that subject. 7. Section 2.2, second paragraph - With this AST LAR, FNP will be using 5 rem TEDE.

8. Section 2.3, first paragraph

- FNP was licensed under the 10 CFR 50 General Design Criteria.

to NL-16-0388 Basis for Proposed Change E1 - 21 9. Section 3.0 - At FNP, the emergency mode of CREFS involves pressurization.

10. Section 3.1, first paragraph

- There are no plant-specific exceptions being taken from TSTF

-448. 11. Section 3.1, second paragraph - The TS Bases Control Program is TS 5.5.14, not TS 5.5.11.

12. Section 3.2 - The editorial correction to replace "irradiate" with "irradiated" in TS 3.7.10 Condition E is not applicable.

13. Section 3.3 - Evaluation 1 is most applicable to FNP. TS 3.7.10 includes the LCO Note described in TSTF

-287, but Action B reads "CRE inoperable" versus "Two CREFS trains inoperable due to inoperable control room boundary in MODES 1, 2, 3, and

4. Currently, the FNP TSs treat Operability of CREFS separately from Operability of the CRE. With the adoption of TSTF

-448, Operability of CREFS requires Operability of the CRE. Accordingly, the first paragraph sentence "The licensee proposed to revise the action requirements of TS 3.7.10, "CREFS," to acknowledge that an inoperable CRE boundary, depending upon the location of the associated degradation, could cause just one, instead of both CREFS trains to be inoperable" should be revised to read "-could cause just one, or both CREFS trains to be inoperable." Similarly, the second paragraph sentence "This change clarifies how to apply the action requirements in the event just one CREFS train is unable to ensure CRE occupant safety-because of an inoperable CRE boundary" should read "This change clarifies how to apply the action requirements in the event that one or both CREFS trains are unable to ensure CRE occupant safety-because of an inoperable CRE boundary."

14. Section 3.3, Evaluation 1, first paragraph, Condition B bullet - Should read "One or more CREFS trains inoperable due to inoperable CRE boundary in MODE 1, 2, 3, or 4, during movement of irradiated fuel assemblies, and during CORE ALTERATIONS."

15. Section 3.3, Evaluation 1, third paragraph - Should read "The licensee proposes to replace existing Required Action B.2.1,

'Restore CRE to OPERABLE status

,' which has a 24

-hour Completion Time-with revised Required Action B.1, to immediately initiate action to implement mitigating actions; revised Required Action B.2

-" 16. Section 3.3, Evaluation 6 and paragraph after Evaluation 6 -

FNP does not have a CRE pressurization surveillance to NL-16-0388 Basis for Proposed Change E1 - 22 requirement as described in Evaluation 6. However, it does have two existing SRs relevant to the CRE:

SR 3.7.10.4

. SR 3.7.10.5 Verify CRE integrity in accordance with the CRIP.

The paragraph after Evaluation 6 should acknowledge that the new SR 3.7.10.4 on performing CRE unfiltered air inleakage testing is replacing these two surveillances.

17. Section 3.3, paragraph after Evaluation 6 - SNC proposes to follow the air inleakage testing methodology described in that paragraph.

18. Section 3.4, first paragraph

- With this AST LAR, FNP will be using 5 rem TEDE.

19. Section 3.4, fifth paragraph

- SNC is not proposing any exceptions to RG 1.197 Sections C.1 and C.2.

20. Section 3.4, sixth paragraph - The staggered test basis interval should be 24 months versus 18 months.

21. Section 4 - The State consultation should refer to the "Alabama" State official.

3.1 0.1.3 License Condition Regarding Initial Performance of New Surveillance and Assessment Requirements SNC proposes the following as a license condition to support implementation of the proposed TS changes:

Upon implementation of Amendment No. xxx adopting TS TF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.10.4, in accordance with TS 5.5.18.c.(i), the assessment of CRE habitability as required by Specification 5.5.18.c.(ii), and the measurement of CRE pressure as required by Specification 5.5.18.d, shall be considered met. Following implementation:

(a) The first performance of SR 3.7.10.4, in accordance with Specification 5.5.18.c.(i), shall be within the specified Frequency of 6 years, plus the 18

-month allowance of SR 3.0.2, as measured from February 1 through 8, 2016, the date of the most recent successful tracer gas test, as stated in the August 25, 2004 letter response to Generic Letter 2003

-01, or within the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 years.

to NL-16-0388 Basis for Proposed Change E1 - 23 (b) The first performance of the periodic assessment of CRE habitability, Specification 5.5.18.c.(ii), shall be within 3 years, plus the 9

-month allowance of SR 3.0.2, as measured from February 1 through 8, 2016, the date of the most recent successful tracer gas test, as stated in the August 25, 2004 letter response to Generic Letter 2003

-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

(c) The first performance of the periodic measurement of CRE pressure, Specification 5.5.18.d, shall be within 24 months, plus the 180 days allowed by SR 3.0.2, as measured from July 11, 2015, the date of the most recent successful pressure measurement test, or within 180 days if not performed previously.

4.0 Regulatory Safety Analysis 4.1 Applicable Regulatory Requirements/Criteria Title 10 Code of Federal Regulations Section 50.36. "Technical specifications" Changes to the FNP TSs are proposed for the adoption of TSTF

-448. A description of these proposed changes and their relationship to applicable regulatory requirements and guidance was provided in the NRC Notice of Availability published in Reference 2, and TSTF-448, Revision 3. Title 10 Code of Federal Regulations Section 50.67. " Accident Source Term" On December 23, 1999, the NRC published 10 CFR 50.67, "Accident Source

Term," in the Federal Register. This regulation provides a mechanism for licensed power reactors to replace the current accident source term used in the DBA analysis with an AST. The direction provided in 10 CFR 50.67 is that licensees who seek to revise their current accident source term in design basis radiological consequence analyses shall apply for a LAR under 10 CFR 50.90.

4.2 Precedent Although a number of alternative AST submittals have been reviewed and approved by the NRC since RG 1.183 was published, and have helped to inform the content of this application, no specific precedent submittals are referenced herein. 4.3 Significant Hazards Consideration Southern Nuclear Operating Company (SNC) evaluated whether or not a significant hazards consideration is involved with the proposed changes by focusing on three standards set forth in 10 CFR 50.92(c) as discussed below.

to NL-16-0388 Basis for Proposed Change E1 - 24 With regard to the proposal to adopt TSTF

-448, SNC has reviewed the proposed no significant hazards consideration determination (NSHCD) published in the Federal Register as part of the Consolidated Line Item Improvement Process. SNC has concluded that the proposed NSHCD presented in the Federal Register notice is applicable to Farley Nuclear Plant (FNP) and is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91(a).

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

There are no physical changes to the plant being introduced by the proposed changes to the accident source term. Implementation of Alternative Source Term (AST) and the new atmospheric dispersion factors have no impact on the probability for initiation of any Design Basis Accident s (DBAs). Once the occurrence of an accident has been postulated, the new accident source term and atmospheric dispersion factors are an input to analyses that evaluate the radiological consequences.

The proposed changes do not involve a revision to the design or manner in which the facility is operated that could increase the probability of an accident previously evaluated in Chapter 15 of the Final Safety Analysis Report (FSAR).

Based on the AST analyses, there are no proposed changes to performance requirements and no proposed revision to the parameters or conditions that could contribute to the consequences of an accident previously discussed in Chapter 1 5 of the FSAR. Plant

-specific radiological analyses have been performed using the AST methodology and new X/Qs have been established.

Based on the results of these analyses, it has been demonstrated that the control room (CR) and off-site dose consequences of the limiting events considered in the analyses meet the regulatory guidance provided for use with the AST, and the doses are within the limits established by 10 CFR 50.67.

Therefore, it is concluded that the proposed amendment does not involve a significant increase in the probability or the consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

No new mod es of operation are introduced by the proposed changes. The proposed changes will not create any failure mode not bounded by previously evaluated accidents. Implementation of AST and the associated proposed Technical Specification changes and new X/Qs have no impact to the initiation of any DBAs. These changes do not to NL-16-0388 Basis for Proposed Change E1 - 25 affect the design function or modes of operation of structures, systems and components in the facility prior to a postulated accident. Since structures, systems and components are operated no differently after the AST implementation, no new failure modes are created by this proposed change. The AST change itself does not have the capability to initiate accidents.

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

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The AST analyses have been performed using approved methodologies to ensure that analyzed events are bounding and safety margin has not been reduced. The dose consequences of these limiting events are within the acceptance criteria presented in 10 CFR 50.67. Thus, by meeting the applicable regulatory limits for AST, there is no significant reduction in a margin of safety.

Therefore, because the proposed changes continue to result in dose consequences within the applicable regulatory limits, the proposed amendment does not involve a significant reduction in margin of safety.

4.4 Conclusions

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 Environmental Consideration A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with th e proposed amendment.

to NL-16-0388 Basis for Proposed Change E1 - 26 Regarding the proposal to adopt TSTF

-448, SNC has reviewed the environmental evaluation included in the model safety evaluation dated January 17, 2007 as part of the CLIIP. SNC has concluded that the staff's findings presented in that evaluation are applicable to FNP and the evaluation is hereby incorporated by reference for this application.

6.0 References

1. Regulatory Guide 1.183, "Alternative Radiological Source Terms For Evaluating Design Basis Accidents At Nuclear Power Reactors," July 2000. 2. Notice of Availability of Technical Specification Improvement To Modify Requirements Regarding Control Room Envelope Habitability Using the Consolidated Line Item Improvement Process, 72 Federal Register 10 (January 17, 2007). 3. Letter from S Peters (NRC) to L. Stinson (SNC), dated September 30, 2004, "Joseph M. Farley Nuclear Plant, Units 1 and 2 Re: Issuance of Amendments (TAC Nos. MC4186 and MC4187)."

4. Regulatory Guide 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants

," June 2003.

5. Letter from S Peters (NRC) to L. Stinson (SNC), dated September 30, 2004, "Joseph M.

Farley Nuclear Plant, Units 1 and 2 Re: Issuance of Amendments (TAC Nos. MC 0625 and MC0626)."

Joseph M. Farley Nuclear Plant - Units 1 and 2 Alternative Source Term License Amendment Request Enclosure 2 Operating License and Technical Specification Pages (Markup)

Farley - Unit 1 Renewed License No. NPF-2 Amendment No. XXX (5) Updated Final Safety Analysis Report Supplement The Updated Final Safety Analysis Report supplement, as revised, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4) following issuance of this renewed license. Until that update is complete, Southern Nuclear may make changes to the programs and activities

described in the supplement without prior Commission approval, provided that Southern Nuclear evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements of that section.

The Southern Nuclear Updated Final Safety Analysis Report supplement, submitted pursuant to 10 CFR 54.21(d), describes certain future activities to be completed prior to the period of extended operation.

Southern Nuclear shall complete these activities no later than June 25, 2017, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

(6) Reactor Vessel Material Surveillance Capsules All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of American Society for Testing and Materials (ASTM) E 185

-82 to the extent practicable for the configuration of the specimens in the capsule.

Any changes to the capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion.

(7) Upon implementation of Amendment No. xxx adopting TSTF

-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.10.4, in accordance with TS 5.5.18.c.(i), the assessment of CRE habitability as required by Specification 5.5.18.c.(ii), and the measurement of CRE pressure as required by Specification 5.5.18.d, shall be considered met. Following implementation:

(a) The first performance of SR 3.7.10.4, in accordance with Specification 5.5.18.c.(i), shall be within the specified Frequency of 6 years, plus the 18

-month allowance of SR 3.0.2, as measured from February 1 through 8, 2016, the date of the most recent successful tracer gas test, as stated in the August 25, 2004 letter response to Generic Letter 2003-01, or within the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 years. (b) The first performance of the periodic assessment of CRE habitability, Specification 5.5.18.c.(ii), shall be within 3 years, plus the 9

-month allowance of SR 3.0.2, as measured from February 1 through 8, 2016, the date of the most recent successful tracer gas test, as stated in the August 25, 2004 letter response to Generic Letter 2003

-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

Farley - Unit 1 Renewed License No. NPF-2 Amendment No. XXX (c) The first performance of the periodic measurement of CRE pressure, Specification 5.5.18.d, shall be within 24 months, plus the 180 days allowed by SR 3.0.2, as measured from July 11, 2015, the date of the most recent successful pressure measurement test, or within 180 days if not performed previousl y. D. Southern Nuclear shall fully implement and maintain in effect all provisions of the Commission

-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

The plan, which contain Safeguards Information protected under 10 CFR 73.21, is entitled:

"Southern Nuclear Operating Company Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan," and was submitted on May 15, 2006. Southern Nuclear shall fully implement and maintain in effect all provisions of the Commission

-approved cyber security (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Southern Nuclear CSP was approved by License Amendment No. 186, as supplemented by a change approved by License Amendment No. 199. E. This renewed license is subject to the following additional conditions for the protection of the environment:

Farley - Unit 2 Renewed License No. NPF-8 Amendment No. XXX to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.

c. Transition License Conditions

1) Before achieving full compliance with 10 CFR 50.48(c), as specified by

2) below, risk

-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in 2) above.

2) The licensee shall implement the modifications to its facility, as described in Attachment S, Table S

-2, "Plant Modifications Committed," of SNC letter NL 1273, dated August 29, 2014, to complete the transition to full compliance with 10 CFR 50.48(c) by November 6, 2017. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

3) The licensee shall implement the items as listed in Attachment S, Table S

-3, "Implementation Items," of SNC letter NL 1273, dated August 29, 2014, within 180 days after NRC approval, except for items 30 and 32. Items 30 and 32 shall be implemented by February 6, 2018.

(7) Upon implementation of Amendment No. xxx adopting TSTF

-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.10.4, in accordance with TS 5.5.18.c.(i), the assessment of CRE habitability as required by Specification 5.5.18.c.(ii), and the measurement of CRE pressure as required by Specification 5.5.18.d, shall be considered met. Following implementation:

(a) The first performance of SR 3.7.10.4, in accordance with Specification 5.5.18.c.(i), shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from February 1 through 8, 2016, the date of the most recent successful tracer gas test, as stated in the August 25, 2004 letter response to Generic Letter 2003

-01, or within the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 years.

(b) The first performance of the periodic assessment of CRE habitability, Specification 5.5.18.c.(ii), shall be within 3 years, plus the 9

-month allowance of SR 3.0.2, as measured from February 1 through 8, 2016, the date of the most recent successful tracer gas test, as stated in the August 25, 2004 letter response to Generic Letter 2003

-01, or within the next 9 Farley - Unit 2 Renewed License No. NPF-8 Amendment No. XXX months if the time period since the most recent successful tracer gas test is greater than 3 years.

(c) The first performance of the periodic measurement of CRE pressure, Specification 5.5.18.d, shall be within 24 months, plus the 180 days allowed by SR 3.0.2, as measured from July 11, 2015, the date of the most recent successful pressure measurement test, or within 180 days if not performed previously.

Deleted per Amendment 144 (8) Deleted per Amendment 144 (9) Deleted per Amendment 144 (10) Deleted per Amendment 144 (11) Deleted per Amendment 144 (12) Deleted per Amendment 144 (13) Deleted per Amendment 144 (14) Deleted per Amendment 144 (15) Deleted per Amendment 144 (16) Deleted per Amendment 144 (17) Deleted per Amendment 144 (18) Deleted per Amendment 144 (19) Deleted per Amendment 144 (20) Deleted per Amendment 144 (21) Deleted per Amendment 144 (22) Additional Conditions The Additional conditions contained in Appendix C, as revised through Amendment No. 137, are hereby incorporated in the renewed license.

The licensee shall operate the facility in accordance with the additional conditions.

Control RoomCREFS 3.7.10 3.7 PLANT SYSTEMS 3.7.10 Control Room Emergency Filtration/Pressurization System (CREFS)

LCO 3.7.10 Two Control Room Emergency Filtration/Pressurization System (CREFS) trains and the Control Room Envelope (CRE) shall be OPERABLE.

NOTE --------------------------------------------- The control room envelope (CRE) boundary may be opened intermittently under administrative control.

APPLICABILITY:

MODES 1, 2, 3, and 4, During movement of irradiated fuel assemblies, During CORE ALTERATIONS.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CREFS train inoperable for reasons other than Condition B. A.1 Restore CREFS train to OPERABLE status.

7 days B. One or more CREFS trains inoperable due to inoperable CRE inoperableboundary. B.1 Initiate actions to implement mitigating actions. AND B.2.1 Restore CRE to OPERABLE status OR B.2.2.1 Verify mitigating actions ensure CRE occupant exposures to radiological, chemical, and smoke hazards will not exceed limitsGeneral Design Criteria (GDC) 19 met using mitigating actions in B.1. AND B.32.2.2 Restore CRE boundary to OPERABLE status.

Immediately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 hours 3090 days Control RoomCREFS 3.7.10 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and associated Completion Time of Condition A or B not met in MODE 1, 2, 3, or 4. C.1 Be in MODE 3.

AND C.2 ------------NOTE-----------

LCO 3.0.4.a is not applicable when entering MODE 4. --------------------------------

Be in MODE 4.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D. Two CREFS trains inoperable in MODE 1, 2, 3, OR 4. D.1 Be in MODE 3.

AND D.2 Be in MODE 5.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours E. Required Action and E.1 Place OPERABLE CREFS Immediately associated Completion Time of Condition A not met during movement of irradiated fuel assemblies or during CORE ALTERATIONS.

train in emergency recirculation mode.

OR E.2.1 Suspend CORE ALTERATIONS.

AND E.2.2 Suspend movement of irradiated fuel assemblies.

Immediately

Control RoomCREFS 3.7.10 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F. Required Action and associated Completion Time of Condition B not met during movement of irradiated fuel assemblies or during CORE ALTERATIONS.

OR Two CREFS trains inoperable during

movement of irradiated fuel assemblies or during CORE ALTERATIONS.

OR One or more CREFS trains inoperable due to an inoperable CRE boundary during movement of irradiated fuel assemblies or during CORE ALTERATIONS.

F.1 Suspend CORE ALTERATIONS.

AND F.2 Suspend movement of irradiated fuel assemblies.

Immediately Immediately SURVEILLANCE REQUREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.1 Operate each CREFS Pressurization train with the heaters operating and each CREFS Recirculation and Filtration train for 15 minutes.

In accordance with the Surveillance Frequency Control Program SR 3.7.10.2 Perform required CREFS filter testing in accordance with the Ventilation Filter Testing Program (VFTP).

In accordance with the VFTP Control RoomCREFS 3.7.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.3

NOTE---------------------------------

Not required to be performed in MODES 5 and 6.

Verify each CREFS train actuates on an actual or simulated actuation signal. In accordance with the Surveillance Frequency Control Program SR 3.7.10.4 Perform required CRE unfiltered air inleakage testing in accordance with the Control Room Envelope Habitability ProgramVerify CRE p within limits in the Control Room Integrity Program (CRIP). In accordance with the Control Room Envelope Habitability Program 2 4 months on a STAGGERED TEST BASIS SR 3.7.10.5 Verify CRE integrity in accordance with the CRIP.

In accordance with the CRIP

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.18 Control Room Integrity Program (CRIP)

A Control Room Integrity Program (CRIP) shall be established and implemented to ensure that the control room integrity is maintained such that a radiological event, hazardous chemicals, or a fire challenge (e.g., fire byproducts, halon, etc.) will not prevent the control room operators from controlling the reactor during normal or accident conditions. The program shall require testing as outlined below. Testing should be performed when changes are made to structures, systems and components which could impact Control Room Impact (CRE) integrity. These structures, systems and components may be internal or external to the CRE. Testing should also be conducted following a modification or a repair that could affect CRE inleakage. Testing should also be performed if the conditions associated with a particular challenge result in a change in operating mode, system alignment or system response that could result in a new limiting condition. Testing should be commensurate with the type and degree of modification or repair. Testing should be conducted in the alignment that results in the greatest consequence to the operators.

A CRIP shall be established to implement the following: a. Demonstrate, using Regulatory Guide (RG) 1.197 and ASTM E741, that CRE inleakage is less than the below values. The values listed below do not include 10 cfm assumed in accident analysis for ingress / egress.

i) 43 cfm when the control room ventilation systems are aligned in the emergency recirculation mode of operation, ii) 600 cfm when the control room ventilation systems are aligned in the isolation mode of operation, and iii) 2,340 cfm when the control room ventilation systems are aligned in the normal mode of operation;

b. Demonstrate that the leakage characteristics of the CRE will not result in simultaneous loss of reactor control capability from the control room and the hot shutdown panels;

c. Maintain a CRE configuration control and a design and licensing bases control program and a preventative maintenance program. As a minimum, the CRE configuration control program will determine whether the i) CRE differential pressure relative to adjacent areas and ii) the control room ventilation system flow rates, as determined in accordance with ASME N510-1989 or ASTM E2029

-99, are consistent with the values measured at the time the ASTM E741 test was performed. If item i or ii has changed, determine how this change has affected the inleakage characteristics of the CRE. If there has been degradation in the inleakage characteristics of the Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.18 Control Room Integrity Program (CRIP)

(continued)

CRE since the E741 test, then a determination should be made whether the licensing basis analyses remain valid. If the licensing basis analyses remain valid, the CRE remains OPERABLE.

d. Test the CRE in accordance with the testing methods and at the frequencies specified in RG 1.197, Revision 0, May 2003.

The provisions of SR 3.0.2 are applicable to the control room inleakage testing frequencies.

5.5.18 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Filtration System (CREFS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:

a. The definition of the CRE and the CRE boundary.

b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREFS, operating at the flow rate required by the VFTP, at a Frequency of 24 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 24 month assessment of the CRE boundary.

e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c.

Programs and Manuals 5.5 The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

Joseph M. Farley Nuclear Plant - Units 1 and 2 Alternative Source Term License Amendment Request Enclosure 3 Bases Pages (Markup) (For information only)

RCS Pressure SL B 2.1.2 Farley Units 1 and 2 B 2.1.2-1 Revision 0 B 2.0 SAFETY LIMITS (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL

BASES BACKGROUND The SL on RCS pressure protects the integrity of the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves

as the primary barrier in preventing the release of fission products into the atmosphere. By establishing an upper limit on RCS pressure, the continued integrity of the RCS is ensured. According to 10 CFR 50, Appendix A, GDC 14, "Reactor Coolant Pressure Boundary," and GDC 15, "Reactor Coolant System Design" (Ref.

1), the reactor coolant pressure boundary (RCPB) design conditions are not to be exceeded during normal operation and anticipated operational occurrences (AOOs). Also, in accordance with GDC 28, "Reactivity Limits" (Ref.

1), reactivity accidents, including rod ejection, do not result in damage to the RCPB greater than limited local yielding.

The design pressure of the RCS is 2500 psia. During normal operation and AOOs, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref.

2). To ensure system integrity, all RCS components were hydrostatically tested at 125% of design pressure, according to the ASME Code requirements prior to initial operation when there was no fuel in the core. Following inception of unit operation, RCS components shall be pressure tested, in accordance with the requirements of ASME Code, Section XI (Ref. 3). Overpressurization of the RCS could result in a breach of the RCPB.

If such a breach occurs in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere, raising concerns relative to limits on radioactive releases specified in 10 CFR 50.67, "Accident Source Term100, "Reactor Site Criteria" (Ref. 4).

APPLICABLE The RCS pressurizer safety valves, the main steam safety valves SAFETY ANALYSES (MSSVs), and the reactor high pressure trip have settings established to ensure that the RCS pressure SL will not be exceeded.

(continued)

RCS Pressure SL B 2.1.2 Farley Units 1 and 2 B 2.1.2-3 Revision 0 BASES SAFETY LIMIT If the RCS pressure SL is violated when the reactor is in MODE 1 VIOLATIONS or 2, the requirement is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for radioactive releases in excess of 10 CFR 50.67, "Accident Source Term,"100, "Reactor Site Criteria," limits (Ref.

4).

The allowable Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of reducing power level to a MODE of operation where the potential for challenges to safety systems is minimized.

If the RCS pressure SL is exceeded in MODE 3, 4, or 5, RCS pressure must be restored to within the SL value within 5 minutes. Exceeding the RCS pressure SL in MODE 3, 4, or 5 is more severe than exceeding this SL in MODE 1 or 2, since the reactor vessel temperature may be lower and the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SL within 5 minutes. The action does not require reducing MODES, since this would require reducing temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure stress. REFERENCES

1. 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC

28. 2. ASME, Boiler and Pressure Vessel Code, Section III, Article NB-7000. 3. ASME, Boiler and Pressure Vessel Code, Section XI, Article IWX-5000. 4. 10 CFR 50.67100.

5. FSAR. Section 7.2.

SDM B 3.1.1 Farley Units 1 and 2 B 3.1.1-4 Revision 0 BASES APPLICABLE SDM satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). Even though SAFETY ANALYSES it is not directly observed from the control room, SDM is considered (continued) an initial condition process variable because it is periodically monitored to ensure that the unit is operating within the bounds of accident analysis assumptions.

With Tavg less than 200°F, the reactivity transients resulting from a postulated steam line break cooldown are minimal, and a 1% delta k/k SHUTDOWN MARGIN provides adequate protection.

LCO SDM is a core design condition that can be ensured during operation through control rod positioning (control and shutdown banks) and through the soluble boron concentration.

The MSLB (Ref.

2) and the boron dilution (Ref.

3) accidents are the most limiting analyses that establish the SDM value of the LCO. For MSLB accidents, if the LCO is violated, there is a potential to exceed the DNBR limit and to exceed 10 CFR 50.67, "Accident Source Term,"100, "Reactor Site Criteria," limits (Ref.

4). For the boron dilution accident, if the LCO is violated, the minimum required time assumed for operator action to terminate dilution may no longer be applicable.

APPLICABILITY In MODE 2 with keff < 1.0 and in MODES 3, 4, and 5, the SDM requirements are applicable to provide sufficient negative reactivity to meet the assumptions of the safety analyses discussed above. In MODE 6, the shutdown reactivity requirements are given in LCO 3.9.1, "Boron Concentration." In MODES 1 and 2, SDM is ensured by complying with LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits."

ACTIONS A.1 If the SDM requirements are not met, boration must be initiated promptly. A Completion Time of Immediately is adequate to ensure prompt operator action to correctly align and start the required

(continued)

SDM B 3.1.1 Farley Units 1 and 2 B 3.1.1-6 Revision 0 BASES SURVEILLANCE SR 3.1.1.1 (continued)

REQUIREMENTS e. Xenon concentration;

f. Samarium concentration; and

g. Isothermal temperature coefficient (ITC).

Using the ITC accounts for Doppler reactivity in this calculation because the reactor is subcritical, and the fuel temperature will be changing at the same rate as the RCS.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES

1. 10 CFR 50, Appendix A, GDC 26.

2. FSAR, Section 15.4.2. 3. FSAR, Section 15.2.4. 4. 10 CFR 50.67100. 5. Letter from D.E. McKinnon to L.K. Mathews, "Operating Procedure for Mode 4/5 Boron Dilution," 90 AP*-G-0041, July 6, 1990.

RTS Instrumentation B 3.3.1 Farley Units 1 and 2 B 3.3.1-1 Revision 0 B 3.3 INSTRUMENTATION B 3.3.1 Reactor Trip System (RTS) Instrumentation

BASES BACKGROUND The RTS initiates a unit shutdown, based on the values of selected unit parameters, to protect against violating the core fuel design limits and Reactor Coolant System (RCS) pressure boundary during anticipated operational occurrences (AOOs) and to assist the Engineered Safety Features (ESF) Systems in mitigating accidents.

The protection and monitoring systems have been designed to assure safe operation of the reactor. This is achieved by specifying limiting safety system settings (LSSS) in terms of parameters directly monitored by the RTS, as well as specifying LCOs on other reactor system parameters and equipment performance.

The LSSS, defined in this specification as the Trip Setpoints, in conjunction with the LCOs, establish the threshold for protective system action to prevent exceeding acceptable limits during Design Basis Accidents (DBAs).

During AOOs, which are those events expected to occur one or more times during the unit life, the acceptable limits are:

1. The Departure from Nucleate Boiling Ratio (DNBR) shall be maintained above the Safety Limit (SL) value to prevent departure from nucleate boiling (DNB);

2. Fuel centerline melt shall not occur; and

3. The RCS pressure SL of 2735 psig shall not be exceeded.

Operation within the SLs of Specification 2.0, "Safety Limits (SLs)," also maintains the above values and assures that offsite dose will be within the 10 CFR 50 and 10 CFR 100 criteria during AOOs.

Accidents are events that are analyzed even though they are not expected to occur during the unit life. The acceptable limit during accidents is that offsite dose shall be maintained within an acceptable fraction of 10 CFR 50.67100 limits. Different accident categories are allowed a different fraction of these limits, based on probability of (continued)

Containment Purge and Exhaust Isolation Instrumentation B 3.3.6 Farley Units 1 and 2 B 3.3.6-2 Revision 0 BASES APPLICABLE purge and exhaust isolation radiation monitors act as backup to the SI SAFETY ANALYSES signal to ensure closing of the purge and exhaust valves. They are (continued) also the primary means for automatically isolating containment in the event of a fuel handling accident during shutdown. Containment isolation in turn ensures meeting the containment leakage rate assumptions of the safety analyses, and ensures that the calculated accidental offsite radiological doses are below 10 CFR 50.67100 (Ref. 1) limits.

The containment purge and exhaust isolation instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The LCO requirements ensure that the instrumentation necessary to initiate Containment Purge and Exhaust Isolation, listed in Table 3.3.6-1, is OPERABLE.

1. Manual Initiation The LCO requires two channels OPERABLE. The operator can initiate Containment Purge Isolation at any time by using either of two valve hand switches in the control room (labeled CTMT PURGE DMPRS). Each switch actuates one train of purge/exhaust isolation valves. Actuation of either handswitch isolates the Containment Purge and Exhaust System.

The LCO for Manual Initiation ensures the proper amount of redundancy is maintained in the manual actuation circuitry to ensure the operator has manual initiation capability.

Each channel consists of one handswitch and the interconnecting wiring to the purge/exhaust isolation valves in that train.

2. Automatic Actuation Logic and Actuation Relays The LCO requires two trains of Automatic Actuation Logic and Actuation Relays OPERABLE to ensure that no single random failure can prevent automatic actuation.

Automatic Actuation Logic and Actuation Relays consist of the same features and operate in the same manner as described for ESFAS Function 1.b (Paragraph 1), SI, and ESFAS Function 3.a, (continued)

Containment Purge and Exhaust Isolation Instrumentation B 3.3.6 Farley Units 1 and 2 B 3.3.6-9 Revision 0 BASES SURVEILLANCE SR 3.3.6.7 REQUIREMENTS (continued)

The CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.6.8

This SR ensures the individual channel response times are less than or equal to the maximum values assumed in the safety analysis. The response time testing acceptance criteria are included in FSAR Table 7.3

-16 (Ref. 4). This surveillance is performed in accordance with the guidance provided in the ESF RESPONSE TIME surveillance requirement in LCO 3.3.2, ESFAS. REFERENCES

1. 10 CFR 50.67100.11. 2. Not used.

3. Not used.

4. FSAR Table 7.3

-16

PRF Actuation Instrumentation B 3.3.8 Farley Units 1 and 2 B 3.3.8-1 Revision 18 B 3.3 INSTRUMENTATION

B 3.3.8 Penetration Room Filtration (PRF) System Actuation Instrumentation BASES BACKGROUND The PRF ensures that radioactive materials in the Spent Fuel Pool Room atmosphere following a fuel handling accident or ECCS pump rooms and penetration rooms of the auxiliary building following a loss of coolant accident (LOCA) are filtered and adsorbed prior to exhausting to the environment. The system is described in the Bases for LCO 3.7.12, "Penetration Room Filtration System." The system initiates filtered ventilation of the Spent Fuel Pool Room (including isolation of the normal ventilation) automatically following receipt of a high radiation signal (gaseous) or a low air flow signal from the normal Spent Fuel Pool Room ventilation system. In addition, the system initiates filtered ventilation of the ECCS pump rooms and penetration rooms following receipt of a Phase B Containment Isolation signal.

Initiation may also be performed manually as needed from the main control room.

High gaseous radiation provides PRF initiation. Each PRF train is initiated by high radiation detected by a channel dedicated to that train. There are a total of two channels, one for each train. Each channel contains a gaseous monitor. High radiation detected by either monitor or a low air flow signal from the normal Spent Fuel Pool Room ventilation or a Phase B Containment Isolation signal from the Engineered Safety Features Actuation System (ESFAS) starts the PRF. These actions function to prevent exfiltration of contaminated air by initiating filtered ventilation, which imposes a negative pressure on the Spent Fuel Pool Room or ECCS pump rooms and penetration rooms. Since the radiation monitors include an air sampling system, various components such as sample line valves and sample pumps are required to support monitor OPERABILITY.

APPLICABLE The PRF ensures that radioactive materials in the Spent Fuel Pool SAFETY ANALYSES Room atmosphere following a fuel handling accident or ECCS pump rooms and penetration rooms following a LOCA are filtered and adsorbed prior to being exhausted to the environment. This action reduces the radioactive content in the plant exhaust following a LOCA or fuel handling accident so that offsite doses remain within the limits specified in 10 CFR 50.67100 (Ref. 1). (continued)

PRF Actuation Instrumentation B 3.3.8 Farley Units 1 and 2 B 3.3.8-9 Revision 52 BASES SURVEILLANCE SR 3.3.8.7 REQUIREMENTS (continued)

The CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES

1. 10 CFR 50.67100.11. 2. FNP - 1/2 - RCP - 252. 3. Not used.

RCS Operational LEAKAGE B 3.4.13 Farley Units 1 and 2 B 3.4.13-2 Revision 24 BASES APPLICABLE Except for primary to secondary LEAKAGE, the safety analyses do not SAFETY ANALYSES address operational LEAKAGE. However, other operational LEAKAGE is typically seen as a precursor to a LOCA; the amount of leakage can affect the probability of such an event. The safety analysis for an event resulting in steam discharge to the atmosphere assumes that primary to secondary LEAKAGE from all steam generators (SGs) is 1 gpm as a result of accident induced conditions. The LCO requirement to limit primary to secondary LEAKAGE through any one SG to less than or equal to 150 gpd (i.e. total leakage less than or equal to 450 gpd) is significantly less than the conditions assumed in the safety analysis (with leakage assumed to occur at room temperature in both cases). Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a main steam line break (MSLB) 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 fluid is released via the main steam safety valves. The majority of the activity released to the atmosphere results from the tube rupture. Therefore, the 1 gpm primary to secondary LEAKAGE safety analysis assumption is relatively inconsequential.

The MSLB is more limiting for primary to secondary LEAKAGE. The safety analysis for the MSLB assumes 0.35 gpm500 gpd and 0.65 gpm470 gpd primary to secondary LEAKAGE in the faulted and both intact steam generators respectively as an initial condition. The offsite dose consequences resulting from the MSLB accident are bounded by a small fraction (i.e., 10%) of the limits defined in 10 CFR 50.67100. The RCS specific activity assumed was 0.5 Ci/gm DOSE EQUIVALENT I

-131 at a conservatively high letdown flow of 145 gpm, with either a pre

-existing or an accident initiated iodine spike. These values bound the Technical Specifications values.

The RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

RCS Specific Activity B 3.4.16 Farley Units 1 and 2 B 3.4.16-1 Revision 15 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.16 RCS Specific Activity BASES BACKGROUND The maximum dose to the whole body and the thyroid that an individual at the exclusion areasite boundary can receive for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months during an accident, or for the duration of the accident at the Low Population Zone, is specified in 10 CFR 50.67100 (Ref. 1). The limits on specific activity ensure that the doses are held to an appropriate fraction of th e 10 CFR 50.67100 limits (i.e., a small fraction of or well within the 10 CFR 50.67100 limits depending on the specific accident analysis) during analyzed transients and accidents.

The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the offsite radioactivity dose consequences in the event of a steam generator tube rupture (SGTR) or main steam line break (MSLB) accident.

The LCO contains specific activity limits for both DOSE EQUIVALENT I-131 and gross specific activity. The allowable levels are intended to limit the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months dose at the exclusion areasite boundary, or at the low population zone outer boundary for the radiological release duration, to an appropriate fraction of the 10 CFR 50.67100 dose guideline limits. The limits in the LCO are standardized, based on parametric evaluations of offsite radioactivity dose consequences for typical site locations.

The parametric evaluations showed the potential offsite dose levels for a SGTR or MSLB accident were an appropriately small fraction of the 10 CFR 100 dose guideline limits. Each evaluation assumes a broad range of site applicable atmospheric dispersion factors in a parametric evaluation.

APPLICABLE The LCO limits on the specific activity of the reactor coolant ensures SAFETY ANALYSES that the resulting doses will not exceed an appropriate fraction of the

10 CFR 50.67100 dose guideline limits following a SGTR or MSLB accident. The SGTR and MSLB safety analyseis (Ref. 2 and 3) assumes the specific activity of the reactor coolant at 0.5 Ci/gm, a conservatively high letdown flow of 145 gpm, and a bounding reactor coolant steam generator (SG) tube leakage of 1 gpm total for three

(continued)

RCS Specific Activity B 3.4.16 Farley Units 1 and 2 B 3.4.16-2 Revision 15 BASES APPLICABLE SGs. The MSLB analysis assumes a steam generator tube leakage of SAFETY ANALYSES 500 gpd in the faulted loop and 470 gpd in each of the intact loops for (continued) a total leakage of 1440 gpd.

These is analyseis resulted in offsite doses bounded by a small fraction (i.e., 10%) of the 10 CFR 50.67100 guidelines using FGR No. 11 and 12ICRP 30 Dose Conversion Factors (DCFs). The initial RCS specific activity assumed was 0.5 Ci/gm DOSE EQUIVALENT I

-131 at a conservatively high letdown flow of 145 gpm with an iodine spike. These values bound the Technical Specifications values. The safety analysis assumes for both the SGTR and MSLB the specific activity of the secondary coolant at its limit of 0.1 Ci/gm DOSE EQUIVALENT I

-131 from LCO 3.7.16, "Secondary Specific Activity." The analysis for the SGTR and MSLB accident s establish es the acceptance limits for RCS specific activity. Reference to these is analyseis are is used to assess changes to the unit that could affect RCS specific activity, as they relate to the acceptance limits.

The SGTR analysis assumes an RCS coolant activity of 0.5 Ci/gm DOSE EQUIVALENT I

-131 at a conservatively high letdown flow of 145 gpm. The SGTR and MSLB analyseis consider s two cases of reactor coolant specific activity. One case assumes specific activity at 0.5 Ci/gm DOSE EQUIVALENT I

-131 at a conservatively high letdown flow of 145 gpm with an accident initiated iodine spike that increases the I

-131 activity release rate into the reactor coolant by a factor of 500 immediately after the accident. The second case assumes the initial reactor coolant iodine activity at 30 Ci/gm DOSE EQUIVALENT I

-131 due to a pre

-accident iodine spike caused by an RCS transient. These values bound the Technical Specifications values. In both cases, the noble gas activity in the reactor coolant assumes 1% failed fuel, which closely Ci/gm for gross specific activity.

The SGTR analysis also assumes a loss of offsite power coincident with a reactor trip. The SGTR causes a reduction in reactor coolant inventory. The reduction initiates a reactor trip from a low pressurizer pressure signal or an RCS overtemperature T signal. The coincident loss of offsite power causes the steam dump valves to close to protect the condenser. The rise in pressure in the ruptured (continued)

RCS Specific Activity B 3.4.16 Farley Units 1 and 2 B 3.4.16-3 Revision 15 BASES APPLICABLE SG discharges radioactively contaminated steam to the atmosphere SAFETY ANALYSES through the SG power operated relief valves and the main steam (continued) safety valves. The unaffected SGs remove core decay heat by venting steam to the atmosphere until the cooldown ends. The MSLB analysis assumes a double

-ended guillotine break of a main steamline outside of containment. The affected steam generator will rapidly depressurize and release both the radionuclides initially contained in the secondary coolant, and the primary coolant activity transferred via SG tube leakage, directly to the outside atmosphere. A portion of the iodine activity initially contained in the intact SGs and noble gas activity due to SG tube leakage is released to the atmosphere through either the SG atmospheric relief valves (ARVs) or the SG safety relief valves. The safety analysis assumes an accident initiated iodine spike and shows the radiological consequences of a MSLB accident are within a small fraction of the Reference 1 dose guideline limits. Operation with iodine specific activity levels greater than the LCO limit is permissible, if the pre

-accident activity levels do not exceed the limits shown in Figure 3.4.16-1, in the applicable specification, for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The MSLB safety analysis has pre

-accident iodine spiking levels up to 30 Ci/gm DOSE EQUIVALENT I

-131. The remainder of the above limit permissible iodine levels shown in Figure 3.4.16-1 are acceptable because of the low probability of a MSLB accident occurring during the established 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months time limit. The occurrence of a MSLB accident at these permissible levels could increase the site boundary dose levels, but still be within 10 CFR 50.67100 dose guideline limits.

The limits on RCS specific activity are also used for establishing standardization in plant personnel radiation protection practices.

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

LCO The specific iodine activity is limited to 0.5 Ci/gm DOSE EQUIVALENT I

-131 at a conservatively high letdown flow of 145 gpm for the SGTR analysis and for the MSLB analysis, and the gross specific activity in the reactor coolant is limited to the number of Ci/gm equal to 100 divided by (average disintegration energy of the sum of the average beta and gamma energies of the coolant (continued)

RCS Specific Activity B 3.4.16 Farley Units 1 and 2 B 3.4.16-4 Revision 33 BASES LCO nuclides). The limit on DOSE EQUIVALENT I

-131 ensures the thyroid (continued) dose to an individual during the Design Basis Accident (DBA) will be an appropriate fraction of the allowed thyroid dose. The limit on gross specific activity ensures the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months whole body dose to an individual at the site boundary during the DBA will be a small fraction of the allowed whole body dose. The SGTR (Ref.

2) and MSLB accident analyses show that the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months site boundary dose levels are within acceptable limits. Violation of the LCO may result in reactor coolant radioactivity levels that could, in the event of an SGTR or MSLB, lead to site boundary doses that exceed the dose guideline limits. APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS average temperature 500°F, operation within the LCO limits for DOSE EQUIVALENT I-131 and gross specific activity are necessary to contain the potential consequences of an SGTR or MSLB to within the acceptable site boundary dose values.

For operation in MODE 3 with RCS average temperature <

500°F, and in MODES 4 and 5, the release of radioactivity in the event of a SGTR is unlikely since the saturation pressure of the reactor coolant is below the lift pressure settings of the main steam safety valves.

ACTIONS A.1 and A.2

With the DOSE EQUIVALENT I

-131 greater than the LCO limit, samples at intervals of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months must be taken to demonstrate that the limits of Figure 3.4.16-1 are not exceeded. The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is required to obtain and analyze a sample. Sampling is done to continue to provide a trend.

The DOSE EQUIVALENT I

-131 must be restored to within limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is required, if the limit violation resulted from normal iodine spiking.

A Note permits the use of the provisions of LCO 3.0.4c. This allowance permits entry into the applicable MODE(S) while relying on the ACTIONS. This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plant remains at, or proceeds to power operation.

(continued)

RCS Specific Activity B 3.4.16 (continued)

Farley Units 1 and 2 B 3.4.16-6 Revision 52 BASES SURVEILLANCE SR 3.4.16.2 REQUIREMENTS (continued)

This Surveillance is performed in MODE 1 only to ensure iodine remains within limit during normal operation and following fast power changes when fuel failure is more apt to occur.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

The Frequency, between 2 and 6 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, is established because the iodine levels peak during this time following fuel failure; samples at other times would provide inaccurate results.

SR 3.4.16.3 A radiochemical analysis for determination is required with the plant operating in MODE 1 equilibrium conditions. The determination directly relates to the LCO and is required to verify plant operation within the specified gross activity LCO limit. The analysis for is a measurement of the average energies per disintegration for isotopes with half lives longer than 15 minutes, excluding iodines. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. This SR has been modified by a Note that indicates sampling is required to be performed within 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for at least

48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . This ensures that the radioactive materials are at equilibrium so the analysis for is representative and not skewed by a crud burst or other similar abnormal event.

REFERENCES 1. 10 CFR 50.67100.11, 1973.

2. FSAR, Section 15.4.3.

SG Tube Integrity B 3.4.17 Farley Units 1 and 2 B 3.4.17-2 Revision 60 BASES BACKGROUND The processes used to meet the SG performance criteria are defined (continued) by the Steam Generator Program Guidelines (Ref. 1).

APPLICABLE The steam generator tube rupture (SGTR) accident is the limiting SAFETY ANALYSES design basis event for SG tubes and avoiding an SGTR is the basis for this Specification. The analysis of a SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to 1 gpmthe operational LEAKAGE rate limits in LCO 3.4.13, "RCS Operational LEAKAGE," plus the leakage rate associated with a double

-ended rupture of a single tube. The accident analysis for a SGTR assumes the contaminated secondary fluid is released via the main steam safety valves. The majority of the activity released to the atmosphere results from the tube rupture.

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 atmosphere is based on the total primary to secondary LEAKAGE from all SGs of 1 gpm as a result of accident induced conditions. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT I

-131 is assumed to be equal to the LCO 3.4.16, "RCS Specific Activity," limits. 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 50.67100 (Ref. 3) or the NRC approved licensing basis (e.g., a small fraction of these limits).

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

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

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

(continued)

SG Tube Integrity B 3.4.17 Farley Units 1 and 2 B 3.4.17-7 Revision 60 BASES SURVEILLANCE SR 3.4.17.2 REQUIREMENTS During an SG inspection, any inspected tube that satisfies the Steam Generator Program plugging criteria is removed from service by plugging. The tube plugging criteria delineated in Specification 5.5.9 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 plugging 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 and Reference 6 provide 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 MODE 4 following a SG inspection" ensures that the Surveillance has been completed and all tubes meeting the plugging 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 19Not used. 3. 10 CFR 50.67100. 4. ASME Boiler and Pressure Vessel Code, Section III, Subsection NB.

5. Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Generator Tubes," August 1976.

6. EPRI TR-107569, "Pressurized Water Reactor Steam Generator Examination Guidelines."

ECCS Recirculation Fluid pH Control System B 3.5.6 Farley Units 1 and 2 B 3.5.6-3 Revision 0 BASES APPLICABLE be increased if the long term pH of the recirculation solution is not SAFETY ANALYSES adjusted to 7.5 or greater. Therefore, long term pH control of the (continued) post-LOCA recirculation fluid helps ensure the offsite and control room thyroid doses are within the limits of 10 CFR 50.67100 and 10 CFR 50, Appendix A, General Design Criterion 19 respectively. The Recirculation Fluid pH Control System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The OPERABILITY of the Recirculation Fluid pH Control System ensures sufficient TSP is maintained in the three TSP storage baskets to increase the long term recirculation fluid pH to between 7.5 and 10.5 following a LOCA. A pH range of 7.5 to 10.5 is sufficient to prevent significant amounts of iodine released from fuel failure and dissolved in the recirculation fluid, from converting to a volatile form and evolving from solution into the containment atmosphere during the ECCS recirculation phase. In addition, an alkaline pH in this range will minimize chloride induced stress corrosion cracking of austenitic stainless steel components, and minimize the hydrogen produced by the corrosion of galvanized surfaces and zinc

-based paints.

In order to achieve the desired pH range of 7.5 to 10.5 in the post

-LOCA recirculation solution a total of between 10,000 pounds (185 ft

3) and 12,900 pounds (215 ft

3) of TSP (or appropriate weights/volumes for equivalent compounds) is required. The required amount of TSP is determined considering the volume of water involved, the target pH range, and the density of different vendor types of TSP that are available. Although the amount of TSP required is based on mass, a required volume is verified since it is not feasible to weigh the entire amount of TSP in containment.

APPLICABILITY In MODES 1, 2, 3, and 4 a DBA could cause the release of radioactive material in containment requiring the operation of the ECCS Recirculation Fluid pH Control System. The ECCS Recirculation Fluid pH Control System assists in reducing the amount of radioactive material available for release to the outside atmosphere after a DBA.

(continued)

MSIVs B 3.7.2 Farley Units 1 and 2 B 3.7.2-4 Revision 0 BASES LCO This LCO provides assurance that the MSIVs will perform their design (continued) safety function to mitigate the consequences of accidents such thatthat could result in offsite exposures are less thancomparable to the 10 CFR 50.67100 (Ref. 4) limits.

APPLICABILITY The MSIVs must be OPERABLE in MODE 1, and in MODES 2 and 3 except when one MSIV in each steam line is closed, when there is significant mass and energy in the RCS and steam generators. When the MSIVs are closed, they are already performing the safety function.

In MODE 4, normally most of the MSIVs are closed, and the stea m generator energy is low.

In MODE 5 or 6, the steam generators do not contain much energy because their temperature is below the boiling point of water; therefore, the MSIVs are not required for isolation of potential high energy secondary system pipe breaks in these MODES.

ACTIONS A Note has been added to the ACTIONS to clarify the application of the Completion Time rules. The Conditions of this Specification may be entered independently for each steam line. The Completion Time(s) of the inoperable MSIV Systems will be tracked separately for each steam line starting from the time the Condition was entered for that steam line.

A.1 With one MSIV inoperable in one or more steam lines in MODE 1, action must be taken to restore the inoperable MSIV to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Some repairs to the MSIV can be made with the unit at power. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasonable, considering the low probability of an accident occurring during this time that would require the MSIVs to close and the remaining OPERABLE MSIV in the steam line. This Completion Time is also consistent with the Completion Times provided for a single inoperable train in other ESF systems that contain redundant trains of equipment.

(continued)

MSIVs B 3.7.2 Farley Units 1 and 2 B 3.7.2-7 Revision 19 BASES SURVEILLANCE SR 3.7.2.1 (continued)

REQUIREMENTS accident and containment analyses. This Surveillance is normally performed while returning the unit to operation following a refueling outage. The Frequency is in accordance with the Inservice Testing Program. Operating experience has shown that these components usually pass the Surveillance when performed in accordance with the Inservice Testing Program. Therefore, the Frequency is acceptable from a reliability standpoint.

This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. If desired, this allows a delay of testing until MODE 3, to establish conditions consistent with those under which the acceptance criterion was generated. This surveillance may be performed in lower modes but must be performed prior to entry into MODE 2.

REFERENCES

1. FSAR, Section 10.3.

2. FSAR, Section 6.2. 3. FSAR, Section 15.4.2. 4. 10 CFR 50.67100.11.

5. ASME, Boiler and Pressure Vessel Code, Section XI.

ARVs B 3.7.4 Farley Units 1 and 2 B 3.7.4-2 Revision 0 BASES APPLICABLE the unit to RHR entry conditions. The limiting design basis accident SAFETY ANALYSES for the ARVs is established by the Steam Generator Tube Rupture (continued)

(SGTR) event (Ref. 2). The SGTR event is analyzed for two cases to determine that the offsite doses meet the NRC acceptance criteria. That is, for the case of an accident initiated Iodine spike, the doses from the accident are a small fraction of the limits defined in 10 CFR 50.67100 and for the case of a pre

-accident Iodine spike, the doses from the accident are within the limits defined in 10 CFR 50.67100. The SGTR event assumes recovery with and without offsite power. The loss of offsite power assumption results in the ARVs being relied upon to reduce RCS temperature to recover from an SGTR and also to reduce RCS temperature and pressure to RHR entry conditions. The accident analysis does not assume a specific method of valve operation to mitigate the accident. The analysis assumes the SG tube break flow is terminated within 30 minutes of the initiation of the accident.

The recovery from the SGTR event requires a rapid cooldown to establish adequate subcooling as a necessary step to allow depressurization of the RCS to terminate the primary to secondary break flow in the ruptured steam generator. The time required to terminate the primary to secondary break flow in the SGTR event is more critical than the time required to cool the RCS down to RHR entry conditions for this event and other accident analyses. Thus, the SGTR is the limiting event for the ARVs.

Each ARV is equipped with two manual isolation valves in the event an ARV spuriously fails to open or fails to close during use.

The ARVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Three ARV lines are required to be OPERABLE. One ARV line is required from each of three steam generators to ensure that at least one ARV line is available to conduct a unit cooldown following an SGTR, in which one steam generator becomes unavailable, accompanied by a single, active failure of a second ARV line on an unaffected steam generator. At least one manual isolation valve must be OPERABLE to isolate a failed open ARV line. A closed manual isolation valve does not render it or its ARV line inoperable. The accident analysis does not model a specific method of valve operation and allows 30 minutes to terminate the SG tube break flow. Sufficient time is available to unisolate and manually operate the ARV.

(continued)

Farley Units 1 and 2 B 3.7.4-2 Revision 0 B 3.7 PLANT SYSTEMS B 3.7.10 Control Room

[Under development due to recent License Amendment

-related changes]

PRF B 3.7.12 Farley Units 1 and 2 B 3.7.12-2 Revision 0 BASES APPLICABLE The PRF System design basis is established by the consequences of SAFETY ANALYSES the limiting Design Basis Accidents (DBAs), which are a fuel handling accident and a large break loss of coolant accident (LOCA). The analysis of the fuel handling accident, given in Reference 3, assumes that all fuel rods in an assembly are damaged. The analysis of the LOCA assumes that radioactive materials leaked from the Emergency Core Cooling System (ECCS) are filtered and adsorbed by the PRF System. The PRF System also functions following a small break LOCA with a Phase B signal or manual operator actuation in those cases where the ECCS goes into the recirculation mode of long term cooling, to clean up releases of smaller leaks, such as from valve steam packing. The DBA analysis of the fuel handling accident and LOCA assumes that only one train of the PRF System is functional due to a single failure that disables the other train. The accident analysis accounts for the reduction in airborne radioactive material provided by the one remaining train of this filtration system. The amount of fission products available for release from the spent fuel pool room is determined for a fuel handling accident and ECCS leakage for a LOCA. The analysis of the effects and consequences of a fuel handling accident and a LOCA are presented in Reference 3. The assumptions and the analysis for the fuel handling accident follow the guidance provided in Regulatory Guide 1.1.183 25 (Ref. 4). The PRF System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Two independent and redundant trains of the PRF System are required to be OPERABLE to ensure that at least one train is available, assuming a single failure that disables the other train, coincident with a loss of offsite power. During movement of irradiated fuel in the spent fuel pool room both trains of PRF are required to be aligned to the spent fuel pool room. Total system failure could result in the atmospheric release from the spent fuel pool room or ECCS pump rooms exceeding 25% of the 10 CFR 50.67100 (Ref. 5) limits in the event of a fuel handling accident or LOCA respectively.

The PRF System is considered OPERABLE when the individual components necessary to control exposure in the spent fuel pool room, ECCS pump rooms, and penetration area are OPERABLE in both trains. A PRF train is considered OPERABLE when its associated:

a. Recirculation and exhaust fans are OPERABLE; (continued)

PRF B 3.7.12 Farley Units 1 and 2 B 3.7.12-7 Revision 52 BASES SURVEILLANCE SR 3.7.12.5 (continued)

REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.7.12.6 During the fuel handling mode of operation, the PRF is designed to maintain a slightly negative pressure in the spent fuel pool room with respect to atmospheric pressure and surrounding areas at a flow rate of 5,500 cfm, to prevent unfiltered leakage. The slightly negative pressure is verified by using a non

-rigorous method that yields some observable identification of the slightly negative pressure. Examples of non

-rigorous methods are smoke sticks, hand held differential pressure indicators, or other measurement devices that do not provide for an absolute measurement. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES

1. FSAR, Section 6.2.3.

2. FSAR, Section 9.4.2.

3. FSAR, Sections 15.4.1 and 15.4.5. 4. Regulatory Guide 1.183 25. 5. 10 CFR 50.67100.

6. ASME N510-1989.

Fuel Storage Pool Water Level B 3.7.13 Farley Units 1 and 2 B 3.7.13-1 Revision 0 B 3.7 PLANT SYSTEMS

B 3.7.13 Fuel Storage Pool Water Level

BASES BACKGROUND The minimum water level in the fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel.

A general description of the fuel storage pool design is given in the FSAR, Section 9.1.2 (Ref.

1). A description of the Spent Fuel Pool Cooling and Cleanup System is given in the FSAR, Section 9.1.3 (Ref. 2). The assumptions of the fuel handling accident are given in the FSAR, Section 15.4.5 (Ref.

3). APPLICABLE The minimum water level in the fuel storage pool meets the SAFETY ANALYSES assumptions of the fuel handling accident described in Regulatory Guide 1.18325 (Ref. 4). The resultant 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose per person at the exclusion areasite boundary is well within the 10 CFR 50.67100 (Ref. 5) limits.

According to Reference 4, there is 23 ft of water between the top of the damaged fuel bundle and the fuel pool surface during a fuel handling accident. With 23 ft of water, the assumptions of Reference 4 can be used directly. In practice, this LCO preserves this assumption for the bulk of the fuel in the storage racks. In the case of a single bundle dropped and lying horizontally on top of the spent fuel racks, however, there may be

< 23 ft of water between the top of the fuel bundle and the surface, indicated by the width of the bundle. To offset this small nonconservatism, the analysis assumes that all fuel rods fail, although analysis shows that only the first few rows fail from a hypothetical maximum drop.

The fuel storage pool water level satisfies Criteria 2 and 3 of 10 CFR 50.36(c)(2)(ii).

Fuel Storage Pool Water Level B 3.7.13 Farley Units 1 and 2 B 3.7.13-3 Revision 52 BASES SURVEILLANCE SR 3.7.13.1 REQUIREMENTS During refueling operations, the level in the fuel storage pool is in equilibrium with the refueling canal, and the level in the refueling canal is checked in accordance with SR 3.9.6.1 (refueling cavity water level verification).

REFERENCES

1. FSAR, Section 9.1.2.

2. FSAR, Section 9.1.3.

3. FSAR, Section 15.4.5. 4. Regulatory Guide 1.183 25, Rev. 0. 5. 10 CFR 50.67100.11.

Secondary Specific Activity B 3.7.16 Farley Units 1 and 2 B 3.7.16-1 Revision 5 B 3.7 PLANT SYSTEMS

B 3.7.16 Secondary Specific Activity

BASES BACKGROUND Activity in the secondary coolant results from steam generator tube outleakage from the Reactor Coolant System (RCS). Under steady state conditions, the activity is primarily iodines with relatively short half lives and, thus, indicates current conditions. During transients, I-131 spikes have been observed as well as increased releases of some noble gases. Other fission product isotopes, as well as activated corrosion products in lesser amounts, may also be found in the secondary coolant.

A limit on secondary coolant specific activity during power operation minimizes releases to the environment because of normal operation, anticipated operational occurrences, and accidents.

This limit is lower than the activity value that might be expected from a 450 gallons per day tube leak (LCO 3.4.13, "RCS Operational LEAKAGE") of primary coolant at the limit of 0.5 Ci/gm (LCO 3.4.16, "RCS Specific Activity"). The steam line failure is assumed to result in the release of the noble gas and iodine activity contained in the steam generator inventory, the feedwater, and the reactor coolant LEAKAGE. Most of the iodine isotopes have short half lives (i.e.,

< 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months ).

With the specified activity limit, the resultant 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose to a person at the site boundary would be within the limits of 10 CFR 20.1001- 20.2402 if the main steam safety valves (MSSVs) and Atmospheric Relief Valves (ARVs) are open for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following a trip from full power.

Operating at the allowable limits results in a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months exclusion areasite boundary exposure well within the 10 CFR 50.67100 (Ref. 1) limits.

APPLICABLE The accident analysis of the main steam line break (MSLB), SAFETY ANALYSES as discussed in the FSAR, Chapter 15 (Ref. 2) assumes the initial secondary coolant specific activity to have a radioactive isotope concentration of 0.10 Ci/gm DOSE EQUIVALENT I-131. This assumption is used in the analysis for determining the radiological (continued)

Secondary Specific Activity B 3.7.16 Farley Units 1 and 2 B 3.7.16-2 Revision 0 BASES APPLICABLE consequences of the postulated accident. The accident analysis, SAFETY ANALYSES based on this and other assumptions, shows that the radiological (continued) consequences of an MSLB do not exceed a small fraction of the exclusion aresite boundary dose limits (Ref.

1) for whole body and thyroid dose rates.

With the loss of offsite power, the remaining steam generators are available for core decay heat dissipation by venting steam to the atmosphere through the MSSVs and steam generator atmospheric relief valves (ARVs). The Auxiliary Feedwater System supplies the necessary makeup to the steam generators. Venting continues until the reactor coolant temperature and pressure have decreased sufficiently for the Residual Heat Removal System to complete the cooldown.

In the evaluation of the radiological consequences of this accident, the activity released from the steam generator connected to the failed steam line is assumed to be released directly to the environment. The unaffected steam generator is assumed to discharge steam and any entrained activity through the MSSVs and ARVs during the event. Since no credit is taken in the analysis for activity plateout or retention, the resultant radiological consequences represent a conservative estimate of the potential integrated dose due to the postulated steam line failure.

Secondary specific activity limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO As indicated in the Applicable Safety Analyses, the specific activity of the secondary coolant is required to be 0.10 Ci/gm DOSE EQUIVALENT I-131 to limit the radiological consequences of a Design Basis Accident (DBA) to a small fraction of the required limit (Ref. 1). Monitoring the specific activity of the secondary coolant in the steam generators ensures that when secondary specific activity limits are exceeded, appropriate actions are taken in a timely manner to place the unit in an operational MODE that would minimize the radiological consequences of a DBA.

Secondary Specific Activity B 3.7.16 Farley Units 1 and 2 B 3.7.16-3 Revision 52 BASES APPLICABILITY In MODES 1, 2, 3, and 4, the limits on secondary specific activity apply due to the potential for secondary steam releases to the atmosphere.

In MODES 5 and 6, the steam generators are not being used for heat removal. Both the RCS and steam generators are depressurized, and primary to secondary LEAKAGE is minimal. Therefore, monitoring of secondary specific activity is not required.

ACTIONS A.1 and A.2 DOSE EQUIVALENT I-131 exceeding the allowable value in the secondary coolant, is an indication of a problem in the RCS and contributes to increased post accident doses. If the secondary specific activity cannot be restored to within limits within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.16.1 REQUIREMENTS This SR verifies that the secondary specific activity in the steam generators is within the limits of the accident analysis. A gamma isotopic analysis of the secondary coolant, which determines DOSE EQUIVALENT I-131, confirms the validity of the safety analysis assumptions as to the source terms in post accident releases. It also serves to identify and trend any unusual isotopic concentrations that might indicate changes in reactor coolant activity or LEAKAGE. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES

1. 10 CFR 50.67100.11. 2. FSAR, Chapter 15.

Containment Penetrations B 3.9.3 Farley Units 1 and 2 B 3.9.3-1 Revision 44 B 3.9 REFUELING OPERATIONS B 3.9.3 Containment Penetrations

BASES BACKGROUND During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, a release of fission product radioactivity within containment will be limited to maintain dose consequences within regulatory limits when the LCO requirements are met. In MODES 1, 2, 3, and 4, this is accomplished by maintaining containment OPERABLE as described in LCO 3.6.1, "Containment." In MODE 6, the potential for containment pressurization as a result of an accident is not likely; therefore, requirements to isolate the containment from the outside atmosphere can be less stringent. The LCO requirements are referred to as "refueling integrity" rather than "containment OPERABILITY." Refueling integrity means that all potential escape paths are closed or capable of being closed. Since there is no potential for containment pressurization, the 10 CFR 50, Appendix J leakage criteria and tests are not required.

The containment serves to contain fission product radioactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained well within the requirements of 10 CFR 50.67100. Additionally, the containment provides radiation shielding from the fission products that may be present in the containment atmosphere following accident conditions.

The containment equipment hatch, which is part of the containment pressure boundary, provides a means for moving large equipment and components into and out of containment. If closed, the equipment hatch must be held in place by at least four bolts. Good engineering practice dictates that the bolts required by this LCO be approximately equally spaced. Alternatively, the equipment hatch can be open provided it can be installed with a minimum of four bolts holding it in place.

The containment air locks, which are also part of the containment pressure boundary, provide a means for personnel access during MODES 1, 2, 3, and 4 unit operation in accordance with LCO 3.6.2, "Containment Air Locks." Each air lock has a door at both ends. The doors are normally interlocked to prevent simultaneous opening when containment OPERABILITY is required. During periods of unit shutdown

(continued)

Containment Penetrations B 3.9.3 Farley Units 1 and 2 B 3.9.3-6 Revision 52 BASES BACKGROUND isolation valve, a manual isolation valve, blind flange, or equivalent.

(continued)

Equivalent isolation methods allowed under the provisions of 10 CFR 50.59 may include use of a material that can provide a temporary, atmospheric pressure, ventilation barrier for the other containment penetrations during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment (Ref.

1). APPLICABLE During CORE ALTERATIONS or movement of irradiated fuel SAFETY ANALYSES assemblies within containment, the most severe radiological consequences result from a fuel handling accident. The fuel handling accident is a postulated event that involves damage to irradiated fuel (Ref. 2). The fuel handling accident analyzed includes dropping a single irradiated fuel assembly. The requirements of LCO 3.9.6, "Refueling Cavity Water Level," and the minimum decay time of

100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to CORE ALTERATIONS ensure that the release of fission product radioactivity, subsequent to a fuel handling accident, results in doses that are less thanwell within the dose limitsguideline values specified in 10 CFR 50.67100, and the more restrictive offsite exposure criteria of. Standard Review Plan, Section 15.0.17.4, Rev. 1 (Ref. 3), defines "well within" 10 CFR 100 to be 25% or less of the 10 CFR 100 values. The acceptance limits for offsite radiation exposure will be 25% of 10 CFR 100 values.

Containment penetrations satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO This LCO limits the consequences of a fuel handling accident in containment by limiting the potential escape paths for fission product radioactivity released within containment. The LCO requires any penetration providing direct access from the containment atmosphere to the outside atmosphere to be closed except for the OPERABLE containment purge and exhaust penetrations, the equipment hatch and the personnel air locks. For the OPERABLE containment purge and exhaust penetrations, this LCO ensures that these penetrations are isolable by the Containment Purge and Exhaust Isolation System. For the equipment hatch and personnel air locks, closure capability is provided by a designated trained closure crew and the necessary equipment. The OPERABILITY requirements for LCO 3.3.6, "Containment Purge and Exhaust Isolation Instrumentation," ensure that the automatic purge and exhaust valve closure times specified in the FSAR can be achieved and, therefore, meet the assumptions used in the safety achieved and, therefore, meet the assumptions used in the safety analysis to ensure that releases through the valves

(continued)

Containment Penetrations B 3.9.3 Farley Units 1 and 2 B 3.9.3-6 Revision 52 BASES SURVEILLANCE SR 3.9.3.2 (continued)

REQUIREMENTS isolation time of each valve is in accordance with the Inservice Testing Program requirements. These Surveillances performed during MODE 6 will ensure that the valves are capable of closing after a postulated fuel handling accident to limit a release of fission product radioactivity from the containment.

SR 3.9.3.3 The equipment hatch is provided with a set of hardware, tools, and equipment for moving the hatch from its storage location and installing it in the opening. The required set of hardware, tools, and equipment shall be inspected to ensure that they can perform the required functions.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

The SR is modified by a Note which only requires that the surveillance be met for an open equipment hatch. If the equipment hatch is installed in its opening, the availability of the means to install the hatch is not required.

REFERENCES

1. GPU Nuclear Safety Evaluation SE

-0002000-001, Rev. 0, May 20, 1988.

2. FSAR, Section 15.4.5. 3. NUREG-0800, Section 15.0.17.4, Rev. 01, July 20001981. 4. Regulatory Guide 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," May 2003.

Refueling Cavity Water Level B 3.9.6 Farley Units 1 and 2 B 3.9.6-18 Revision 0 B 3.9 REFUELING OPERATIONS B 3.9.6 Refueling Cavity Water Level

BASES BACKGROUND The movement of irradiated fuel assemblies or performance of CORE ALTERATIONS, except during latching and unlatching of control rod drive shafts, within containment requires a minimum water level of 23 ft above the top of the reactor vessel flange. During refueling, this maintains sufficient water level in the containment, refueling canal, fuel transfer canal, refueling cavity, and spent fuel pool. Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident (Refs.

1 and 2). Sufficient iodine activity would be retained to limit offsite doses from the accident to less than< 25% of 10 CFR 50.67100 limits (Ref. 4), as well as the more restrictiveprovided by the guidance of Reference

3. APPLICABLE During CORE ALTERATIONS and movement of irradiated fuel SAFETY ANALYSES assemblies, the water level in the refueling canal and the refueling cavity is an initial condition design parameter in the analysis of a fuel handling accident in containment, as postulated by Regulatory Guide 1.183 25 (Ref. 1). A minimum water level of 23 ft (Regulatory Position C.1.c of Ref.

1) allows a decontamination factor of 2100 (Regulatory Position C.1.g of Ref.

1) to be used in the accident analysis for iodine. This relates to the assumption that 99% of the total iodine released from the pellet to cladding gap of all the dropped fuel assembly rods is retained by the refueling cavity water. The fuel pellet to cladding gap is assumed to contain 10% (except I

-131 is 12%) of the total fuel rod iodine inventory (Refs.

1 and 6). The fuel handling accident analysis inside containment is described in Reference 2. With a minimum water level of 23 ft and a minimum decay time of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to fuel handling, the analysis and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water and offsite doses are maintained within allowable limits (Refs.

3 and 4 and 5).

Refueling cavity water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

Refueling Cavity Water Level B 3.9.6 Farley Units 1 and 2 B 3.9.6-3 Revision 52 BASES SURVEILLANCE SR 3.9.6.1 (continued)

REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES

1. Regulatory Guide 1.18325, March 23, 1972 , July 2000. 2. FSAR, Section 15.4.5. 3. NUREG-0800, Section 15.0.17.4. 4. 10 CFR 50.67100.10. 5. Malinowski, D. D., Bell, M. J., Duhn, E., and Locante, J., WCAP-828, Radiological Consequences of a Fuel Handling Accident, December 1971. 6. NUREG/CR 5009.

Farley Units 1 and 2 B 3.9.6-3 Revision 52

Joseph M. Farley Nuclear Plant - Units 1 and 2 Alternative Source Term License Amendment Request Enclosure 4 Operating License and Technical Specification Pages (Retyped)

Farley - Unit 1 Renewed License No. NPF-2 Amendment No. ___

(5) Updated Final Safety Analysis Report Supplement The Updated Final Safety Analysis Report supplement, as revised, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4) following issuance of this renewed license. Until that update is complete, Southern Nuclear may make changes to the programs and activities

described in the supplement without prior Commission approval, provided that Southern Nuclear evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements of that section.

The Southern Nuclear Updated Final Safety Analysis Report supplement, submitted pursuant to 10 CFR 54.21(d), describes certain future activities to be completed prior to the period of extended operation. Southern Nuclear shall complete these activities no later than June 25, 2017, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

(6) Reactor Vessel Material Surveillance Capsule s All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of American Society for Testing and Materials (ASTM) E 185

-82 to the extent practicable for the configuration of the specimens in the capsule.

Any changes to the capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion.

(7) Upon implementation of Amendment No.

___ adopting TSTF

-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.10.4, in accordance with TS 5.5.18.c.(i), the assessment of CRE habitability as required by Specification 5.5.18.c.(ii), and the measurement of CRE pressure as required by Specification 5.5.18.d, shall be considered met. Following implementation:

(a) The first performance of SR 3.7.10

.4, in accordance with Specification 5.5.18.c.(i), shall be within the specified Frequency of 6 years, plus the 18

-month allowance of SR 3.0.2, as measured from February 1 through 8, 2016, the date of the most recent successful tracer gas test, as stated i n the August 25, 2004 letter response to Generic Letter 2003

-01, or within the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 years.

(b) The first performance of the periodic assessment of CRE habitability, Specification 5.5.18.c.(ii), shall be within 3 years, plus the 9

-month Farley - Unit 1 Renewed License No. NPF-2 Amendment No. ___

allowance of SR 3.0.2, as measured from February 1 through 8, 2016, the date of the most recent successful tracer gas test, as stated in the August 25, 2004 letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

(c) The first performance of the periodic measurement of CRE pressure, Specification 5.5.18.d, shall be within 24 months, plus the 180 days allowed by SR 3.0.2, as measured from July 11, 2015, the date of the most recent successful pressure measurement test, or within 180 days if not performed previously.

D. Southern Nuclear shall fully implement and maintain in effect all provisions of the Commission

-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to

provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

The plan, which contain Safeguards Information protected under 10 CFR 73.21, is entitled:

"Southern Nuclear Operating Company Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan," and was submitted on May 15, 2006. Southern Nuclear shall fully implement and maintain in effect all provisions of the Commission

-approved cyber security (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Southern Nuclear CSP was approved by License Amendment No. 186, as supplemented by a change approved by License Amendment No. 199. E. This renewed license is subject to the following additional conditions for

the protection of the environment:

(1) Southern Nuclear shall operate the facility within applicable Federal and State air and water quality standards and the Environmental Protection Plan (Appendix B).

(2) Before engaging in an operational activity not evaluated by the Commission, Southern Nuclear will prepare and record an environmental evaluation of such activity.

When the evaluation indicates that such activity may result in a significant adverse environmental impact that was

not evaluated, or that is significantly greater than evaluated in the Final Environmental Statement, Southern Nuclear shall provide a written evaluation of such activities and obtain prior approval of the Director, Office of Nuclear Reactor Regulation, for the activities.

F. Alabama Power Company shall meet the following antitrust conditions:

(1) Alabama Power Company shall recognize and accord to Alabama Electric Cooperative (AEC) the status of a competing electric utility in central and southern Alabama.

Farley - Unit 1 Renewed License No. NPF-2 Amendment No. ___

(2) Alabama Power Company shall offer to sell to AEC an undivided ownership interest in Units 1 and 2 of the Farley Nuclear Plant.

The percentage of ownership interest to be so offered shall be an amount based on the relative sizes of the respective peak loads of AEC and the Alabama Power Company (excluding from the Alabama Power Company

's peak load that amount imposed by members of AEC upon the electric system of Alabama Power Company) occurring in 1976. The price to be paid by AEC for its proportionate share of Units 1 and 2, determined in accordance with the foregoing formula, will be established by the parties through good faith negotiations.

The price shall be sufficient to fairly reimburse Alabama Power Company for the proportionate share of its total costs related to the Units 1 and 2 including, but not limited to, all costs of construction, installation, ownership and licensing, as of a date, to be agreed to by the two parties, which fairly accommodates both their respective interests.

The offer by Alabama Power Company to sell an undivided ownership interest in Units 1 and 2 may be conditioned, at Alabama Power Company's option, on the agreement by AEC to waive any right of partition ofthe Farley Plant and to avoid interference in the day

-to-day operation of the plant.

(3) Alabama Power Company will provide, under contractual arrangements between Alabama Power Company and AEC, transmission services via its electric system (a) from AEC

's electric system to AEC

's off-system members; and (b) to AEC

's electric system from electric systems other than Alabama Power Company

's and from AEC

's electric system to electric systems other than Alabama Power Company

's. The contractual arrangements covering such transmission services shall embrace rates and charges reflecting conventional accounting and ratemaking concepts followed by the Federal Energy Regulatory Commission (or its successor in function) in testing the reasonableness of rates and charges for transmission services. Such contractual arrangements shall contain provisions protecting Alabama Power Company against economic detriment resulting from transmission line or transmission losses associated therewith.

(4) Alabama Power Company shall furnish such other bulk power supply services as are reasonably available from its system.

(5) Alabama Power Company shall enter into appropriate contractual arrangements amending the 1972 Interconnection Agreement as last amended to provide for a reserve sharing arrangement between Alabama Power Company and AEC under which Alabama Power Company will provide reserve generating capacity in accordance with practices applicable to its responsibility to the operating companies of the Southern Company System.

AEC shall maintain a minimum level expressed as a percentage of coincident peak one

-hour kilowatt load equal to the percent reserve level similarly expressed for Alabama

Farley - Unit 1 Renewed License No. NPF-2 Amendment No. ___

Power Company as determined by the Southern Company System under its minimum reserve criterion then in effect. Alabama Power Company shall provide to AEC such data as needed from time to time to demonstrate the basis for the need for such minimum reserve level.

(6) Alabama Power Company shall refrain from taking any steps, including but not limited to, the adoption of restrictive provisions in rate filings or negotiated contracts for the sale of wholesale power, that serve to prevent any entity or group of entities engaged in the retail sale of firm electric power from fulfilling all or part of their bulk power requirements through self

-generation or through purchases from some other source other than Alabama Power Company.

Alabama Power Company shall further, upon request and subject to reasonable terms and conditions, sell partial requirements power to any such entity. Nothing in this paragraph shall be construed as preventing an applicant from taking reasonable steps, in accord with general practice in the industry, to ensure that the reliability of its system is not endangered by any action called for herein.

(7) Alabama Power Company shall engage in wheeling for and at the request of any municipally

-owned distribution system:

a. of electric energy from delivery points of Alabama Power Company to said distribution system(s); and

b. of power generated by or available to a distribution system as a result of its ownership or entitlement2 in generating facilities, to delivery points of Alabama Power Company designated by the distribution system.

Such wheeling services shall be available with respect to any unused capacity on the transmission lines of Alabama Power Company, the use of which will not jeopardize Alabama Power Company

's system.

The contractual arrangements covering such wheeling services shall be determined in accordance with the principles set forth in Condition (3) herein. Alabama Power Company shall make reasonable provisions for disclosed transmission requirements of any distribution system(s) in planning future transmission. "Disclosed" means the giving of reasonable advance notification of future requirements by said distribution system(s) utilizing wheeling services to be made available by Alabama Power Company.

(8) The foregoing conditions shall be implemented in a manner consistent with the provisions of the Federal Power Act and the Alabama Public Utility laws and regulations thereunder and all rates, charges, services or

___________________

2 "Entitlement" includes, but is not limited to, power made available to an entity pursuant to an exchange agreement.

Farley - Unit 1 Renewed License No. NPF-2 Amendment No. ___

practices in connection therewith are to be subject to the approval of regulatory agencies having jurisdiction over them.

Southern Nuclear shall not market or broker power or energy from Joseph M. Farley Nuclear Plant, Units 1 and 2. Alabama Power Company shall continue to be responsible for compliance with the obligations imposed on

it by the antitrust conditions contained in this paragraph 2.F. of the renewed license. Alabama Power Company shall be responsible and accountable for the actions of its agent, Southern Nuclear, to the extent said agent

's actions may, in any way, contravene the antitrust conditions of this paragraph 2.F.

G. Mitigation Strategy License Condition The licensee shall develop and maintain strategies for addressing large fires and explosions that include the following key areas:

(a) Fire fighting response strategy with the following elements:

1. Pre-defined coordinated fire response strategy and guidance 2. Assessment of mutual aid fire fighting assets

3. Designated staging areas for equipment and materials

4. Command and control

5. Training of response personnel (b) Operations to mitigate fuel damage considering the following:

1. Protection and use of personnel assets

2. Communications

3. Minimizing fire spread

4. Procedures for implementing integrated fire response strategy

5. Identification of readily

-available pre

-staged equipment

6. Training on integrated fire response strategy (c) Actions to minimize release to include consideration of:

1. Water spray scrubbing

2. Dose to onsite responders H. In accordance with the requirement imposed by the October 8, 1976 order of the United States Court of Appeals for the District of Columbia Circuit in Natural Resources Defense Council vs. Nuclear Regulatory Commission , No. 74-1385 and 74

-1586, that the Nuclear Regulatory Commission "shall make any licenses granted between July 21, 1976 and such time when the mandate is issued subject to the outcome of such proceeding herein," this renewed license shall be subject to the outcome of such proceedings.

Farley - Unit 1 Renewed License No. NPF-2 Amendment No. ___

I. This renewed operating license is effective as of the date of issuance and shall expire at midnight on June 25, 2037.

FOR THE NUCLEAR REGULATORY COMMISSION J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments:

1. Appendix A - Technical Specifications

2. Preoperational Tests, Startup Tests and Other Items Which Must Be Completed Prior to Proceeding to Succeeding Operational Modes

3. Appendix B - Environmental Protection Plan

4. Appendix C - Additional conditions

Date of Issuance: May 12, 2005

Farley - Unit 2 Renewed License No. NPF-2 Amendment No. ___

to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense

-in-depth and safety margins are maintained when changes are made to the fire protection program.

c. Transition License Conditions

1) Before achieving full compliance with 10 CFR 50.48(c), as specified by 2) below, risk

-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in 2) above.

2) The licensee shall implement the modifications to its facility, as described in Attachment S, Table S

-2, "Plant Modifications Committed," of SNC letter NL 1273, dated August 29, 2014, to complete the transition to full compliance with 10 CFR 50.48(c) by November 6, 2017. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

3) The licensee shall implement the items as listed in Attachment S, Table S

-3, "Implementation Items," of SNC letter NL-14-1273, dated August 29, 2014, within 180 days after NRC approval, except for items 30 and 32. Items 30 and 32 shall be implemented by February 6, 2018.

(7) Upon implementation of Amendment No.

___ adopting TSTF

-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.10.4, in accordance with TS 5.5.18.c.(i), the assessment of CRE habitability as required by Specification 5.5.18.c.(ii), and the measurement of CRE pressure as required by Specification 5.5.18.d, shall be considered met. Following implementation:

(a) The first performance of SR 3.7.10.4, in accordance with Specification 5.5.18.c.(i), shall be within the specified Frequency of 6 years, plus the 18

-month allowance of SR 3.0.2, as measured from February 1 through 8, 2016, the date of the most recent successful tracer gas test, as stated in the August 25, 2004 letter response to Generic Letter 2003

-01, or within the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 years.

(b) The first performance of the periodic assessment of CRE habitability, Specification 5.5.18.c.(ii), shall be within 3 years, plus the 9

-month allowance of SR 3.0.2, as measured from February 1 through 8, 2016, the date of the most recent successful tracer gas test, as stated in the August 25, 2004 letter response to Generic Letter

Farley - Unit 2 Renewed License No. NPF-2 Amendment No. ___

2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

(c) The first performance of the periodic measurement of CRE pressure, Specification 5.5.18.d, shall be within 24 months, plus the 180 days allowed by SR 3.0.2, as measured from July 11, 2015, the date of the most recent successful pressure measurement test, or within 180 days if not performed previously. (8) Deleted per Amendment 144 (9) Deleted per Amendment 144 (10) Deleted per Amendment 144 (11) Deleted per Amendment 144 (12) Deleted per Amendment 144 (13) Deleted per Amendment 144 (14) Deleted per Amendment 144 (15) Deleted per Amendment 144 (16) Deleted per Amendment 144 (17) Deleted per Amendment 144 (18) Deleted per Amendment 144 (19) Deleted per Amendment 144 (20) Deleted per Amendment 144 (21) Deleted per Amendment 144 (22) Additional Conditions The Additional conditions contained in Appendix C, as revised through Amendment No. 137, are hereby incorporated in the renewed license.

The licensee shall operate the facility in accordance with the additional conditions.

(23) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4) following issuance of this renewed license.

Until that update is complete, Southern Nuclear may make changes to the programs and activities described in the supplement without prior Commission approval, provided that Southern Nuclear evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements of that section.

The Southern Nuclear Updated Final Safety Analysis Report supplement, submitted pursuant to 10 CFR 54.21(d), describes certain future activities to be completed prior to the period of extended operation.

Southern Nuclear shall complete these activities no later than June 25, 2017, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

Farley - Unit 2 Renewed License No. NPF-2 Amendment No. ___

(24) Reactor Vessel Material Surveillance Capsules All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of American Society for Testing and Materials (ASTM) E 185

-82 to the extent practicable for the configuration of the specimens in the capsule. Any changes to the capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation.

All capsules placed in storage must be maintained for future insertion.

D. Southern Nuclear shall fully implement and maintain in effect all provisions of the Commission

-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plan, which contain Safeguards Information protected under 10 CFR 73.21, is entitled:

"Southern Nuclear Operating Company Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan," and was submitted on May 15, 2006. Southern Nuclear shall fully implement and maintain in effect all provisions of the Commission

-approved cyber security (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Southern Nuclear CSP was approved by License Amendment No. 181, as supplemented by a change approved by License Amendment No. 195. E. Deleted per Amendment 144 F. Alabama Power Company shall meet the following antitrust conditions:

(1) Alabama Power Company shall recognize and accord to Alabama Electric Cooperative (AEC) the status of a competing electric utility in central and southern Alabama.

(2) Alabama Power Company shall offer to sell to AEC an undivided ownership interest in Units 1 and 2 of the Farley Nuclear Plant. The percentage of ownership interest to be so offered shall be an amount based on the relative sizes of the respective peak loads of AEC and Alabama Power

Company (excluding from the Alabama Power Company

's peak load that amount imposed by members of AEC upon the electric system of Alabama Power Company) occurring in 1976. The price to be paid by AEC for its proportionate share of Units 1 and 2, determined in accordance with the foregoing formula, will be established by the parties through good faith negotiations.

The price shall be sufficient to fairly reimburse Alabama Power Company for the proportionate share of its total costs related to the Units 1 and 2 including, but not limited to, all costs of construction, Farley - Unit 2 Renewed License No. NPF-2 Amendment No. ___

installation, ownership and licensing, as of a date, to be agreed to by the two parties, which fairly accommodates both their respective interests.

The offer by Alabama Power Company to sell an undivided ownership

interest in Units 1 and 2 may be conditioned, at Alabama Power Company's option, on the agreement by AEC to waive any right of partition of the Farley Plant and to avoid interference in the day-to-day operation of the plant. (3) Alabama Power Company will provide, under contractual arrangements between Alabama Power Company and AEC, transmission services via its electric system (a) from AEC

's electric system to AEC

's off-system members; and (b) to AEC

's electric system from electric systems other than Alabama Power Company

's, and from AEC

's electric system to electric systems other than Alabama Power Company

's. The contractual arrangements covering such transmission services shall embrace rates and charges reflecting conventional accounting and ratemaking concepts followed by the Federal Energy Regulatory Commission (or its successor in function) in testing the reasonableness of rates and charges for transmission services. Such contractual arrangements shall contain provisions protecting Alabama Power Company against economic detriment resulting from transmission line or transmission losses associated therewith.

(4) Alabama Power Company shall furnish such other bulk power supply services as are reasonably available from its system.

(5) Alabama Power Company shall enter into appropriate contractual arrangements amending the 1972 Interconnection Agreement as last amended to provide for a reserve sharing arrangement between Alabama Power Company and AEC under which Alabama Power Company will provide reserve generating capacity in accordance with practices applicable to its responsibility to the operating companies of the Southern Company System. AEC shall maintain a minimum level expressed as a percentage of coincident peak one

-hour kilowatt load equal to the percent reserve level similarly expressed for Alabama Power Company as determined by the Southern Company System under its minimum reserve criterion then in effect. Alabama Power Company shall provide to AEC such data as needed from time to time to demonstrate the basis for the need for such minimum reserve level.

(6) Alabama Power Company shall refrain from taking any steps, including but not limited, to the adoption of restrictive provisions in rate filings or negotiated contracts for the sale of wholesale power, that serve to prevent any entity or group of entities engaged in the retail sale of firm electric power from fulfilling all or part of their bulk power requirements through se lf-generation or through purchases from some other source other than Alabama Power Company.

Alabama Power Company shall further, upon request and subject to reasonable terms and conditions, sell partial requirements power to any such entity. Nothing in this paragraph shall be

Farley - Unit 2 Renewed License No. NPF-2 Amendment No. ___

construed as preventing an applicant from taking reasonable steps, in accord with general practice in the industry, to ensure that the reliability of its system is not endangered by any action called for herein.

(7) Alabama Power Company shall engage in wheeling for and at the request of any municipally

-owned distribution system:

a. of electric energy from delivery points of Alabama Power Company to said distribution system(s); and

b. of power generated by or available to a distribution system as a result of its ownership or entitlement in generating facilities, to delivery points of Alabama Power Company designated by the distribution system.

Such wheeling services shall be available with respect to any unused capacity on the transmission lines of Alabama Power Company, the use of which will not jeopardize Alabama Power Company's system.

The contractual arrangements covering such wheeling services shall be determined in accordance with the principles set forth in Condition (3) herein.

Alabama Power Company shall make reasonable provisions for disclosed transmission requirements of any distribution system(s) in planning future transmission. "Disclosed" means the giving of reasonable advance notification of future requirements by said distribution system(s) utilizing wheeling services to be made available by Alabama Power Company.

(8) The foregoing conditions shall be implemented in a manner consistent with the provisions of the Federal Power Act and the Alabama Public Utility laws and regulations thereunder and all rates, charges, services or practices in connection therewith are to be subject to the approval of regulatory agencies having jurisdiction over them.

Southern Nuclear shall not market or broker power or energy from Joseph M. Farley Nuclear Plant, Units 1 and 2.

Alabama Power Company shall continue to be responsible for compliance with the obligations imposed on it by the antitrust conditions contained in this paragraph 2.F. of the renewed license.

Alabama Power Company shall be responsible and accountable for the actions of its agent, Southern Nuclear, to the extent said agent's actions may, in any way, contravene the antitrust conditions of this paragraph 2.F.

___________________

2 "Entitlement" includes, but is not limited to, power made available to an entity pursuant to an exchange agreement.

Farley - Unit 2 Renewed License No. NPF-2 Amendment No. ___

G. The facility requires relief from certain requirements of 10 CFR 50.55a(g) and exemptions from Appendices G, H and J to 10 CFR Part 50. The relief and exemptions are described in the Office of Nuclear Reactor Regulation

's Safety Evaluation Report, Supplement No. 5.

They are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest.

Therefore, the relief and exemptions are hereby granted.

With the granting of these relief and exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.

H. Southern Nuclear shall immediately notify the NRC of any accident at this facility which could result in an unplanned release of quantities of fission products in excess of allowable limits for normal operation established by the Commission.

I. Mitigation Strategy License Condition The licensee shall develop and maintain strategies for addressing large fires and explosions that include the following key areas:

(a) Fire fighting response strategy with the following elements:

1. Pre-defined coordinated fire response strategy and guidance

2. Assessment of mutual aid fire fighting assets

3. Designated staging areas for equipment and materials

4. Command and control

5. Training of response pers onnel (b) Operations to mitigate fuel damage considering the following:

1. Protection and use of personnel assets

2. Communications

3. Minimizing fire spread

4. Procedures for implementing integrated fire response strategy

5. Identification of readily-available pre

-staged equipment

6. Training on integrated fire response strategy (c) Actions to minimize release to include consideration of:

1. Water spray scrubbing

2. Dose to onsite responders J. Alabama Power Company shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

Farley - Unit 2 Renewed License No. NPF-2 Amendment No. ___

K. This renewed operating license is effective as of the date of issuance and shall expire at midnight on March 31, 2041.

FOR THE NUCLEAR REGULATORY COMMISSION J. E. Dyer, Director Office of Nuclear Reactor Regulation

Attachment:

1. Appendix A - Technical Specifications (NUREG

-0697, as revised) 2. Appendix B - Environmental Protection Plan

3. Appendix C - Additional conditions Date of Issuance: May 12, 2005

CREFS 3.7.10 Farley Units 1 and 2 3.7.10-1 Amendment No. ___ (Unit 1)

Amendment No. ___ (Unit 2) 3.7 PLANT SYSTEMS

3.7.10 CREFS LCO 3.7.10 Two CREFS trains shall be OPERABLE.

NOTE --------------------------------------------- The control room envelope (CRE) boundary may be opened intermittently under administrative control.

APPLICABILITY:

MODES 1, 2, 3, and 4, During movement of irradiated fuel assemblies, During CORE ALTERATIONS.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CREFS train inoperable for reasons other than Condition B.

A.1 Restore CREFS train to OPERABLE status.

7 days B. One or more CREFS trains inoperable due to inoperable CRE boundary. B.1 Initiate actions to implement mitigating actions. AND B.2 Verify mitigating actions ensure CRE occupant exposures to radiological, chemical, and smoke hazards will not exceed limits. AND B.3 Restore CRE boundary to OPERABLE status.

Immediately

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

90 days CREFS 3.7.10 Farley Units 1 and 2 3.7.10-2 Amendment No. ___ (Unit 1)

Amendment No. ___ (Unit 2)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and associated Completion Time of Condition A or B not met in MODE 1, 2, 3, or 4. C.1 Be in MODE 3.

AND C.2 ------------NOTE-----------

LCO 3.0.4.a is not applicable when entering MODE 4. --------------------------------

Be in MODE 4.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 12 hours D. Two CREFS trains inoperable in MODE 1, 2, 3, OR 4. D.1 Be in MODE 3. AND D.2 Be in MODE 5.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours E. Required Action and E.1 Place OPERABLE CREFS Immediately associated Completion Time of Condition A not met during movement of irradiated fuel assemblies or during CORE ALTERATIONS.

train in emergency recirculation mode.

OR E.2.1 Suspend CORE ALTERATIONS.

AND E.2.2 Suspend movement of irradiated fuel assemblies.

Immediately

CREFS 3.7.10 Farley Units 1 and 2 3.7.10-3 Amendment No. ___ (Unit 1)

Amendment No. ___ (Unit 2)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F. Required Action and associated Completion Time of Condition B not met during movement of irradiated fuel assemblies or during CORE ALTERATIONS.

OR Two CREFS trains inoperable during movement of irradiated fuel assemblies or during CORE ALTERATIONS.

OR One or more CREFS trains inoperable due to an inoperable CRE boundary during movement of irradiated fuel assemblies or during CORE ALTERATIONS.

F.1 Suspend CORE ALTERATIONS.

AND F.2 Suspend movement of irradiated fuel assemblies.

Immediately Immediately SURVEILLANCE REQUREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.1 Operate each CREFS Pressurization train with the heaters operating and each CREFS Recirculation and Filtration train for 15 minutes.

In accordance with the Surveillance Frequency Control Program SR 3.7.10.2 Perform required CREFS filter testing in accordance with the Ventilation Filter Testing Program (VFTP).

In accordance with the VFTP CREFS 3.7.10 Farley Units 1 and 2 3.7.10-4 Amendment No. ___ (Unit 1)

Amendment No. ___ (Unit 2)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.3

NOTE---------------------------------

Not required to be performed in MODES 5 and 6.

Verify each CREFS train actuates on an actual or simulated actuation signal. In accordance with the Surveillance Frequency Control Program SR 3.7.10.4 Perform required CRE unfiltered air inleakage testing in accordance with the Control Room Envelope Habitability Program. In accordance with the Control Room Envelope Habitability Program

Farley Units 1 and 2 5.5-5 Amendment No. ___ (Unit 1)

Amendment No. ___ (Unit 2) 5.5 Programs and Manuals 5.5.18 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Filtration System (CREFS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements: a. The definition of the CRE and the CRE boundary.

b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREFS, operating at the flow rate required by the VFTP, at a Frequency of 24 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 24 month assessment of the CRE boundary. e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous

Farley Units 1 and 2 5.5-6 Amendment No. ___ (Unit 1)

Amendment No. ___ (Unit 2) 5.5.18 Control Room Envelope Habitability Program (continued) chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

5.5.19 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04

-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

to NL 0388 Regulatory Guide 1.183 Conformance Tables

Joseph M. Farley Nuclear Plant - Units 1 and 2 Alternative Source Term License Amendment Request Enclosure 5 Regulatory Guide 1.183 Conformance Tables

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 1 REGULATORY GUIDE 1.183 COMPARISON Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis 1.1.1 The proposed uses of an AST and the associated proposed facility modifications and changes to procedures should be evaluated to determine whether the proposed changes are consistent with the principle that sufficient safety margins are maintained, including a margin to account for analysis uncertainties. The safety margins are products of specific values and limits contained in the technical specifications (which cannot be changed without NRC approval) and other values, such as assumed accident or transient initial conditions or assumed safety system response times. Changes, or the net effects of multiple changes, that result in a reduction in safety margins may require prior NRC approval. Once the initial AST implementation has been approved by the staff and has become part of the facility design basis, the licensee may use 10 CFR 50.59 and its supporting guidance in assessing safety margins related to subsequent facility modifications and changes to procedures.

Conforms- Adequate safety margins are maintained, as discussed in the No Significant Hazards Consideration. Future changes will be evaluated under the provisions of 10 CFR 50.59. 1.1.2 The proposed uses of an AST and the associated proposed facility modifications and changes to procedures should be evaluated to determine whether the proposed changes are consistent with the principle that adequate defense in depth is maintained to compensate for uncertainties in accident progression and analysis data. Consistency with the defense

-in-depth philosophy is maintained if system redundancy, independence, and diversity are preserved commensurate with the expected frequency, consequences of challenges to the system, and uncertainties. In all cases, compliance with the Conforms - There are no facility modifications being proposed to implement AST, and compliance with the GDCs are maintained. No reliance is placed on compensatory programmatic actions (including manual operator actions) to maintain adequate defense-in-depth. to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 2 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis General Design Criteria in Appendix A to 10 CFR Part 50 is essential. Modifications proposed for the facility generally should not create a need for compensatory programmatic activities, such as reliance on manual operator actions.

1.1.2 Proposed modifications that seek to downgrade or remove required engineered safeguards equipment should be evaluated to be sure that the modification does not invalidate assumptions made in facility PRAs and does not adversely impact the facility's severe accident management program.

Not Applicable

- There are no modifications being proposed with this License Amendment Request.

1.1.3 The design basis accident source term is a fundamental assumption upon which a significant portion of the facility design is based. Additionally, many aspects of facility operation derive from the design analyses that incorporated the earlier accident source term. Although a complete re

-assessment of all facility radiological analyses would be desirable, the NRC staff determined that recalculation of all design analyses would generally not be necessary. Regulatory Position 1.3 of this guide provides guidance on which analyses need updating as part of the AST implementation submittal and which may need updating in the future as additional modifications are performed.

Conforms - See RG Section 1.3 discussions

. 1.1.3 This approach would create two tiers of analyses, those based on the previous source term and those based on an AST. The radiological acceptance criteria would also be different with some analyses based on whole body and thyroid criteria and some based on TEDE criteria. Full implementation of the AST revises the plant licensing basis to specify the AST in place of the previous accident source term and establishes the TEDE dose as the new acceptance criteria. Selective implementation of the AST Conforms - This is a full scope AST implementation for the radiological dose consequences of the FNP Design Basis Accidents.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 3 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis also revises the plant licensing basis and may establish the TEDE dose as the new acceptance criteria. Selective implementation differs from full implementation only in the scope of the change. In either case, the facility design bases should clearly indicate that the source term assumptions and radiological criteria in these affected analyses have been superseded and that future revisions of these analyses, if any, will use the updated approved assumptions and criteria.

1.1.3 Radiological analyses generally should be based on assumptions and inputs that are consistent with corresponding data used in other design basis safety analyses, radiological and nonradiological, unless these data would result in nonconservative results or otherwise conflict with the guidance in this guide.

Conforms- This License Amendment Request includes re

-evaluation of the radiological consequences of the most severe DBAs. It relies on assumptions and inputs that do not create a conflict with, or render non

-conservative, other design basis safety analyses.

1.1.4 Although the AST provided in this guide was based on a limited spectrum of severe accidents, the particular characteristics have been tailored specifically for DBA analysis use. The AST is not representative of the wide spectrum of possible events that make up the planning basis of emergency preparedness. Therefore, the AST is insufficient by itself as a basis for requesting relief from the emergency preparedness requirements of 10 CFR 50.47 and Appendix E to 10 CFR Part 50.

Conforms - No changes are proposed in this License Amendment Request to Emergency Preparedness requirements. 1.2.1 Full implementation is a modification of the facility design basis that addresses all characteristics of the AST, that is, composition and magnitude of the radioactive material, its chemical and physical form, and the timing of its release. Full implementation revises the plant licensing basis to specify the AST in place of the previous accident source term and establishes the TEDE dose as Conforms - This License Amendment Request involves recalculation of the dose consequences of the most severe DBAs. The characteristics of the AST methods are addressed in the recalculations. The DBA LOCA has been re

-analyzed per Appendix A.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 4 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis the new acceptance criteria. This applies not only to the analyses performed in the application (which may only include a subset of the plant analyses), but also to all future design basis analyses. At a minimum for full implementations, the DBA LOCA must be re-analyzed using the guidance in Appendix A of this guide. Additional guidance on analysis is provided in Regulatory Position 1.3 of this guide. Since the AST and TEDE criteria would become part of the facility design basis, new applications of the AST would not require prior NRC approval unless stipulated by 10 CFR 50.59, "Changes, Tests, and Experiments," or unless the new application involved a change to a technical specification. However, a change from an approved AST to a different AST that is not approved for use at that facility would require a license amendment under 10 CFR 50.67.

1.2.2 Selective implementation is a modification of the facility design basis that (1) is based on one or more of the characteristics of the AST or (2) entails re

-evaluation of a limited subset of the design basis radiological analyses. The NRC staff will allow licensees flexibility in technically justified selective implementations provided a clear, logical, and consistent design basis is maintained. An example of an application of selective implementation would be one in which a licensee desires to use the release timing insights of the AST to increase the required closure time for a containment isolation valve by a small amount.

Another example would be a request to remove the charcoal filter media from the spent fuel building ventilation exhaust. For the latter, the licensee may only need to re

-analyze DBAs that credited the iodine removal by the charcoal media. Additional Not Applicable

- This License Amendment Request is for full scope AST implementation for the radiological dose consequences of the major FNP DBA.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 5 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis analysis guidance is provided in Regulatory Position 1.3 of this guide. NRC approval for the AST (and the TEDE dose criterion) will be limited to the particular selective implementation proposed by the licensee. The licensee would be able to make subsequent modifications to the facility and changes to procedures based on the selected AST characteristics incorporated into the design basis under the provisions of 10 CFR 50.59. However, use of other characteristics of an AST or use of TEDE criteria that are not part of the approved design basis, and changes to previously approved AST characteristics, would require prior staff approval under 10 CFR 50.67. As an example, a licensee with an implementation involving only timing, such as relaxed closure time on isolation valves, could not use 10 CFR 50.59 as a mechanism to implement a modification involving a reanalysis of the DBA LOCA. However, this licensee could extend use of the timing characteristic to adjust the closure time on isolation valves not included in the original approval.

1.3.1 There are several regulatory requirements for which compliance is demonstrated, in part, by the evaluation of the radiological consequences of design basis accidents. These requirements include, but are not limited to, the following.

Environmental Qualification of Equipment (10 CFR 50.49)

Control Room Habitability (GDC 19 of Appendix A to 10 CFR Part 50)

Emergency Response Facility Habitability (Paragraph IV.E.8 of Appendix E to 10 CFR Part 50)

Alternative Source Term (10 CFR 50.67)

Conforms- This full scope AST License Amendment Request is salient to: a) Control Room Habitability (GDC 19 and NUREG

-0737 Item III.D.3.4), b) AST (10 CFR 50.67), and c) Facility Siting (10 CFR 100.11). Control Room Habitability and compliance with the Alternative Source Term requirements are the principal subjects of this submittal and are discussed in Sections 3 and 4 of this License Amendment Request.

Regarding Emergency Response Facility Habitability, to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 6 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis Environmental Reports (10 CFR Part 51)

Facility Siting (10 CFR 100.11) 5 There may be additional applications of the accident source term identified in the technical specification bases and in various licensee commitments. These include, but are not limited to, the following from Reference 2, NUREG

-0737. Post-Accident Access Shielding (NUREG

-0737, II.B.2)

Post-Accident Sampling Capability (NUREG

-0737, II.B.3)

Accident Monitoring Instrumentation (NUREG

-0737, II.F.1) Leakage Control (NUREG

-0737, III.D.1.1)

Emergency Response Facilities (NUREG

-0737, III.A.1.2)

Control Room Habitability (NUREG

-0737, III.D.3.4)

FNP will continue to meet the NUREG

-0654 Planning Standard for Emergency Facilities and Equipment as described in the FNP Emergency Plan.

As stated in Footnote 5 of this RG, the dose guidelines of 10 CFR 100.11 are superseded by 10 CFR 50.67 for licensees that have implemented an AST. 1.3.2 Any implementation of an AST, full or selective, and any associated facility modification should be supported by evaluations of all significant radiological and nonradiological impacts of the proposed actions. This evaluation should consider the impact of the proposed changes on the facility's compliance with the regulations and commitments listed above as well as any other facility

-specific requirements. These impacts may be due to (1) the associated facility modifications or (2) the differences in the AST characteristics. The scope and extent of the re

-evaluation will necessarily be a function of the specific proposed facility modification 6 and whether a full or selective implementation is being pursued. The NRC staff does not expect a complete recalculation of all facility radiological analyses, but does expect licensees to evaluate all impacts of the proposed Conforms- The License Amendment Request for this full scope application of the AST evaluated the impact of the proposed change against the Current Licensing Basis, mitigating system design basis requirements, and Technical Specifications.

No facility modifications are proposed as part of this License Amendment Request and compliance with regulations and commitments are maintained.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 7 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis changes and to update the affected analyses and the design bases appropriately. An analysis is considered to be affected if the proposed modification changes one or more assumptions or inputs used in that analysis such that the results, or the conclusions drawn on those results, are no longer valid. Generic analyses, such as those performed by owner groups or vendor topical reports, may be used provided the licensee justifies the applicability of the generic conclusions to the specific facility and implementation. Sensitivity analyses, discussed below, may also be an option. If affected design basis analyses are to be re

-

calculated, all affected assumptions and inputs should be updated and all selected characteristics of the AST and the TEDE criteria should be addressed. The license amendment request should describe the licensee's re

-analysis effort and provide statements regarding the acceptability of the proposed implementation, including modifications, against each of the applicable analysis requirements and commitments identified in Regulatory Position 1.3.1 of this guide.

1.3.2 The NRC staff has performed an evaluation of the impact of the AST on three representative operating reactors (Ref. 14). This evaluation determined that radiological analysis results based on the TID-14844 source term assumptions (Ref. 1) and the whole body and thyroid methodology generally bound the results from analyses based on the AST and TEDE methodology. Licensees may use the applicable conclusions of this evaluation in addressing the impact of the AST on design basis radiological analyses. However, this does not exempt the licensee from evaluating the remaining radiological and nonradiological impacts Conforms- There are no plant modifications that are planned to implement the AST analyses. The radiological and nonradiological impacts of full scope implementation of the AST have been considered and discussed in the License Amendment Request, as applicable.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 8 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis of the AST implementation and the impacts of the associated plant modifications. For example, a selective implementation based on the timing insights of the AST may change the required isolation time for the containment purge dampers from 2.5 seconds to 5.0 seconds. This application might be acceptable without dose calculations. However, evaluations may need to be performed regarding the ability of the damper to close against increased containment pressure or the ability of ductwork downstream of the dampers to withstand increased stresses.

1.3.2 For full implementation, a complete DBA LOCA analysis as described in Appendix A of this guide should be performed, as a minimum. Other design basis analyses are updated in accordance with the guidance in this section.

Conforms - T he DBA LOCA analysis is provided in this License Amendment Request which is consistent with Appendix A.

1.3.2 A selective implementation of an AST and any associated facility modification based on the AST should evaluate all the radiological and nonradiological impacts of the proposed actions as they apply to the particular implementation. Design basis analyses are updated in accordance with the guidance in this section. There is no minimum requirement that a DBA LOCA analysis be performed. The analyses performed need to address all impacts of the proposed modification, the selected characteristics of the AST, and if dose calculations are performed, the TEDE criteria. For selective implementations based on the timing characteristic of the AST, e.g., change in the closure timing of a containment isolation valve, re

-analysis of radiological calculations may not be necessary if the modified elapsed time remains a fraction (e.g., 25%) of the time between accident initiation and the onset of the gap release phase. Longer Not Applicable

- This License Amendment Request is a full scope AST implementation that evaluates the dose consequences of the most severe FNP DBAs. to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 9 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis time delays may be considered on an individual basis. For longer time delays, evaluation of the radiological consequences and other impacts of the delay, such as blockage by debris in sump water, may be necessary. If affected design basis analyses are to be re-calculated, all affected assumptions and inputs should be updated and all selected characteristics of the AST and the TEDE criteria should be addressed.

1.3.3 It may be possible to demonstrate by sensitivity or scoping evaluations that existing analyses have sufficient margin and need not be recalculated. As used in this guide, a sensitivity analysis is an evaluation that considers how the overall results vary as an input parameter (in this case, AST characteristics) is varied. A scoping analysis is a brief evaluation that uses conservative, simple methods to show that the results of the analysis bound those obtainable from a more complete treatment. Sensitivity analyses are particularly applicable to suites of calculations that address diverse components or plant areas but are otherwise largely based on generic assumptions and inputs. Such cases might include postaccident vital area access dose calculations, shielding calculations, and equipment environmental qualification (integrated dose). It may be possible to identify a bounding case, re

-analyze that case, and use the results to draw conclusions regarding the remainder of the analyses. It may also be possible to show that for some analyses the whole body and thyroid doses determined with the previous source term would bound the TEDE obtained using the AST. Where present, arbitrary "designer margins" may be adequate to bound any impact of the AST and TEDE criteria. If sensitivity or scoping Not Applicable

- The FNP AST analysis does not rely on sensitivity or scoping analyses.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 10 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis analyses are used, the license amendment request should include a discussion of the analyses performed and the conclusions drawn. Scoping or sensitivity analyses should not constitute a significant part of the evaluations for the design basis exclusion area boundary (EAB), low population zone (LPZ), or control room dose.

1.3.4 Full implementation of the AST replaces the previous accident source term with the approved AST and the TEDE criteria for all design basis radiological analyses. The implementation may have been supported in part by sensitivity or scoping analyses that concluded many of the design basis radiological analyses would remain bounding for the AST and the TEDE criteria and would not require updating. After the implementation is complete, there may be a subsequent need (e.g., a planned facility modification) to revise these analyses or to perform new analyses. For these recalculations, the NRC staff expects that all characteristics of the AST and the TEDE criteria incorporated into the design basis will be addressed in all affected analyses on an individual as

-needed basis. Re-evaluation using the previously approved source term may not be appropriate. Since the AST and the TEDE criteria are part of the approved design basis for the facility, use of the AST and TEDE criteria in new applications at the facility do not constitute a change in analysis methodology that would require NRC approval.

7 Not Applicable

- The FNP AST design basis radiological analyses do not rely on sensitivity or scoping analyses.

1.3.4 This guidance is also applicable to selective implementations to the extent that the affected analyses are within the scope of the approved implementation as described in the facility design basis. In these cases, the characteristics of the AST and TEDE criteria Not Applicable

- This is a full scope License Amendment Request that evaluates the dose consequences of the most severe FNP DBAs. to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 11 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis identified in the facility design basis need to be considered in updating the analyses. Use of other characteristics of the AST or TEDE criteria that are not part of the approved design basis, and changes to previously approved AST characteristics, requires prior NRC staff approval under 10 CFR 50.67.

1.3.5 Current environmental qualification (EQ) analyses may be impacted by a proposed plant modification associated with the AST implementation. The EQ analyses that have assumptions or inputs affected by the plant modification should be updated to address these impacts. The NRC staff is assessing the effect of increased cesium releases on EQ doses to determine whether licensee action is warranted. Until such time as this generic issue is resolved, licensees may use either the AST or the TID14844 assumptions for performing the required EQ analyses. However, no plant modifications are required to address the impact of the difference in source term characteristics (i.e., AST vs TID14844) on EQ doses pending the outcome of the evaluation of the generic issue. The EQ dose estimates should be calculated using the design basis survivability period.

Conforms - The FNP AST License Amendment Request is not proposing to modify the equipment qualification design basis to adopt AST. The FNP EQ analysis will continue to be based on TID

-14844 assumptions.

1.4 The use of an AST changes only the regulatory assumptions regarding the analytical treatment of the design basis accidents.

The AST has no direct effect on the probability of the accident. Use of an AST alone cannot increase the core damage frequency (CDF) or the large early release frequency (LERF). However, facility modifications made possible by the AST could have an impact on risk. If the proposed implementation of the AST involves changes to the facility design that would invalidate assumptions made in the facility's PRA, the impact on the existing Not Applicable

- No facility modifications are proposed or planned as implementation actions of th e FHA AST analysis.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 12 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis PRAs should be evaluated.

1.4 Consideration should be given to the risk impact of proposed implementations that seek to remove or downgrade the performance of previously required engineered safeguards equipment on the basis of the reduced postulated doses. The NRC staff may request risk information if there is a reason to question adequate protection of public health and safety.

Not Applicable

- The FNP AST License Amendment Request is not seeking to remove or downgrade the performance of previously required engineered safeguards equipment on the basis of the reduced postulated doses. 1.4 The licensee may elect to use risk insights in support of proposed changes to the design basis that are not addressed in currently approved NRC staff positions. For guidance, refer to Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (Ref. 15).

Not Applicable

- The FNP AST License Amendment Request is not utilizing risk insights as a basis for any proposed changes. 1.5 According to 10 CFR 50.90, an application for an amendment must fully describe the changes desired and should follow, as far as applicable, the form prescribed for original applications.

Regulatory Guide 1.70, "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants (LWR Edition)" (Ref 16), provides additional guidance. The NRC staff's finding that the amendment may be approved must be based on the licensee's analyses, since it is these analyses that will become part of the design basis of the facility. The amendment request should describe the licensee's analyses of the radiological and nonradiological impacts of the proposed modification in sufficient detail to support review by the NRC staff. The staff recommends that licensees submit affected FSAR pages annotated with changes that reflect the revised analyses or submit the actual Conforms- The License Amendment Request is formatted in accordance with accepted NRC/industry guidance. The request describes the radiological and nonradiological impacts of the FNP AST analysis. Consistent with previous precedent, affected USAR pages are not included in the analyses. However, a detailed summary of the AST dose calculations are included. Approval of this License Amendment Request will result in the necessary revisions to the USAR, with revised USAR pages submitted pursuant to 10 CFR 50.71(e).

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 13 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis calculation documentation.

1.5 If the licensee has used a current approved version of an NRC

-sponsored computer code, the NRC staff review can be made more efficient if the licensee identifies the code used and submits the inputs that the licensee used in the calculations made with that code. In many cases, this will reduce the need for NRC staff confirmatory analyses. This recommendation does not constitute a requirement that the licensee use NRC

-sponsored computer codes. The LOCA dose calculation was performed using RADTRAD 3.10. The FHA dose calculations uses RADTRAD Version 3.03.

The MSLB, CRE, and Locked Rotor dose calculations were performed using the Bechtel standard computer program LocaDose, Version 7.1.

The SGTR dose calculation was performed using LocaDose, Version 7.11.

LocaDose is designed to calculate radioactive isotope activities within regions, radioactive releases from regions, doses and dose rates within regions for humans and equipment, and inhalation and immersion doses and dose rates at offsite locations to plant personnel and the general public.

1.6 Requirements for updating the facility's final safety analysis report (FSAR) are in 10 CFR 50.71, "Maintenance of Records, Making of Reports." The regulations in 10 CFR 50.71(e) require that the FSAR be updated to include all changes made in the facility or procedures described in the FSAR and all safety evaluations performed by the licensee in support of requests for license amendments or in support of conclusions that changes did not involve unreviewed safety questions. The analyses required by 10 CFR 50.67 are subject to this requirement. The affected radiological analysis descriptions in the FSAR should be updated Conforms- Approval of this License Amendment Request will result in the necessary revisions to the USAR, with revised USAR pages submitted pursuant to 10 CFR 50.71(e).

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 14 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis to reflect the replacement of the design basis source term by the AST. The analysis descriptions should contain sufficient detail to identify the methodologies used, significant assumptions and inputs, and numeric results. Regulatory Guide 1.70 (Ref. 16) provides additional guidance. The descriptions of superseded analyses should be removed from the FSAR in the interest of maintaining a clear design basis.

2.1 The AST must be based on major accidents, hypothesized for the purposes of design analyses or consideration of possible accidental events, that could result in hazards not exceeded by those from other accidents considered credible. The AST must address events that involve a substantial meltdown of the core with the subsequent release of appreciable quantities of fission products. Conforms- This License Amendment Request applies the AST methods when evaluating the dose consequences of the most severe DBAs applicable to FNP. 2.2 The AST must be expressed in terms of times and rates of appearance of radioactive fission products released into containment, the types and quantities of the radioactive species released, and the chemical forms of iodine released.

Conforms - For the DBAs that release to Containment (LOCA, FHA, and Control Rod Ejection), the AST is expressed in terms of times and rates of release of radioactive fission products, the types and quantities of the radioactive species released, and the chemical forms of iodine released.

2.3 The AST must not be based upon a single accident scenario but instead must represent a spectrum of credible severe accident events. Risk insights may be used, not to select a single risk

-significant accident, but rather to establish the range of events to be considered. Relevant insights from applicable severe accident research on the phenomenology of fission product release and transport behavior may be considered.

Conforms- This License Amendment Request considers a number of release scenarios, as applicable, for the DBAs being revised for use of AST. The most limiting of these releases are analyzed for radiological consequences.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 15 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis 2.4 The AST must have a defensible technical basis supported by sufficient experimental and empirical data, be verified and validated, and be documented in a scrutable form that facilitates public review and discourse.

Conforms- The DBA AST dose calculations ha ve been developed based on NUREG

-1465 and this Regulatory Guide. The calculation s, which utilizes RADTRAD and Bechtel LocaDose were developed in accordance with 10 CFR 50 Appendix B, Criterion III.

2.5 The AST must be peer

-reviewed by appropriately qualified subject matter experts. The peer

-review comments and their resolution should be part of the documentation supporting the AST. Conforms- The FNP AST dose calculations have been developed by industry experts and reviewed and accepted by SNC Engineering. The calculation was developed in accordance with 10 CFR 50 Appendix B program, Criterion III.

3.1 The inventory of fission products in the reactor core and available for release to the containment should be based on the maximum full power operation of the core with, as a minimum, current licensed values for fuel enrichment, fuel burnup, and an assumed core power equal to the current licensed rated thermal power times the ECCS evaluation uncertainty.8 The period of irradiation should be of sufficient duration to allow the activity of dose

-significant radionuclides to reach equilibrium or to reach maximum values.9 The core inventory should be determined using an appropriate isotope generation and depletion computer code such as ORIGEN 2 (Ref. 17) or ORIGEN

-ARP (Ref. 18). Core inventory factors (Ci/MWt) provided in TID14844 and used in some analysis computer codes were derived for low burnup, low enrichment fuel and should not be used with higher burnup and higher enrichment fuels.

Conforms - The FNP DBAs that release to the Containment are the LOCA, FHA, and Control Rod Ejection. to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 16 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis 3.1 For the DBA LOCA, all fuel assemblies in the core are assumed to be affected and the core average inventory should be used. 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, radial peaking factors from the facility's core operating limits report (COLR) or technical specifications should be applied in determining the inventory of the damaged rods.

With the exception of the Fuel Handling Accident (FHA), the analyses of events which involve fuel damage assume that the entire core is affected with a source term based upon full power, core average conditions. The FHA source term is derived from the core source term, the number of damaged fuel rods, and a conservative assembly peaking factor, which exceeds the maximum fuel rod peaking factor specified in the COLR

. 3.1 No adjustment to the fission product inventory should be made for events postulated to occur during power operations at less than full rated power or those postulated to occur at the beginning of core life. For events postulated to occur while the facility is shutdown, e.g., a fuel handling accident, radioactive decay from the time of shutdown may be modeled.

The analysis of the FHA considers radioactive decay between the time of core shutdown and the beginning of fuel movement. to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 17 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis 3.2 The core inventory release fractions, by radionuclide groups, for the gap release and early in-vessel damage phases for DBA LOCAs are listed in Table 1 for BWRs and Table 2 for PWRs. These fractions are applied to the equilibrium core inventory described in Regulatory Position 3.1.

Table 2 PWR Core Inventory Fraction Released Into Containment Group Gap Release Phase Early In-vessel Phase Total Noble Gases 0.05 0.95 1.0 Halogens 0.05 0.35 0.4 Alkali Metals 0.05 0.25 0.3 Tellurium Metals 0.00 0.05 0.05 Ba, Sr 0.00 0.02 0.02 Noble Metals 0.00 0.0025 0.0025 Cerium Group 0.00 0.0005 0.0005 Lanthanides 0.00 0.0002 0.0002 Conforms - The LOCA AST calculation models Table 2 in the release fraction and timing file.

3.2 For non-LOCA events, the fractions of the core inventory assumed to be in the gap for the various radionuclides are given in Table 3. The release fractions from Table 3 are used in conjunction with the fission product inventory calculated with the maximum core radial peaking factor.

Conforms - The FHA, CRE, and Locked Rotor accidents result in fuel damage, so the non

-LOCA gap fractions of Table 3 are used. While the SGTR and MSLB accidents conservatively assume a pre

-existing 1% failed fuel source term, this is not the result of damage caused by the accident, and so the to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 18 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis Table 3.11 Non-LOCA Fraction of Fission Product Inventory in Gap Table 3 Group Fraction I-131 0.08 Kr-85 0.10 Other Noble Gases 0.05 Other Halogens 0.05 Alkali Metals 0.12 non-LOCA gap fractions of Ta ble 3 are not included for these events.

3.3 Table 4 tabulates the onset and duration of each sequential release phase for DBA LOCAs at PWRs and BWRs. The specified onset is the time following the initiation of the accident (i.e., time = 0). The early in

-vessel phase immediately follows the gap release phase. The activity released from the core during each release phase should be modeled as increasing in a linear fashion over the duration of the phase.12 For non

-LOCA DBAs in which fuel damage is projected, the release from the fuel gap and the fuel pellet should be assumed to occur instantaneously with the onset of the projected damage.

Table 4 LOCA Release Phases (PWR)

Phase Onset Duration Gap Release 30 sec 0.5 hr Early In-vessel 0.5 hr 1.3 hr Conforms - The LOCA AST calculation models Table 4 in the release fraction and timing file.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 19 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis 3.3 For facilities licensed with leak

-before-break methodology, the onset of the gap release phase may be assumed to be 10 minutes. A licensee may propose an alternative time for the onset of the gap release phase, based on facility

-specific calculations using suitable analysis codes or on an accepted topical report shown to be applicable to the specific facility. In the absence of approved alternatives, the gap release phase onsets in Table 4 should be used.

Conforms - The LOCA AST calculation models Table 4 in the release fraction and timing file.

3.4 Elements listed in Table 5 in each radionuclide group that should be considered in design basis analyses.

Table 5 Radionuclide Groups Group Elements Noble Gases Xe, Kr Halogens I, Br Alkali Metals Cs, Rb Tellurium Group Te, Sb, Se, Ba,Sr Noble Metals Ru, Rh, Pd, Mo, Tc, Co Lanthenides La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am Cerium Ce, Pu, Np Conforms The source term in the design basis analysis represents the most dose significant isotopes from the elements listed in Table 5 of Regulatory Guide 1.183.

3.5 Of the radioiodine released from the reactor coolant system (RCS) to the containment in a postulated accident, 95 percent of the iodine released should be assumed to be cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide.

This includes releases from the gap and the fuel pellets. With the exception of elemental and organic iodine and noble gases, Conforms - The chemical composition of the iodine released from the RCS to containment in the LOCA event is 95% aerosol, 4.85% elemental, and 0.15% organic. All non iodine and non

-noble gas fission products are assumed to be in particulate form. The chemical composition of iodines in the non-LOCA to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 20 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis fission products should be assumed to be in particulate form. The same chemical form is assumed in releases from fuel pins in FHAs and from releases from the fuel pins through the RCS in DBAs other than FHAs or LOCAs. However, the transport of these iodine species following release from the fuel may affect these assumed fractions. The accident

-specific appendices to this regulatory guide provide additional details.

events are based upon the guidance in the respective appendices of Reg. Guide 1.183.

3.6 The amount of fuel damage caused by non

-LOCA design basis events should be analyzed to determine, for the case resulting in the highest radioactivity release, the fraction of the fuel that reaches or exceeds the initiation temperature of fuel melt and the fraction of fuel elements for which the fuel clad is breached. Although the NRC staff has traditionally relied upon the departure from nucleate boiling ratio (DNBR) as a fuel damage criterion, licensees may propose other methods to the NRC staff, such as those based upon enthalpy deposition, for estimating fuel damage for the purpose of establishing radioactivity releases.

Conforms - The amount of fuel damage in the Locked Rotor event is based upon the fraction of the core which experiences DNB as reported in the Updated Final Safety Analysis Report (FSAR). The fraction of the fuel rods assumed to melt in the CRE event is conservatively based upon the portion of the fuel centerline that is calculated to exceed the melting temperature as documented in the FSAR. 4.1.1 The dose calculations should determine the TEDE. TEDE is the sum of the committed effective dose equivalent (CEDE) from inhalation and the deep dose equivalent (DDE) from external exposure. The calculation of these two components of the TEDE should consider all radionuclides, including progeny from the decay of parent radionuclides, that are significant with regard to dose consequences and the released radioactivity.

13 Conforms - The AST dose consequences are calculated in TEDE.

4.1.2 The exposure

-to-CEDE factors for inhalation of radioactive material should be derived from the data provided in ICRP Publication 30, "Limits for Intakes of Radionuclides by Workers" (Ref. 19). Table 2.1 of Federal Guidance Report 11, "Limiting Conforms - Dose Conversion Factors for inhalation in this analysis are taken from Table 2.1 of Federal Guidance Report 11.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 21 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion" (Ref. 20), provides tables of conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the CEDE.

4.1.3 For the first 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the breathing rate of persons offsite should be assumed to be 3.5 x 10

-4 cubic meters per second. From 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the accident, the breathing rate should be assumed to be 1.8 x 10

-4 cubic meters per second. After that and until the end of the accident, the rate should be assumed to be 2.3 x 10-4 cubic meters per second.

Conforms - Offsite breathing rates used in the analysis are consistent with the values specified in Section 4.1.3 of Reg. Guide 1.

183. 4.1.4 The DDE should be calculated assuming submergence in semi

-infinite cloud assumptions with appropriate credit for attenuation by body tissue. The DDE is nominally equivalent to the effective dose equivalent (EDE) from external exposure if the whole body is irradiated uniformly. Since this is a reasonable assumption for submergence exposure situations, EDE may be used in lieu of DDE in determining the contribution of external dose to the TEDE. Table III.1 of Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil" (Ref. 21), provides external EDE conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the EDE.

Conforms - Dose Conversion Factors for air submergence are taken from the Table 111. 1 of Federal Guidance Report 12.

4.1.5 The TEDE should be determined for the most limiting person at the EAB. The maximum EAB TEDE for any two

-hour period following the start of the radioactivity release should be determined and used in determining compliance with the dose criteria in 10 CFR 50.67. The maximum two

-hour TEDE should Conforms - The TEDE was determined for the most limiting person at the EAB. The maximum two

-hour TEDE was determined by calculating the postulated dose for a series of small time increments and performing a 'sliding' sum over increments for to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 22 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis be determined by calculating the postulated dose for a series of small time increments and performing a "sliding" sum over the increments for successive two

-hour periods. The maximum TEDE obtained is submitted. The time increments should appropriately reflect the progression of the accident to capture the peak dose interval between the start of the event and the end of radioactivity release (see also Table 6).

successive two

-hour periods.

4.1.6 TEDE should be determined for the most limiting receptor at the outer boundary of the low population zone (LPZ) and should be used in determining compliance with the dose criteria in 10 CFR 50.67. Conforms - The TEDE is determined for the most limiting person at the LPZ.

4.1.7 No correction should be made for depletion of the effluent plume by deposition on the ground.

Conforms - No correction is made for deposition of the effluent plume by deposition on the ground. 4.2.1 The TEDE analysis should consider all sources of radiation that will cause exposure to control room personnel. The applicable sources will vary from facility to facility, but typically will include:

Contamination of the control room atmosphere by the intake or infiltration of the radioactive material contained in the radioactive plume released from the facility, Contamination of the control room atmosphere by the intake or infiltration of airborne radioactive material from areas and structures adjacent to the control room envelope, Radiation shine from the external radioactive plume released from the facility, Radiation shine from radioactive material in the reactor containment, Conforms - The analyses consider the applicable sources of contamination to the control room atmosphere for each event.

With respect to external and containment shine sources and their impact on control room doses, the physical design of the control room envelop and the surrounding auxiliary building provide more than 18" of concrete shielding between the operators and shine sources in all directions around the control room. The Control Room Emergency Filtration System filters are located outside of and above the control room envelope. The control room ceiling is approximately to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 23 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis Radiation shine from radioactive material in systems and components inside or external to the control room envelope, e.g., radioactive material buildup in recirculation filters. 24" thick. Accordingly, shielding from the walls and the filter unit casings prevents an appreciable dose to the operators during the accident.

The control room is surrounded by the Auxiliary Building (and s o does not abut the containment

), and is shielded from containment by more than 2 feet of concrete in all directions. The containment walls are 3'9" thick as well.

Accordingly, the control room is adequately shielded from containment shine, as well as shine from containment leakage sources.

With respect to shine from the release plume, the exterior Auxiliary Building surrounds the control room and the exterior concrete walls are approximately 21" thick. The floors, walls, and ceilings of the control room add to the concrete shielding from the plume. Therefore, shine from the release plume to the control room occupants will not be significant.

For the Fuel Handling Accident scenario where the Personnel Airlock is open, the Auxiliary Building area around the control room could become contaminated.

A small section of the control room envelop wall is only 1 foot thick inside the Auxiliary Building (between the control room and an interior hallway).

Doses to the control room operators due to shine from the contaminated area through the 1 foot thick to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 24 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis wall are included in the Fuel Handling Accident evaluation of control room doses and were found to be not significant

. 4.2.2 The radioactive material releases and radiation levels used in the control room dose analysis should be determined using the same source term, transport, and release assumptions used for determining the EAB and the LPZ TEDE values, unless these assumptions would result in non

-conservative results for the control room.

Conforms - The FNC AST dose calculations use the same source term, transport, and release assumptions for Control Room, EAB, and EPZ dose values. 4.2.3 The models used to transport radioactive material into and through the control room, 15 and the shielding models used to determine radiation dose rates from external sources, should be structured to provide suitably conservative estimates of the exposure to control room personnel.

Conforms - The models used to transport radioactive material into and through the control room have been structured to provide suitably conservative estimates of the exposure to control room personnel.

No shielding models have been used in this application.

4.2.4 Credit for engineered safety features that mitigate airborne radioactive material within the control room may be assumed.

Such features may include control room isolation or pressurization, or intake or recirculation filtration. Refer to Section 6.5.1, "ESF Atmospheric Cleanup System," of the SRP (Ref. 3) and Regulatory Guide 1.52, "Design, Testing, and Maintenance Criteria for Postaccident Engineered

-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light

-Water-Cooled Nuclear Power Plants" (Ref. 25), for guidance. The control room design is often optimized for the DBA LOCA and the protection afforded for other accident sequences may not be as advantageous. In most designs, control room isolation is actuated by engineered safeguards feature (ESF) signals or radiation monitors (RMs). In some cases, the ESF signal is effective only Conforms - For the AST DBAs covered under this License Amendment Request, credit is taken for control room isolation and reconfiguring into the emergency ventilation mode upon accident initiation by a high radiation or Safety Injection signal.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 25 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis for selected accidents, placing reliance on the RMs for the remaining accidents. Several aspects of RMs can delay the control room isolation, including the delay for activity to build up to concentrations equivalent to the alarm setpoint and the effects of different radionuclide accident isotopic mixes on monitor response. 4.2.5 Credit should generally not be taken for the use of personal protective equipment or prophylactic drugs. Deviations may be considered on a case

-by-case basis.

Conforms- No credit is taken for the use of personal protective equipment or prophylactic drugs.

4.2.6 The dose receptor for these analyses is the hypothetical maximum exposed individual who is present in the control room for 100% of the time during the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the event, 60% of the time between 1 and 4 days, and 40% of the time from 4 days to 30 days.16 For the duration of the event, the breathing rate of this individual should be assumed to be 3.5 x 10

-4 cubic meters per second.

Conforms - Control room occupancy and breathing rates are consistent with this regulatory position.

4.2.7 Control room doses should be calculated using dose conversion factors identified in Regulatory Position 4.1 above for use in offsite dose analyses. The DDE from photons may be corrected for the difference between finite cloud geometry in the control room and the semi

-infinite cloud assumption used in calculating the dose conversion factors. The following expression may be used to correct the semi

-infinite cloud dose, DDE, to a finite cloud dose, DDEfinite , where the control room is modeled as a hemisphere that has a volume, V, in cubic feet, equivalent to that of the control room (Ref. 22).

Conforms - Control room doses are calculated using dose conversion factors identified in Position 4.1 above.

Equation 1 from Reg. Guide 1.183 is used for finite cloud correction when calculating the DDE immersion doses due to airborne activity inside the control room in the Fuel Handling Accident.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 26 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis DDEfinite= 1173338.0 V DDE 4.3 The guidance provided in Regulatory Positions 4.1 and 4.2 should be used, as applicable, in re

-assessing the radiological analyses identified in Regulatory Position 1.3.1, such as those in NUREG-0737 (Ref. 2). Design envelope source terms provided in NUREG-0737 should be updated for consistency with the AST. In general, radiation exposures to plant personnel identified in Regulatory Position 1.3.1 should be expressed in terms of TEDE. Integrated radiation exposure of plant equipment should be determined using the guidance of Appendix I of this guide.

Not Applicable

- This full scope AST implementation LAR is for the radiological consequences of major FNP DBAs. 4.4 The radiological criteria for the EAB, the outer boundary of the LPZ, and for the control room are in 10 CFR 50.67. These criteria are stated for evaluating reactor accidents of exceedingly low probability of occurrence and low risk of public exposure to radiation, e.g., a large

-break LOCA. The control room criterion applies to all accidents. For events with a higher probability of occurrence, postulated EAB and LPZ doses should not exceed the criteria tabulated in Table 6.

Table 6 17 Accident Dose Criteria Accident or Case EAB and LPZ Dose Criteria Analysis Release Duration LOCA 25 rem TEDE 30 days for containment and ECCS leakage The EAB and LPZ acceptance criteria from Table 6 of RG 1.183 are applied. The control room acceptance

of 5 rem TEDE is taken from 10 CFR 50.67(b)(2)(iii).

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 27 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis PWR Steam Generator Tube Rupture Affected SG: time to isolate; Unaffected SG(s): until cold shutdown is established Fuel Damage or Pre-incident Spike 25 rem TEDE Coincident Iodine Spike 2.5 rem TEDE PWR Main Steam Line Break Until cold shutdown is established Fuel Damage or Pre-incident Spike 25 rem TEDE Coincident Iodine Spike 2.5 rem TEDE PWR Locked Rotor Accident 2.5 rem TEDE Until cold shutdown is established PWR Rod Ejection Accident 6.3 rem TEDE 30 days for containment pathway; until cold shutdown is established for secondary pathway Fuel Handling Accident 6.3 rem TEDE 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 28 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis The column labeled "Analysis Release Duration" is a summary of the assumed radioactivity release durations identified in the individual appendices to this guide. Refer to these appendices for complete descriptions of the release pathways and durations.

4.4 The acceptance criteria for the various NUREG

-0737 (Ref. 2) items generally reference General Design Criteria 19 (GDC 19) from Appendix A to 10 CFR Part 50 or specify criteria derived from GDC 19. These criteria are generally specified in terms of whole body dose, or its equivalent to any body organ. For facilities applying for, or having received, approval for the use of an AST, the applicable criteria should be updated for consistency with the TEDE criterion in 10 CFR 50.67(b)(2)(iii).

Conforms - The EAB and LPZ acceptance criteria from Table 6 of RG 1.183 are applied. The control room occupant acceptance criteria of 5 rem TEDE is taken from 10 CFR 50.67(b)(2)(iii).

5.1.1 The evaluations required by 10 CFR 50.67 are re

-analyses of the design basis safety analyses and evaluations required by 10 CFR 50.34; they are considered to be a significant input to the evaluations required by 10 CFR 50.92 or 10 CFR 50.59. These analyses should be prepared, reviewed, and maintained in accordance with quality assurance programs that comply with Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50.

Conforms- The FNP AST dose calculations were prepared and accepted by SNC under a 10 CFR 50 Appendix B Quality Assurance program.

5.1.1 These design basis analyses were structured to provide a conservative set of assumptions to test the performance of one or more aspects of the facility design. Many physical processes and phenomena are represented by conservative, bounding assumptions rather than being modeled directly. The staff has selected assumptions and models that provide an appropriate and prudent safety margin against unpredicted events in the Not Applicable

- This License Amendment Request is not proposing deviations to conformance with this Regulatory Guide.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 29 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis course of an accident and compensate for large uncertainties in facility parameters, accident progression, radioactive material transport, and atmospheric dispersion. Licensees should exercise caution in proposing deviations based upon data from a specific accident sequence since the DBAs were never intended to represent any specific accident sequence

-- the proposed deviation may not be conservative for other accident sequences.

5.1.2 Credit may be taken for accident mitigation features that are classified as safety

-related, are required to be operable by technical specifications, are powered by emergency power sources, and are either automatically actuated or, in limited cases, have actuation requirements explicitly addressed in emergency operating procedures. The single active component failure that results in the most limiting radiological consequences should be assumed. Assumptions regarding the occurrence and timing of a loss of offsite power should be selected with the objective of maximizing the postulated radiological consequences.

Conforms - Only safety-related Engineered Safety Features are credited in the analysis with an assumed single active failure that results in the greatest impact on the radiological consequences. A loss of offsite power is assumed concurrent with the start of each event as that maximizes the dose impact. 5.1.3 The numeric values that are chosen as inputs to the analyses required by 10 CFR 50.67 should be selected with the objective of determining a conservative postulated dose. In some instances, a particular parameter may be conservative in one portion of an analysis but be nonconservative in another portion of the same analysis. For example, assuming minimum containment system spray flow is usually conservative for estimating iodine scrubbing, but in many cases may be nonconservative when determining sump pH. Sensitivity analyses may be needed to determine the appropriate value to use. As a Conforms - Numerical values are selected and biased for each application in a conservative direction with the objective of maximizing the dose consequences. Numerical values for parameters which are controlled by Technical Specifications are either used as direct inputs in the analysis, or more conservative values may be used to enhance safety margin. to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 30 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis conservative alternative, the limiting value applicable to each portion of the analysis may be used in the evaluation of that portion. A single value may not be applicable for a parameter for the duration of the event, particularly for parameters affected by changes in density. For parameters addressed by technical specifications, the value used in the analysis should be that specified in the technical specifications.

18 If a range of values or a tolerance band is specified, the value that would result in a conservative postulated dose should be used. If the parameter is based on the results of less frequent surveillance testing, e.g., steam generator nondestructive testing (NDT), consideration should be given to the degradation that may occur between periodic tests in establishing the analysis value.

5.1.4 The NRC staff considers the implementation of an AST to be a significant change to the design basis of the facility that is voluntarily initiated by the licensee. In order to issue a license amendment authorizing the use of an AST and the TEDE dose criteria, the NRC staff must make a current finding of compliance with regulations applicable to the amendment. The characteristics of the ASTs and the revised dose calculational methodology may be incompatible with many of the analysis assumptions and methods currently reflected in the facility's design basis analyses. The NRC staff may find that new or unreviewed issues are created by a particular site

-specific implementation of the AST, warranting review of staff positions approved subsequent to the initial issuance of the license. This is not considered a backfit as defined by 10 CFR 50.109, "Backfitting." However, prior design bases that are unrelated to the use of the AST, or are unaffected Conforms- The FNP DBA analysis assumptions and methods are compatible with the AST and the TEDE criteria. to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 31 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis by the AST, may continue as the facility's design basis. Licensees should ensure that analysis assumptions and methods are compatible with the ASTs and the TEDE criteria.

5.2 The appendices to this regulatory guide provide accident

-specific assumptions that are acceptable to the staff for performing analyses that are required by 10 CFR 50.67. The DBAs addressed in these attachments were selected from accidents that may involve damage to irradiated fuel. This guide does not address DBAs with radiological consequences based on technical specification reactor or secondary coolant

-specific activities only. The inclusion or exclusion of a particular DBA in this guide should not be interpreted as indicating that an analysis of that DBA is required or not required. Licensees should analyze the DBAs that are affected by the specific proposed applications of an AST.

Conforms - See Tables B, C, D, E, F, and G of this Enclosure.

5.2 The NRC staff has determined that the analysis assumptions in the appendices to this guide provide an integrated approach to performing the individual analyses and generally expects licensees to address each assumption or propose acceptable alternatives. Such alternatives may be justifiable on the basis of plant-specific considerations, updated technical analyses, or, in some cases, a previously approved licensing basis consideration.

The assumptions in the appendices are deemed consistent with the AST identified in Regulatory Position 3 and internally consistent with each other. Although licensees are free to propose alternatives to these assumptions for consideration by the NRC staff, licensees should avoid use of previously approved staff positions that would adversely affect this consistency.

Conforms - See Tables B, C, D, E, F, and G of this Enclosure.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 32 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis 5.2 The NRC is committed to using probabilistic risk analysis (PRA) insights in its regulatory activities and will consider licensee proposals for changes in analysis assumptions based upon risk insights. The staff will not approve proposals that would reduce the defense in depth deemed necessary to provide adequate protection for public health and safety. In some cases, this defense in depth compensates for uncertainties in the PRA analyses and addresses accident considerations not adequately addressed by the core damage frequency (CDF) and large early release frequency (LERF) surrogate indicators of overall risk.

Conforms- PRA was not used as a basis for acceptability of this AST License Amendment Request. 5.3 Atmospheric dispersion values (X/Q) for the EAB, the LPZ, and the control room that were approved by the staff during initial facility licensing or in subsequent licensing proceedings may be used in performing the radiological analyses identified by this guide. Methodologies that have been used for determining X/Q values are documented in Regulatory Guides 1.3 and 1.4, Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," and the paper, "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19" (Refs. 6, 7, 22, and 28).

Conforms - The X/Q used for the EAB and the LPZ were previously approved by the NRC in License Amendment 165/157.

5.3 References 22 and 28 should be used if the FSAR X/Q values are to be revised or if values are to be determined for new release points or receptor distances. Fumigation should be considered where applicable for the EAB and LPZ. For the EAB, the assumed fumigation period should be timed to be included in the worst 2

-hour exposure period. The NRC computer code PAVAN (Ref. 29) implements Regulatory Guide 1.145 (Ref. 28) Not Applicable

- The X/Q values used are those described in the FNP FSAR.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 33 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis and its use is acceptable to the NRC staff. The methodology of the NRC computer code ARCON96 19 (Ref. 26) is generally acceptable to the NRC staff for use in determining control room X/Q values. Meteorological data collected in accordance with the site-specific meteorological measurements program described in the facility FSAR should be used in generating accident X/Q values. Additional guidance is provided in Regulatory Guide 1.23, "Onsite Meteorological Programs" (Ref. 30). All changes in X/Q analysis methodology should be reviewed by the NRC staff.

6.0 The assumptions in Appendix I to this guide are acceptable to the NRC staff for performing radiological assessments associated with equipment qualification. The assumptions in Appendix I will supersede Regulatory Positions 2.c(1) and 2.c(2) and Appendix D of Revision 1 of Regulatory Guide 1.89, "Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants" (Ref. 11), for operating reactors that have amended their licensing basis to use an alternative source term. Except as stated in Appendix I, all other assumptions, methods, and provisions of Revision 1 of Regulatory Guide 1.89 remain effective.

The NRC staff is assessing the effect of increased cesium releases on EQ doses to determine whether licensee action is warranted. Until such time as this generic issue is resolved, licensees may use either the AST or the TID14844 assumptions for performing the required EQ analyses. However, no plant modifications are required to address the impact of the difference in source term characteristics (i.e., AST vs TID14844) on EQ Conforms - FNP is retaining the use of the TID 14844 source term as the basis for Environmental Qualification.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 34 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis doses pending the outcome of the evaluation of the generic issue. Footnote 6 For example, a proposed modification to change the timing of a containment isolation valve from 2.5 seconds to 5.0 seconds might be acceptable without any dose calculations. However, a proposed modification that would delay containment spray actuation could involve recalculation of DBA LOCA doses, re

-assessment of the containment pressure and temperature transient, recalculation of sump pH, re

-assessment of the emergency diesel generator loading sequence, integrated doses to equipment in the containment, and more.

Conform s - No modifications are being proposed as part of this AST License Amendment Request.

Footnote 7 In performing screenings and evaluations pursuant to 10 CFR 50.59, it may be necessary to compare dose results expressed in terms of whole body and thyroid with new results expressed in terms of TEDE. In these cases, the previous thyroid dose should be multiplied by 0.03 and the product added to the whole body dose. The result is then compared to the TEDE result in the screenings and evaluations. This change in dose methodology is not considered a change in the method of evaluation if the licensee was previously authorized to use an AST and the TEDE criteria under 10 CFR 50.67.

Not Applicable

- This activity is a License Amendment Request made pursuant to 10 CFR Part

90. Footnote 8 The uncertainty factor used in determining the core inventory should be that value provided in Appendix K to 10 CFR Part 50, typically 1.02.

Conforms - A 1.02 uncertainty factor is used for those events resulting in fuel damage.

Footnote 9 Note that for some radionuclides, such as Cs

-137, equilibrium will not be reached prior to fuel offload. Thus, the maximum inventory at the end of life should be used.

Conforms - A conservative core factor is applied the principal radionuclides to account for cycle

-to-cycle variations.

Footnote The release fractions listed here have been determined to be Conforms - Burnup does not exceed 62,000 to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 35 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis 10 acceptable for use with currently approved LWR fuel with a peak burnup up to 62,000 MWD/MTU. The data in this section may not be applicable to cores containing mixed oxide (MOX) fuel.

MWD/MTU at FNP.

Footnote 11 The release fractions listed here have been determined to be acceptable for use with currently approved LWR fuel with a peak burnup up to 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. As an alternative, fission gas release calculations performed using

NRC-approved methodologies may be considered on a case

-by-case basis. To be acceptable, these calculations must use a projected power history that will bound the limiting projected plant-specific power history for the specific fuel load. For the BWR rod drop accident the PWR rod ejection accident, the gap fractions are assumed to be 10% for iodines and noble gases.

Conforms - Burnup does not exceed 54 GWD/MTU at FNP. Footnote 12 In lieu of treating the release in a linear ramp manner, the activity for each phase can be modeled as being released instantaneously at the start of that release phase, i.e., in step increases.

Conforms - Both RADTRAD and LOCADOSE can model the release either in a linear ramp manner, or instantaneous release, as required.

Footnote 13 The prior practice of basing inhalation exposure on only radioiodine and not including radioiodine in external exposure calculations is not consistent with the definition of TEDE and the characteristics of the revised source term.

Conforms - Offsite inhalation doses are calculated consistent with the definition of TEDE.

Footnote 14 With regard to the EAB TEDE, the maximum two

-hour value is the basis for screening and evaluation under 10 CFR 50.59. Changes to doses outside of the two

-hour window are only considered in the context of their impact on the maximum two

-hour EAB TEDE.

Not Applicable

- This activity is a License Amendment Request made pursuant to 10 CFR Part

90. to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 36 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis Footnote 15 The iodine protection factor (IPF) methodology of Reference 22 may not be adequately conservative for all DBAs and control room arrangements since it models a steady

-state control room condition. Since many analysis parameters change over the duration of the event, the IPF methodology should only be used with caution. The NRC computer codes HABIT (Ref. 23) and RADTRAD (Ref. 24) incorporate suitable methodologies.

Conforms - The iodine protection factor methodology of Reference 22 is not used in this application.

Footnote 16 This occupancy is modeled in the X/Q values determined in Reference 22 and should not be credited twice. The ARCON96 Code (Ref. 26) does not incorporate these occupancy assumptions, making it necessary to apply this correction in the dose calculations.

Conforms - The control room occupancy assumptions are incorporated in the dose calculations Footnote 17 For PWRs with steam generator alternative repair criteria, different dose criteria may apply to steam generator tube rupture and main steam line break analyses.

Conforms - Refer to ARC line items in Tables D and E. Footnote 18 Note that for some parameters, the technical specification value may be adjusted for analysis purposes by factors provided in other regulatory guidance. For example, ESF filter efficiencies are based on the guidance in Regulatory Guide 1.52 (Ref. 25) and in Generic Letter 99

-02 (Ref. 27) rather than the surveillance test criteria in the technical specifications. Generally, these adjustments address potential changes in the parameter between scheduled surveillance tests.

Conforms - There are no parameters used in the dose calculations that are directly based on Technical Specification surveillance requirements

. Footnote 19 The ARCON96 computer code contains processing options that may yield X/Q values that are not sufficiently conservative for use in accident consequence assessments or may be incompatible with release point and ventilation intake configurations at particular sites. The applicability of these options and associated Conforms - The ARCON96 processing options and input parameters were based on the release point and ventilation intake configurations at FNP.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 37 Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FNP Analysis input parameters should be evaluated on a case

-by-case basis. The assumptions made in the examples in the ARCON96 documentation are illustrative only and do not imply NRC staff acceptance of the methods or data used in the example.

Table B: Conformance With Regulatory Guide 1.183 Appendix A (Loss of Coolant Accident) RG Section RG Position FNP Analysis A-1 Acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are provided in Regulatory Position 3 of this guide.

Conforms - See discussions in Table A.

A-2 If the sump or suppression pool pH is controlled at values of 7 or greater, the chemical form of radioiodine released to the containment should be assumed to be 95% cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide.

Iodine species, including those from iodine re

-evolution, for sump or suppression pool pH values less than 7 will be evaluated on a case-by-case basis. Evaluations of pH should consider the effect of acids and bases created during the LOCA event, e.g., radiolysis products. With the exception of elemental and organic

iodine and noble gases, fission products should be assumed to be in particulate form.

Conforms - The pH of the containment sump is maintained equal to or greater than 7.0 after the onset of the spray recirculation mode. Therefore the radioiodine composition of 95 percent cesium iodide, 4.85 percent elemental iodine, and 0.15 percent organic iodide is used. A-3.1 The radioactivity released from the fuel should be assumed to mix instantaneously and homogeneously throughout the free air volume of the primary containment in PWRs or the drywell in BWRs as it is released. This distribution should be adjusted if Conforms - The radioactivity released from the fuel is modeled as mixing instantaneously and homogeneously in the Containment.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 38 Table B: Conformance With Regulatory Guide 1.183 Appendix A (Loss of Coolant Accident) RG Section RG Position FNP Analysis there are internal compartments that have limited ventilation exchange. The suppression pool free air volume may be included provided there is a mechanism to ensure mixing between the drywell to the wetwell. The release into the containment or drywell should be assumed to terminate at the end of the early in

-vessel phase.

A-3.2 Reduction in airborne radioactivity in the containment by natural deposition within the containment may be credited. Acceptable models for removal of iodine and aerosols are described in Chapter 6.5.2, "Containment Spray as a Fission Product Cleanup System," of the Standard Review Plan (SRP), NUREG

-0800 (Ref. A-1) and in NUREG/CR

-6189, "A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments" (Ref. A-2). The latter model is incorporated into the analysis code RADTRAD (Ref. A

-3). The prior practice of deterministically assuming that a 50% plateout of iodine is released from the fuel is no longer acceptable to the NRC staff as it is inconsistent with the characteristics of the revised source terms.

Conforms - An aerosol natural deposition rate of 0.1 hr-1 is assumed based upon values presented Section VI of NUREG/CR

-6189. A-3.3 Reduction in airborne radioactivity in the containment by containment spray systems that have been designed and are maintained in accordance with Chapter 6.5.2 of the SRP (Ref. A-1) may be credited. Acceptable models for the removal of iodine and aerosols are described in Chapter 6.5.2 of the SRP a nd NUREG/CR-5966, "A Simplified Model of Aerosol Removal by Containment Sprays"1 (Ref. A-4). This simplified model is incorporated into the analysis code RADTRAD (Refs. A

-1 to A-3). Conforms - Containment Spray is credited for elemental and particulate iodine removal.

A-3.3 The evaluation of the containment sprays should address areas within the primary containment that are not covered by the spray Conforms - Containment Spray covers less than 90% of the Containment volume, so the modeling includes to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 39 Table B: Conformance With Regulatory Guide 1.183 Appendix A (Loss of Coolant Accident) RG Section RG Position FNP Analysis drops. The mixing rate attributed to natural convection between sprayed and unsprayed regions of the containment building, provided that adequate flow exists between these regions, is assumed to be two turnovers of the unsprayed regions per hour, unless other rates are justified. The containment building atmosphere may be considered a single, well

-mixed volume if the spray covers at least 90%

of the volume and if adequate mixing of unsprayed compartments can be shown.

both the sprayed volume and unsprayed volume. A flow rate of 12,045 cfm is used between the sprayed and unsprayed volume which correlates to two turnovers of the unsprayed region per hour.

A-3.3 The SRP sets forth a maximum decontamination factor (DF) for elemental iodine based on the maximum iodine activity in the primary containment atmosphere when the sprays actuate, divided by the activity of iodine remaining at some time after decontamination.

The SRP also states that the particulate iodine removal rate should be reduced by a factor of 10 when a DF of 50 is reached. The reduction in the removal rate is not required if the removal rate is based on the calculated time

-dependent airborne aerosol mass. There is no specified maximum DF for aerosol removal by sprays. The maximum activity to be used in

determining the DF is defined as the iodine activity in the columns labeled "Total" in Tables 1 and 2 of this guide multiplied by 0.05 for elemental iodine and by 0.95 for particulate iodine (i.e., aerosol treated as particulate in SRP methodology).

Conforms - Elemental and aerosol removal coefficients are calculated for the sprayed regions of the containment using the guidelines of Chapter 6.5.2 of the Standard Review Plan. The elemental iodine removal coefficients are limited to a maximum value of 13.7/hr, and are set to zero when the elemental iodine decontamination factor (DF) reaches a value of 200. The aerosol removal coefficients are reduced by a factor of 10 when the aerosol DF reaches 50.

A-3.4 Reduction in airborne radioactivity in the containment by in

-containment recirculation filter systems may be credited if these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99

-02 (Refs. A

-5 and A-6). The filter media loading caused by the increased aerosol release associated with the revised source term should be addressed.

Conforms - No credit is taken for in

-containment recirculation filter systems.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 40 Table B: Conformance With Regulatory Guide 1.183 Appendix A (Loss of Coolant Accident) RG Section RG Position FNP Analysis A-3.5 Reduction in airborne radioactivity in the containment by suppression pool scrubbing in BWRs should generally not be credited. However, the staff may consider such reduction on an individual case basis. The evaluation should consider the relative timing of the blowdown and the fission product release from the fuel, the force driving the release through the pool, and the potential for any bypass of the suppression pool (Ref. 7). Analyses should consider iodine re-evolution if the suppression pool liquid pH is not maintained greater than 7.

Not Applicable

- FNP is a PWR.

A-3.6 Reduction in airborne radioactivity in the containment by retention in ice condensers, or other engineering safety features not addressed above, should be evaluated on an individual case basis. See Section 6.5.4 of the SRP (Ref. A

-1). Conforms - No credit is taken for ice condensers or other engineering safety features to reduce airborne radioactivity in containment.

A-3.7 The primary containment (i.e., drywell for Mark I and II containment designs) should be assumed to leak at the peak pressure technical specification leak rate for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . For PWRs, the leak rate may be reduced after the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 50% of the technical specification leak rate. For BWRs, leakage may be reduced after the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , if supported by plant configuration and analyses, to a value not less than 50% of the technical specification leak rate. Leakage from subatmospheric containments is assumed to terminate when the containment is brought to and maintained at a subatmospheric condition as defined by technical specifications.

Conforms - The containment leak rate for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is the maximum value allowed by the FNP Technical Specifications. It is reduced to 50% of that value after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

A-3.7 For BWRs with Mark III containments, the leakage from the drywell into the primary containment should be based on the steaming rate of the heated reactor core, with no credit for core debris relocation. This leakage should be assumed during the Not Applicable. FNP is a PWR.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 41 Table B: Conformance With Regulatory Guide 1.183 Appendix A (Loss of Coolant Accident) RG Section RG Position FNP Analysis two-hour period between the initial blowdown and termination of the fuel radioactivity release (gap and early in-vessel release phases). After two hours, the radioactivity is assumed to be uniformly distributed throughout the drywell and the primary containment.

A-3.8 If the primary containment is routinely purged during power operations, releases via the purge system prior to containment isolation should be analyzed and the resulting doses summed with the postulated doses from other release paths. The purge release evaluation should assume that 100% of the radionuclide inventory in the reactor coolant system liquid is released to the containment at the initiation of the LOCA. This inventory should be based on the technical specification reactor coolant system equilibrium activity. Iodine spikes need not be considered. If the purge system is not isolated before the onset of the gap release phase, the release fractions associated with the gap release and early in-vessel phases should be considered as applicable.

Conforms - Based upon the isolation of the mini

-purge flow within 30 seconds, the mini

-purge system will be isolated before the onset of the gap release as defined in Table 4 of this Regulatory Guide. Therefore, only those nuclides in the RCS source term are available for release.

A-4 For facilities with dual containment systems, the acceptable assumptions related to the transport, reduction, and release of radioactive material in and from the secondary containment or enclosure buildings are as follows.

Not Applicable. FNP does not have a dual containment.

A-5.1 With the exception of noble gases, all the fission products released from the fuel to the containment (as defined in Tables 1 and 2 of this guide) should be assumed to instantaneously and homogeneously mix in the primary containment sump water (in PWRs) or suppression pool (in BWRs) at the time of release from the core. In lieu of this deterministic approach, suitably conservative mechanistic models for the transport of airborne Conforms - With the exception of noble gases, all the fission products released from the fuel to the containment instantaneously and homogeneously mix in the primary sump water.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 42 Table B: Conformance With Regulatory Guide 1.183 Appendix A (Loss of Coolant Accident) RG Section RG Position FNP Analysis activity in containment to the sump water may be used. Note that many of the parameters that make spray and deposition models conservative with regard to containment airborne leakage are nonconservative with regard to the buildup of sump activity.

A-5.2 The leakage should be taken as two times the sum of the simultaneous leakage from all components in the ESF recirculation systems above which the technical specifications, or licensee commitments to item III.D.1.1 of NUREG

-0737 (Ref. A

-8), would require declaring such systems inoperable. The leakage should be assumed to start at the earliest time the recirculation flow occurs in these systems and end at the latest time the releases from these systems are terminated. Consideration should also be given to design leakage through valves isolating ESF recirculation systems from tanks vented to atmosphere, e.g., emergency core cooling system (ECCS) pump miniflow return to the refueling water storage tank.

The FNP Technical Specifications do not provide a specific limit for operational leakage from ECCS systems. However, administrative limits ensure that operational leakage is adequately controlled. In the analysis, an assumed leakage from ECCS systems is taken as 20,000 cc/hr for leakage of sump water outside of containment into the Auxiliary Building , which is multiplied by two, consistent with this Regulatory Position.

In addition, two times the assumed leak rate o f 1.0 gpm past valves that isolate return flow to the Refueling Water Storage Tank (RWST) is evaluated separately. The leakage is assumed to start at the earliest time that recirculation occurs in the ECCS systems and continues for the 30-day duration of the event. A-5.3 With the exception of iodine, all radioactive materials in the recirculating liquid should be assumed to be retained in the liquid phase. Conforms - With the exception of iodine, all radioactive materials in the recirculating liquid is modeled as being retained in the liquid phase.

A-5.4 If the temperature of the leakage exceeds 212°F, the fraction of total iodine in the liquid that becomes airborne should be assumed equal to the fraction of the leakage that flashes to vapor. This flash fraction, FF, should be determined using a constant enthalpy, h, process, based on the maximum time

-dependent temperature of the sump water circulating outside the Conforms - It is assumed for the case when the temperature of the ESF leakage exceeds 212 F that the fraction of total iodine in the liquid that becomes airborne is equal to the fraction of the leakage that flashes to vapor. This flash fraction, FF, is determined assuming a constant enthalpy, h, process, and is to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 43 Table B: Conformance With Regulatory Guide 1.183 Appendix A (Loss of Coolant Accident) RG Section RG Position FNP Analysis containment:

FF = fgff hhh 2 1 Where: h f1 is the enthalpy of liquid at system design temperature and pressure; h f2 is the enthalpy of liquid at saturation conditions (14.7 psia, 212ºF); and h fg is the heat of vaporization at 212ºF.

based on the maximum time

-dependent sump water temperature. A-5.5 If the temperature of the leakage is less than 212°F or the calculated flash fraction is less than 10%, the amount of iodine that becomes airborne should be assumed to be 10% of the total iodine activity in the leaked fluid, unless a smaller amount can be justified based on the actual sump pH history and area ventilation rates. Conforms - Since the calculated flashing fraction is less than 10%, and without a basis for justifying a smaller value, 10% of the iodine in the ESF leakage is assumed to be released.

A-5.6 The radioiodine that is postulated to be available for release to the environment is assumed to be 97% elemental and 3% organic. Reduction in release activity by dilution or holdup within buildings, or by ESF ventilation filtration systems, may be credited where applicable. Filter systems used in these applications should be evaluated against the guidance of Regulatory Guide 1.52 (Ref. A

-5) and Generic Letter 99

-02 (Ref. A-6). Conforms - The radioiodine that is postulated to be available for release to the environment is modeled as 97% elemental and 3% organic.

A-6 For BWRs, the main steam isolation valves (MSIVs) have design leakage that may result in a radioactivity release. The radiological consequences from postulated MSIV leakage should be analyzed and combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA. The following Not Applicable. FNP is a PWR.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 44 Table B: Conformance With Regulatory Guide 1.183 Appendix A (Loss of Coolant Accident) RG Section RG Position FNP Analysis assumptions are acceptable for evaluating the consequences of MSIV leakage.

A-7 The radiological consequences from post

-LOCA primary containment purging as a combustible gas or pressure control measure should be analyzed. If the installed containment purging capabilities are maintained for purposes of severe accident management and are not credited in any design basis analysis, radiological consequences need not be evaluated. If the primary containment purging is required within 30 days of the LOCA, the results of this analysis should be combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA. Reduction in the amount of radioactive material released via ESF filter systems may be taken into account provided that these systems meet the guidance in Regulatory Guide 1.52 (Ref. A

-5) and Generic Letter 99-02 (Ref. A

-6). Conforms - FNP use s hydrogen recombiners for post-accident hydrogen control. As such, the containment mini

-purge system is assumed to not be available for combustible gas management and this pathway is assumed to remain closed following a containment isolation signal.

Footnote A-1 This document describes statistical formulations with differing levels of uncertainty. The removal rate constants selected for use in design basis calculations should be those that will maximize the dose consequences. For BWRs, the simplified model should be used only if the release from the core is not directed through the suppression pool. Iodine removal in the suppression pool affects the iodine species assumed by the model to be present initially.

Conforms - The removal rate constants selected for use in the LOCA calculation are those that will maximize the dose consequences.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 45 Table C: Conformance With Regulatory Guide 1.183 Appendix B (Fuel Handling Accident)

RG Section RG Position FNP Analysis B-1 Acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are provided in Regulatory Position 3 of this guide.

Conforms - See discussions in Table A.

B-1.1 The number of fuel rods damaged during the accident should be based on a conservative analysis that considers the most limiting case. This analysis should consider parameters such as the weight of the dropped heavy load or the weight of a dropped fuel assembly (plus any attached handling grapples), the height of the drop, and the compression, torsion, and shear stresses on the irradiated fuel rods. Damage to adjacent fuel assemblies, if applicable (e.g., events over the reactor vessel), should be considered.

Conforms - The FHA is a single fuel assembly dropped from within either the Containment, the Fuel Handling Building, or the Auxiliary Building without interaction with any other fuel assemblies. The number of fuel rods damaged is equal to one fuel assembly. B-1.2 The fission product release from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached. All the gap activity in the damaged rods is assumed to be instantaneously released.

Radionuclides that should be considered include xenons, kryptons, halogens, cesiums, and rubidiums.

Conforms - The fission product release is equal to the gap release, with isotopic fractions as given in Table 3 of RG 1.183 (8% for I-131, 10% for Kr

-85, and 5% for the other Halogens and Noble Gases).

Cycle to cycle fuel load variations are accounted for with adjustments to the core source term: +15% for Kr

-85, +5% for Xe

-133, and +3% for the other isotopes.

B-1.3 The chemical form of radioiodine released from the fuel to the spent fuel pool should be assumed to be 95% cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. The CsI released from the fuel is assumed to completely dissociate in the pool water. Because of the low pH of the pool water, the iodine re

-evolves as elemental iodine. This is assumed to occur instantaneously. The NRC staff will consider, on a case

-by-case basis, justifiable mechanistic treatment of the iodine release from the pool. Conforms - The chemical forms of radioiodine released from the fuel to the spent fuel pool is assumed to be 95% cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide.

The CsI released from the fuel completely dissociates in the pool water and re-evolves as elemental iodine.

The dissociation and re

-evolution occurs instantaneously.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 46 Table C: Conformance With Regulatory Guide 1.183 Appendix B (Fuel Handling Accident)

RG Section RG Position FNP Analysis B-2 If the depth of water above the damaged fuel is 23 feet or greater, the decontamination factors for the elemental and organic species are 500 and 1, respectively, giving an overall effective decontamination factor of 200 (i.e., 99.5% of the total iodine released from the damaged rods is retained by the water). This difference in decontamination factors for elemental (99.85%) and organic iodine (0.15%) species results in the iodine above the water being composed of 57% elemental and 43% organic species. If the depth of water is not 23 feet, the decontamination factor will have to be determined on a case

-by-case method (Ref. B

-1). Conforms - Water level is greater than 23 feet for each case. Therefore, the pool water is assumed to have a decontamination factor of 500 for iodine isotopes in an organic form. This assumption leads to an overall effective decontamination factor of 200 for the iodine isotopes released from the gap.

B-3 The retention of noble gases in the water in the fuel pool or reactor cavity is negligible (i.e., decontamination factor of 1). Particulate radionuclides are assumed to be retained by the water in the fuel pool or reactor cavity (i.e., infinite decontamination factor).

Conforms - Noble gases are not scrubbed by the pool water (decontamination factor of 1). Particulate releases are assumed to be entirely scrubbed (infinite decontamination factor).

B-4.1 The radioactive material that escapes from the fuel pool to the fuel building is assumed to be released to the environment over a 2-hour time period.

Conforms - For releases in containment and the Fuel Handling Building , the FNP fuel handling analysis considers a release to the environment over a 2-hour time period

. B-4.2 A reduction in the amount of radioactive material released from the fuel pool by engineered safety feature (ESF) filter systems may be taken into account provided these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99

-02 (Refs. B-2, B-3). Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system1 should be determined and accounted for in the radioactivity release analyses.

Conforms - A reduction in the amount of radioactive material released from the spent fuel pool area of the Auxiliary Building is credited by use of the Penetration Room Filter (PRF) system. This system meets the requirements of Reg. Guide 1.52 and is required to be in service prior to the movement of irradiated fuel in the building. It is analyzed as actuating from the source term seen from the FHA.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 47 Table C: Conformance With Regulatory Guide 1.183 Appendix B (Fuel Handling Accident)

RG Section RG Position FNP Analysis B-4.3 The radioactivity release from the fuel pool should be assumed to be drawn into the ESF filtration system without mixing or dilution in the fuel building. If mixing can be demonstrated, credit for mixing and dilution may be considered on a case

-by-case basis.

This evaluation should consider the magnitude of the building volume and exhaust rate, the potential for bypass to the environment, the location of exhaust plenums relative to the surface of the pool, recirculation ventilation systems, and internal walls and floors that impede stream flow between the surface of the pool and the exhaust plenums.

Conforms - There is no credit taken for mixing or dilution in the spent fuel pool are of the Auxiliary Building. B-5.1 If the containment is isolated 2 during fuel handling operations, no radiological consequences need to be analyzed.

Not Applicable

- Containment is not assumed to be isolated during fuel handling operations.

B-5.2 If the containment is open during fuel handling operations, but designed to automatically isolate in the event of a fuel handling accident, the release duration should be based on delays in radiation detection and completion of containment isolation. If it can be shown that containment isolation occurs before radioactivity is released to the environment, 1 no radiological consequences need to be analyzed.

Not Applicable

-The containment equipment hatch and personnel airlock are modeled as being open during an FHA and no credit it taken in the analysis for closing them. B-5.3 If the containment is open during fuel handling operations (e.g., personnel air lock or equipment hatch is open), 3 the radioactive material that escapes from the reactor cavity pool to the containment is released to the environment over a 2

-hour time period. Conforms - The FHA radiological release is over a two-hour period.

B-5.4 A reduction in the amount of radioactive material released from the containment by ESF filter systems may be taken into account provided that these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99

-02 (Refs. B

-2 and Not Applicable

- No credit is taken for ESF filter systems to mitigate radioactive material release from the Containment.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 48 Table C: Conformance With Regulatory Guide 1.183 Appendix B (Fuel Handling Accident)

RG Section RG Position FNP Analysis B-3). Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system should be determined and accounted for in the radioactivity release analyses.

1 B-5.5 Credit for dilution or mixing of the activity released from the reactor cavity by natural or forced convection inside the containment may be considered on a case

-by-case basis. Such credit is generally limited to 50% of the containment free volume. This evaluation should consider the magnitude of the containment volume and exhaust rate, the potential for bypa ss to the environment, the location of exhaust plenums relative to the surface of the reactor cavity, recirculation ventilation systems, and internal walls and floors that impede stream flow between the surface of the reactor cavity and the exhaust plenums. The free volume of the FNP containment is 2.0E6 cubic feet. The free volume used in the FHA dose calculation was 1.0E6 cubic feet.

Footnote B-1 These analyses should consider the time for the radioactivity concentration to reach levels corresponding to the monitor setpoint, instrument line sampling time, detector response time, diversion damper alignment time, and filter system actuation, as applicable.

Conforms - The FHA calculation demonstrates that a sufficient concentration of radioactivity occurs at the Control Room Ventilation System sensor to result in control room isolation within the assumed 60 second delay time.

Footnote B-2 Containment isolation does not imply containment integrity as defined by technical specifications for non

-shutdown modes.

The term isolation is used here collectively to encompass both containment integrity and containment closure, typically in place during shutdown periods. To be credited in the analysis, the appropriate form of isolation should be addressed in technical specifications.

Not Applicable

- Containment is not assumed to be isolated during fuel handling operations.

Footnote The staff will generally require that technical specifications Conforms - FNP TS 3.9.3 establishes the to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 49 Table C: Conformance With Regulatory Guide 1.183 Appendix B (Fuel Handling Accident)

RG Section RG Position FNP Analysis B-3 allowing such operations include administrative controls to close the airlock, hatch, or open penetrations within 30 minutes. Such administrative controls will generally require that a dedicated individual be present, with necessary equipment available, to restore containment closure should a fuel handling accident occur. Radiological analyses should generally not credit this manual isolation.

requirements for containment penetrations during refueling operations. The FHA dose calculation takes no credit for manual isolation of containment after the event.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 50 Table D: Conformance With Regulatory Guide 1.183 Appendix E (Main Steam Line Break Accident

) RG Section RG Position FNP Analysis E-1 Assumptions acceptable to the NRC staff regarding core inventory and the release of radionuclides from the fuel are provided in Regulatory Position 3 of this regulatory guide. The release from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached. The fuel damage estimate should assume that the highest worth control rod is stuck at its fully withdrawn position.

Conforms - See discussions in Table A.

E-2 If no or minimal fuel damage is postulated for the limiting event, the activity released should be the maximum coolant activity allowed by the technical specifications. Two cases of iodine spiking should be assumed.

Consistent with the FNP current licensing basis a failed fuel term is conservatively included with the two cases of iodine spiking.

E-2.1 A reactor transient has occurred prior to the postulated main steam line break (MSLB) and has raised the primary coolant iodine concentration to the maximum value (typically 60 µCi/gm DE I-131) permitted by the technical specifications (i.e., a preaccident iodine spike case).

Conforms - The Main Steam Line Break Accident dose calculation includes a case for a preaccident iodine spike with the maximum iodine concentration permitted by the FNP technical specifications.

E-2.2 The primary system transient associated with the MSLB causes an iodine spike in the primary system. The increase in primary coolant iodine concentration is estimated using a spiking model that assumes that the iodine release rate from the fuel rods to the primary coolant (expressed in curies per unit time) increases to a value 500 times greater than the release rate corresponding to the iodine concentration at the equilibrium value (typically 1.0 µCi/gm DE I

-131) specified in technical specifications (i.e., concurrent iodine spike case). A concurrent iodine spike need not be considered if fuel damage is postulated.

The assumed iodine spike duration should be 8 Conforms - The Main Steam Line Break Accident dose calculation includes a case for a concurrent iodine spike causing the iodine release rate from the fuel rods to the RCS to increase to a value 500 times greater than the release rate that yields the equilibrium iodine concentration specified in the technical specifications.

The iodine spike duration is 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . For conservatism, the concurrent iodine spike is assumed even with 1% fuel damage.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 51 Table D: Conformance With Regulatory Guide 1.183 Appendix E (Main Steam Line Break Accident

) RG Section RG Position FNP Analysis hours. Shorter spike durations may be considered on a case

-by-case basis if it can be shown that the activity released by the 8-hour spike exceeds that available for release from the fuel gap of all fuel pins.

E-3 The activity released from the fuel should be assumed to be released instantaneously and homogeneously through the primary coolant.

Conforms - The activity from the fuel is assumed to be released instantaneously and homogeneously to the reactor coolant system. E-4 The chemical form of radioiodine released from the fuel should be assumed to be 95%

cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. Iodine releases from the steam generators to the environment should be assumed to be 97% elemental and 3% organic. These fractions apply to iodine released as a result of fuel damage and to iodine released during normal operations, including iodine spiking.

Conforms - The iodine releases from the steam generators to the environment are 97% elemental and 3% organic for the pre

-accident case and the concurrent iodine spike case, including failed fuel. E-5.1 For facilities that have not implemented alternative repair criteria (see Ref. E

-1, DG-1074), the primary

-to-secondary leak rate in the steam generators should be assumed to be the leak rate limiting condition for operation specified in the technical specifications.

For facilities with traditional generator specifications (both per generator and total of all generators), the leakage should be apportioned between affected and unaffected steam generators in such a manner that the calculated dose is maximized.

Conforms - FNP is licensed to ARC. The assumed primary-to-secondary leak rate in the two intact steam generators are 0.65 gpm (936 gallons per day). This is conservative relative to FNP TS 3.4.13 which allows 150 gallons per day per Steam Generator.

E-5.2 The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g., lbm/hr) should be consistent with the basis of the parameter being converted. The ARC leak rate correlations are generally based on the collection of cooled Conforms - The assumed density is 62.4 lbm/ft

3. to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 52 Table D: Conformance With Regulatory Guide 1.183 Appendix E (Main Steam Line Break Accident

) RG Section RG Position FNP Analysis liquid. Surveillance tests and facility instrumentation used to show compliance with leak rate technical specifications are typically based on cooled liquid. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft 3). E-5.3 The primary

-to-secondary leakage should be assumed to continue until the primary system pressure is less than the secondary system pressure, or until the temperature of the leakage is less than 100°C (212°F). The release of radioactivity from unaffected steam generators should be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated.

Conforms - For the faulted steam generator, primary

-to-secondary leakage continues for the duration of the event. The release from the unaffected steam generators continues until the Reactor Coolant System is reduced tocold shutdown conditions in 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . E-5.4 All noble gas radionuclides released from the primary system are assumed to be released to the environment without reduction or mitigation.

Conforms - All noble gases are released from the steam generator water without credit for scrubbing.

E-5.5 The transport model described in this section should be utilized for iodine and particulate releases from the steam generators. This model is shown in Figure E

-1 and summarized below: Conforms - See below. E-5.5.1 A portion of the primary

-to-secondary leakage will flash to vapor, based on the thermodynamic conditions in the reactor and secondary coolant.

During periods of steam generator dryout, all of the primary-to-secondary leakage is assumed to flash to vapor and be released to the environment with no mitigation.

With regard to the unaffected steam generators used for plant cooldown, the primary

-to-secondary leakage can be assumed to mix with the secondary water without flashing during periods of total tube submergence.

Conforms - The leakage of the faulted steam generator is modeled as a direct vapor flow from the RCS to the environment without partitioning. For the intact steam generators, primary

-to-secondary leakage mixes with the secondary water without flashing for the duration of the event.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 53 Table D: Conformance With Regulatory Guide 1.183 Appendix E (Main Steam Line Break Accident

) RG Section RG Position FNP Analysis E-5.5.2 The leakage that immediately flashes to vapor will rise through the bulk water of the steam generator and enter the steam space. Credit may be taken for scrubbing in the generator, using the models in NUREG

-0409, "Iodine Behavior in a PWR Cooling System Following a Postulated Steam Generator Tube Rupture Accident" (Ref. E-2), during periods of total submergence of the tubes.

Conforms - For conservatism, no credit is taken for scrubbing.

E-5.5.3 The leakage that does not immediately flash is assumed to mix with the bulk water. Conforms - The leakage that does not immediately flash mixes with the bulk water.

E-5.5.4 The radioactivity in the bulk water is assumed to become vapor at a rate that is the function of the steaming rate and the partition coefficient. A partition coefficient for iodine of 100 may be assumed. The retention of particulate radionuclides in the steam generators is limited by the moisture carryover from the steam generators.

Conforms - For flows out of the intact SGs, radioactivity to the environment is a function of the steaming rate, and the iodine partition factor is assumed to be 100. Moisture carryover is modeled at 0.1%. E-5.6 Operating experience and analyses have shown that for some steam generator designs, tube uncovery may occur for a short period following any reactor trip (Ref. E

-3). The potential impact of tube uncovery on the transport model parameters (e.g., flash fraction, scrubbing credit) needs to be considered. The impact of emergency operating procedure restoration strategies on steam generator water levels should be evaluated.

Conforms - The steam generator with the faulted main steamline in the MSLB accident is assumed to blow completely dry, causing a direct release of radioactivity from that soulrce to the environment.

Footnote E-1 Facilities licensed with, or applying for, alternative repair criteria (ARC) should use this section in conjunction with the guidance that is being developed in Draft Regulatory Guide DG

-1074, "Steam Generator Tube Integrity," for acceptable assumptions and methodologies for performing radiological analyses.

Conforms - FNP is licensed to ARC.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 54 Table D: Conformance With Regulatory Guide 1.183 Appendix E (Main Steam Line Break Accident

) RG Section RG Position FNP Analysis Footnote E-2 The activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximum technical specification values, whichever maximizes the radiological consequences. In determining dose equivalent I-131 (DE I-131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.

Consistent with the FNP current licensing basis a failed fuel term is conservatively included with the two cases of iodine spiking.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 55 Table E: Conformance With Regulatory Guide 1.183 Appendix F (Steam Generator Tube Rupture Accident

) RG Section RG Position FNP Analysis F-1 Assumptions acceptable to the NRC staff regarding core inventory and the release of radionuclides from the fuel are in Regulatory Position 3 of this guide. The release from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached.

Conforms - See discussions in Table A.

F-2 If no or minimal 2 fuel damage is postulated for the limiting event, the activity released should be the maximum coolant activity allowed by technical specification. Two cases of iodine spiking should be assumed.

Consistent with the FNP current licensing basis a failed fuel term is conservatively included with the two cases of iodine spiking.

F-2.1 A reactor transient has occurred prior to the postulated steam generator tube rupture (SGTR) and has raised the primary coolant iodine concentration to the maximum value (typically 60 µCi/gm DE I

-131) permitted by the technical specifications (i.e., a preaccident iodine spike case).

Conforms - Case 1 is a pre

-accident spike using th e maximum Dose Equivalent Iodine permitted by the FNP Technical Specifications.

F-2.2 The primary system transient associated with the SGTR causes an iodine spike in the primary system. The increase in primary coolant iodine concentration is estimated usi ng a spiking model that assumes that the iodine release rate from the fuel rods to the primary coolant (expressed in curies per unit time) increases to a value 335 times greater than the release rate corresponding to the iodine concentration at the equilibrium value (typically 1.0 µCi/gm DE I

-131) specified in technical specifications (i.e., concurrent iodine spike case). A concurrent iodine spike need not be considered if fuel damage is postulated.

The assumed iodine spike duration should be 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Shorter spike durations may be considered on a case

-by-case basis if it can be shown that the activity released by the 8-hour spike exceeds that available for release from the Conforms - The concurrent iodine spike case assumes the RCS transient associated with the accident creates an iodine spike, causing the iodine release rate from the fuel rods to the RCS to increase to a value 335 times greater than the release rate that yields the equilibrium iodine concentration specified in the technical specifications.

A 1% failed fuel source term is conservatively included, consistent with the FNP current licensing basis for this event. An 8

-hour release duration is modeled.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 56 Table E: Conformance With Regulatory Guide 1.183 Appendix F (Steam Generator Tube Rupture Accident

) RG Section RG Position FNP Analysis fuel gap of all fuel pins.

F-3 The activity released from the fuel, if any, should be assume d to be released instantaneously and homogeneously through the primary coolant.

Conforms - Mixing in the primary coolant is assumed to be instantly and homogeneously.

F-4 Iodine releases from the steam generators to the environment should be assumed to be 97% elemental and 3% organic.

Conforms - The iodine released to the environment is assumed to be 97% elemental and 3% organic

. F-5.1 The primary

-to-secondary leak rate in the steam generators should be assumed to be the leak rate limiting condition for operation specified in the technical specifications. The leakage should be apportioned between affected and unaffected steam generators in such a manner that the calculated dose is maximized.

Conforms - The assumed primary

-to-secondary leak rate in the two intact steam generators are 0.65 gpm (936 gallons per day). This is conservative relative to FNP TS 3.4.13 which allows 150 gallons per day per Steam Generator. F-5.2 The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g., lbm/hr) should be consistent with the basis of surveillance tests used to show compliance with leak rate technical specifications. These tests are typically based on cool liquid.

Facility instrumentation used to determine leakage is typically located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft 3). Conforms - The assumed density is 62.4 lbm/ft

3. F-5.3 The primary

-to-secondary leakage should be assumed to continue until the primary system pressure is less than the secondary system pressure, or until the temperature of the

leakage is less than 100°C (212°F). The release of radioactivity from the unaffected steam generators should be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated.

Conforms - It is assumed that cold shutdown is established at 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , terminating the accident.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 57 Table E: Conformance With Regulatory Guide 1.183 Appendix F (Steam Generator Tube Rupture Accident

) RG Section RG Position FNP Analysis F-5.4 The release of fission products from the secondary system should be evaluated with the assumption of a coincident loss of offsite power.

Conforms - The SGTR assumes a concurrent LOOP to maximize the release to the environment. However, continued feedwater flow is modeled with its secondary side iodine contribution for conservatism.

F-5.5 All noble gas radionuclides released from the primary system are assumed to be rel eased to the environment without reduction or mitigation.

Conforms - Noble gases are modeled as going directly to the environment without reduction or mitigation.

F-5.6 The transport model described in Regulatory Positions 5.5 and 5.6 of Appendix E should be utilized for iodine and particulates.

Conforms - The transport model described in Position 5.5 and 5.6 of Appendix E is applied to releases from the steam generators.

Footnote F-1 Facilities licensed with, or applying for, alternative repair criteria (ARC) should use this section in conjunction with the guidance that is being developed in Draft Regulatory Guide DG

-1074, "Steam Generator Tube Integrity" (USNRC, December 1998), for acceptable assumptions and methodologies for performing radiological analyses.

Conforms - FNP is licensed to ARC.

Footnote F-2 The activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximum technical specification values, whichever maximizes the radiological consequences. In determining dose equivalent I-131 (DE I-131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.

Consistent with the FNP current licensing basis a failed fuel term is conservatively included with the two cases of iodine spiking.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 58 to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 59 Table F: Conformance With Regulatory Guide 1.183 Appendix G (Locked Rotor Accident) RG Section RG Position FNP Analysis G-1 Assumptions acceptable to the NRC staff regarding core inventory and the release of radionuclides from the fuel are in Regulatory Position 3 of this regulatory guide. The release from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached.

Conforms - See discussions in Table A.

G-2 If no fuel damage is postulated for the limiting event, a radiological analysis is not required as the consequences of this event are bounded by the consequences projected for the main steam line break outside containment.

Conforms - The transient causes fuel damage and so a radiological analysis is provided.

G-3 The activity released from the fuel should be assumed to be released instantaneously and homogeneously through the primary coolant.

Conforms - The gap activity in the damaged rods is instantaneously released to and uniformly mixed within the reactor coolant system at the onset of the accident. G-4 The chemical form of radioiodine released from the fuel should be assumed to be 95%

cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. Iodine

releases from the steam generators to the environment should be assumed to be 97% elemental and 3% organic. These fractions apply to iodine released as a result of fuel damage and to iodine released during normal operations, including iodine spiking.

Conforms - The iodine releases from the steam generators to the environment are 97%

elemental and 3% organic for the pre

-accident case and the concurrent iodine spike case, including failed fuel.

G-5.1 The primary

-to-secondary leak rate in the steam generators should be assumed to be the leak-rate-limiting condition for operation specified in the technical specifications. The leakage should be apportioned between the steam generators in such a manner that the calculated dose is maximized.

Conforms - Leakage is 1 gpm, which is bounding over the Technical Specification limit of 150 gallons per day per steam generator.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 60 Table F: Conformance With Regulatory Guide 1.183 Appendix G (Locked Rotor Accident) RG Section RG Position FNP Analysis G-5.2 The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g., lbm/hr) should be consistent with the basis of surveillance tests used to show compliance with leak rate technical specifications. These tests are typically based on cool liquid.

Facility instrumentation used to determine leakage is typically located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft 3). Conforms - The assumed density is 62.4 lbm/ft

3. G-5.3 The primary

-to-secondary leakage should be assumed to continue until the primary system pressure is less than the secondary system pressure, or until the temperature of the

leakage is less than 100°C (212°F). The release of radioactivity should be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated.

Conforms - The accident terminates after 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and cold shutdown conditions have been achieved.

G-5.4 The release of fission products from the secondary system should be evaluated with the assumption of a coincident loss of offsite power.

Conforms - The Locked Rotor Accident assumes a concurrent LOOP to maximize the release to the environment. However, continued feedwater flow is modeled with its secondary side iodine contribution for conservatism.

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

Conform s - Noble gases are assumed to leak directly to the environment without holdup in the SG.

G-5.6 The transport model described in assumptions 5.5 and 5.6 of Appendix E should be utilized for iodine and particulates.

Conforms - The transport model described in Position 5.5 and 5.6 of Appendix E is applied to releases from the steam generators.

Footnote G-1 Facilities licensed with, or applying for, alternative repair criteria (ARC) should use this section in conjunction with the guidance Conforms - FNP is licensed to ARC.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 61 Table F: Conformance With Regulatory Guide 1.183 Appendix G (Locked Rotor Accident) RG Section RG Position FNP Analysis that is being developed in Draft Regulatory Guide DG

-1074, "Steam Generator Tube Integrity" (USNRC, December 1998), for acceptable assumptions and methodologies for performing radiological analyses.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 62 Table G: Conformance With Regulatory Guide 1.183 Appendix H (Rod Ejection Accident

) RG Section RG Position FNP Analysis H-1 Assumptions acceptable to the NRC staff regarding core inventory are in Regulatory Position 3 of this guide. For the rod ejection accident, the release from the breached fuel is based on the estimate of the number of fuel rods breached and the assumption that 10% of the core inventory of the noble gases and iodines is in the fuel gap. The release attributed to fuel melting is based on the fraction of the fuel that reaches or exceeds the initiation temperature for fuel melting and the assumption that 100% of the noble gases and 25% of the iodines contained in that fraction are available for release from containment. For the secondary system release pathway, 100% of the noble gases and 50% of the iodines in that fraction are released to the reactor coolant.

Conforms - See discussions in Table A. The fission product release is based upon Appendix H the amount of damaged fuel and the assumption that 10% of the core inventory of noble gases and iodines are in the fuel rod gap.

For releases from containment involve fuel melting, 100% of the noble gases and 50 % of the iodines contained in the portion of the fuel which melts is available for release from containme nt and to the RCS for the secondary release pathway. H-2 If no fuel damage is postulated for the limiting event, a radiological analysis is not required as the consequences of this event are bounded by the consequences projected for the loss-of-coolant accident (LOCA), main steam line break, and steam generator tube rupture.

Not Applicable

- Failed fuel is postulated for this event. H-3 Two release cases are to be considered. In the first, 100% of the activity released from the fuel should be assumed to be released instantaneously and homogeneously through the containment atmosphere. In the second, 100% of the activity released from the fuel should be assumed to be completely dissolved in the primary coolant and available for release to the secondary system. Conforms - Two release pathways are considered. In the release from containment, 100% of the activity from fuel melting and fuel cladding damage instantaneously reaches the containment at the onset of the accident and is available for release to the environment. In the case with the release from the secondary system, 100% of the activity from fuel melting and fuel cladding damage instantaneously reaches the RCS at the onset of the accident and is available for release to the secondary system and to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 63 Table G: Conformance With Regulatory Guide 1.183 Appendix H (Rod Ejection Accident

) RG Section RG Position FNP Analysis eventually to the environment.

H-4 The chemical form of radioiodine released to the containment atmosphere should be assumed to be 95% cesium iodide (CsI), 4.85% elemental iodine, and 0.15% organic iodide. If containment sprays do not actuate or are terminated prior to accumulating sump water, or if the containment sump pH is not controlled at values of 7 or greater, the iodine species should be evaluated on an individual case basis. Evaluations of pH should consider the effect of acids created during the rod ejection accident event, e.g., pyrolysis and radiolysis products. With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form. Conforms - The chemical form of radioiodine released to the containment atmosphere is assumed to be 95% cesium iodide, 4.85%

elemental iodine, and 0.15% organic iodide. Since containment sprays will not necessarily be activated in this event, no credit is taken for pH being controlled at values of 7 or greater. H-5 Iodine releases from the steam generators to the environment should be assumed to be 97% elemental and 3% organic.

Conforms - Accounting for particulate scrubbing in the steam generators, the release to the environment is 97% elemental and 3% organic.

H-6.1 A reduction in the amount of radioactive material available for leakage from the containment that is due to natural deposition, containment sprays, recirculating filter systems, dual containments, or other engineered safety features may be taken into account. Refer to Appendix A to this guide for guidance on acceptable methods and assumptions for evaluating these mechanisms.

Conforms - Radioactive material removal from the containment atmosphere by sprays and other engineered safety features is not credited. Natural deposition of elemental iodine is credited. H-6.2 The containment should be assumed to leak at the leak rate incorporated in the technical specifications at peak accident pressure for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and at 50% of this leak rate for the remaining duration of the accident. Peak accident pressure is the maximum pressure defined in the technical specifications Conforms - The containment is assumed to leak to the environment at the technical specification limit of 0.15%/day for the first 24 h ours of the accident and half this rate thereafter. to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 64 Table G: Conformance With Regulatory Guide 1.183 Appendix H (Rod Ejection Accident

) RG Section RG Position FNP Analysis for containment leak testing. Leakage from subatmospheric containments is assumed to be terminated when the containment is brought to a subatmospheric condition as defined in technical specifications.

H-7.1 A leak rate equivalent to the primary

-to-secondary leak rate limiting condition for operation specified in the technical specifications should be assumed to exist until shutdown cooling is in operation and releases from the steam generators have been terminated.

Conforms - The total leakage from the primary system to the secondary system is assumed to be 1 gpm, conservatively bounding the technical specification limit of 150 gpd per generator. This leakage lasts for the first 2500 sec of the accident and is conservatively modeled as being direct to the environment

. H-7.2 The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g., lbm/hr) should be consistent with the basis of surveillance tests used to show compliance with leak rate technical specifications. These tests typically are based on cooled liquid.

The facility's instrumentation used to determine leakage typically is located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft 3). Conforms - The water density of both the primary and secondary coolants is assumed to be 62.4 lbm/ft

3. H-7.3 All noble gas radionuclides released to the secondary system are assumed to be released to the environment without reduction or mitigation.

Conforms - It is assumed that noble gases are not retained in the secondary water. H-7.4 The transport model described in assumptions 5.5 and 5.6 of Appendix E should be utilized for iodine and particulates.

Conforms - The transport model described in Position 5.5 and 5.6 of Appendix E is applied to releases from the steam generators.

Footnote H-1 Facilities licensed with, or applying for, alternative repair criteria (ARC) should use this section in conjunction with the guidance that is being developed in Draft Regulatory Guide DG

-1074, Conforms - FNP is licensed to ARC.

to NL 0388 Regulatory Guide 1.183 Conformance Tables E5 - 65 Table G: Conformance With Regulatory Guide 1.183 Appendix H (Rod Ejection Accident

) RG Section RG Position FNP Analysis "Steam Generator Tube Integrity" (USNRC, December 1998), for acceptable assumptions and methodologies for performing radiological analyses.

E5 - 66 Joseph M. Farley Nuclear Plant - Units 1 and 2 Alternative Source Term License Amendment Request Enclosure 6 Loss of Coolant Accident Analysis to NL 0388 Loss of Coolant Accident Analysis E5 - 67 to NL 0388 Fuel Handling Accident Analysis E5 - 68

Joseph M. Farley Nuclear Plant - Units 1 and 2 Alternative Source Term License Amendment Request Enclosure 7 Fuel Handling Accident Analysis to NL 0388 Main Steam Line Break Analysis E5 - 69

Joseph M. Farley Nuclear Plant - Units 1 and 2 Alternative Source Term License Amendment Request Enclosure 8 Main Steam Line Break Analysis to NL 0388 Steam Generator Tube Rupture Analysis E5 - 70

Joseph M. Farley Nuclear Plant - Units 1 and 2 Alternative Source Term License Amendment Request Enclosure 9 Steam Generator Tube Rupture Analysis 0 to NL 0388 Control Rod Ejection Analysis E5 - 71

Joseph M. Farley Nuclear Plant - Units 1 and 2 Alternative Source Term License Amendment Request Enclosure 10 Control Rod Ejection Analysis

1 to NL 0388 Locked Rotor Analysis E5 - 72

Joseph M. Farley Nuclear Plant - Units 1 and 2 Alternative Source Term License Amendment Request Enclosure 11 Locked Rotor Analysis 2 to NL 0388 FNP AST Accident Analysis Input Values Comparison Tables E5 - 73

Joseph M. Farley Nuclear Plant - Units 1 and 2 Alternative Source Term License Amendment Request Enclosure 12 FNP AST Accident Analysis Input Values Comparison Tables 2 to NL 0388 FNP AST Accident Analysis Input Values Comparison Tables E5 - 74 FNP AST Accident Analysis Input Values Comparison Tables To facilitate the review and to more readily assess the impact of the adoption of the Alternative Source Term at Farley Nuclear Plant, summary tables are provided in this enclosure for each accident being analyzed including a comparison between current licensing basis (CLB) input parameters and the values utilized in the new AST accident analysis, and the basis for any changes. The tables are provided within this enclosure for the following accident scenarios:

Table 2 - Loss of Coolant Accident (LOCA)

Table 3 - Fuel Handling Accident (FHA)

Table 4 - Main Steam Line Break (MSLB)

Table 5 - Steam Generator Tube Rupture (SGTR)

Table 6 - Locked Rotor Accident (LRA)

Table 7 - Control Rod Ejection (CRE)

Additionally, Table 1, "Control Room Parameters," is provided to show the parameters of interest for control room habitability (CRH).

In this table, the LOCA parameters are provided as that resulted in the most limiting dose to the Control Room occupants.

2 to NL 0388 FNP AST Accident Analysis Input Values Comparison Tables E5 - 75 Table 1: Control Room Parameters Input/Assumption CLB CRH Value New AST Value Reason for Change Control Room Volume 114,000 ft 3 114,ooo ft 3 No change Normal Operation Filtered Make

-up Flow Rate 0 cfm 0 cfm No change Filtered Recirculation Flow Rate 0 cfm 0 cfm No change Unfiltered Make

-up Flow Rate 0 cfm 2340 cfm 60 seconds of normal Control Room HVAC operation is assumed after accident initiation.

Unfiltered Inleakage 0 cfm 0 cfm No change Emergency Operation Recirculation Mode Filtered Make

-up Flow Rate 375 cfm 375 cfm No change Filtered Recirculation Flow Rate 2700 cfm 2700 cfm No change Unfiltered Make

-up Flow Rate 0 cfm 0 cfm No change Unfiltered Inleakage 53 cfm 325 cfm The revised value is intended to provide operational margin to the control room measured control room inleakage.

Filter Efficiencies Elemental 94.5% 94.5% No change Organic 94.5% 94.5% No change Particulate 98.5% 98.5% No change Occupancy 0-24 hours 1-4 days 4-30 days 100% 60% 40% 100% 60% 40% No change Breathing Rate 3.47E-4 m 3/sec 3.5E-4 m 3/sec Rounded up for conservatism.

2 to NL 0388 FNP AST Accident Analysis Input Values Comparison Tables E5 - 76 Table 2: LOCA Inputs and Assumptions Input/Assumption CLB Offsite Value CLB CRH AST Value New AST Value For Offsite and CRH Reason for Change Containment Purge Iodine Chemical Form 2.5% particulate, 95.5% elemental, 2.0% organic 2.5% particulate, 95.5% elemental, 2.0% organic 95% cesium iodide, 4.85% elemental , 0.15% organic Adoption of RG 1.183 methodology.

Containment Volume 2,030,000 ft 3 2,030,000 ft 3 2,030,000 ft 3 No change Containment Purge Filtration 0% 0% 0% No change Removal by Wall Deposition None None None No change Removal by Sprays None None None No change Containment Leakage Iodine Chemical Form 2.5% particulate, 95.5% elemental, 2.0% organic 2.5% particulate, 95.5% elemental, 2.0% organic 95% cesium iodide, 4.85% elemental , 0.15% organic Adoption of RG 1.183 methodology.

Containment Sump pH

>7.0 >7.0 >7.0 No change Containment Sprayed Volume 1,668,660 ft 3 1,668,660 ft 3 1,668,660 ft 3 No change Containment unsprayed Volume 361,340 ft 3 361,340 ft 3 361,340 ft 3 No change Containment Spray Start Time 0 seconds 0 seconds 90 seconds Provides additional conservatism to Containment Leakage Pathway.

Containment Spray Stop Time 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months No change Containment Spray Flow Rate 2480 gpm injection mode 2480 gpm injection mode 2480 gpm injection mode 2290 gpm recirculation mode No change Elemental Iodine Spray Removal Coefficient 2.7 hr-1 2.7 hr-1 13.7 hr-1 Revision is consistent with RG 1.183 Appendix A RP 3.3 Aerosol Spray Removal Coefficient 5.45 hr-1injection mode 5.03 hr-1recirculation mode 5.45 hr-1injection mode 5.03 hr-1recirculation mode 5.45 hr-1injection mode 5.03 hr-1recirculation mode No change Organic Iodine Spray Removal None None None No change Natural Deposition Elemental, Organic, Aerosol

- None Elemental, Organic, Aerosol

- None Elemental, Organic iodine

- None Aerosols - 0.1 hr-1in unsprayed regions only Aerosol natural deposition is permitted per Appendix A of RG 1.183. Containment Leakage Rate 0 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 hours to 30 days

0.15%/day 0.075%/day

0.15%/day 0.075%/day No change Containment Leakage Filtration 0% 0% 0% No change ESF Leakage to the Auxiliary Building Iodine Chemical Form 0% aerosol, 98% elemental, 2% organic 0% aerosol, 98% elemental, 2% organic 0% aerosol, 97% elemental, 3% organic The revised percentages are as specified in RG 1.183.

Containment Sump Volume 49,200 ft 3 49,200 ft 3 49,200 ft 3 No change ECCS Recirculation Start Time 20 minutes 20 minutes 20 minutes No change 2 to NL 0388 FNP AST Accident Analysis Input Values Comparison Tables E5 - 77 Table 2: LOCA Inputs and Assumptions Input/Assumption CLB Offsite Value CLB CRH AST Value New AST Value For Offsite and CRH Reason for Change ESF Leakage Flow Rate 4,000 cc/hr 6,000 cc/hr 20,000 cc/hr ESF flow rate increased to provide additional operating margin in the analysis. ESF Flashing Fraction 15% 15% 10% ESF flashing fraction recalculated based on Section 5.4 of RG 1.183. ESF Leakage to the RWST (Not Explicitly Modeled in the CLB)

Table 3: FHA Inputs and Assumptions Input/Assumption CLB Offsite Value CLB CRH AST Value New AST Value For Offsite and CRH Reason for Change Iodine Chemical Form 0% aerosol, 99.75% elemental, 0.25% organic 0% aerosol, 99.75% elemental, 0.25% organic 0% aerosol, 99.85% elemental, 0.15% organic Chemical composition is as described in RG 1.183 Appendix B Section 2.

Number of Fuel Assemblies Damaged 1 1 1 No change Percentage of Fuel Rods Failed 100% 100% 100% No change No. of rods exceeding 6.3 kw/ft above 54 GWD/MTU 0 0 0 No change Water Level Above Damaged Fuel 23 ft 23 ft 23 No change Pool Decontamination Factors Elementary

- 400 Organic - 1 Elementary

- 500 Organic - 1 Decontamination Factors are as described in RG 1.183 Appendix B Section 2.

Delay Before Fuel Movement 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months 100 hours 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months No change Containment Release Filtration 0% 0% 0% No change Table 4: MSLB Inputs and Assumptions Input/Assumption CLB Offsite Value CLB CRH AST Value New AST Value For Offsite and CRH Reason for Change Maximum Pre

-Accident Iodine Spike Concentration

-131 -131 I-131 No change Concurrent Iodine Spike Appearance Rate No change Initial Steam Generator Iodine No change 2 to NL 0388 FNP AST Accident Analysis Input Values Comparison Tables E5 - 78 Table 4: MSLB Inputs and Assumptions Input/Assumption CLB Offsite Value CLB CRH AST Value New AST Value For Offsite and CRH Reason for Change Source Term Iodine Chemical Form Not provided Not provided 0% aerosol, 97% elemental, 3% organic The AST chemical form is as provided in RG 1.183 Appendix E, Section 4.

Percentage of Fuel Rods Failed 0% 0% 1% No change. Note

- the MSLB does not result in failed fuel. This is a pre

-condition included for conservatism.

RCS Mass 440,000 lbm 440,000 lbm 440,900 lbm AST calculation more precisely accounts for CVCS mass.

Mass 168,000 lbm/SG 168,000 lbm/SG 168,000 lbm/SG No change Intact Steam Generator Steam Release 0 - 2 hrs: 316,715 lbm 2 - 8 hrs: 703,689 lbm 8 - 24 hrs: 948,000 lbm 0 - 2 hrs: 316,715 lbm 2 - 8 hrs: 703,689 lbm 8 - 24 hrs: 948,000 lbm 0 - 2 hrs: 316,715 lbm 2 - 8 hrs: 703,689 lbm 8 - 24 hrs: 948,000 lbm No change Primary-Secondary Leak Rate 0.65 gpm to two intact SGs 0.35 gpm to faulted SG 0.65 gpm to two intact SGs 0.35 gpm to faulted SG 0.65 gpm to two intact SGs 0.35 gpm to faulted SG No change Density Used for Leakage Volume

-to-Mass Conversion 62.4 lbm/ft 3 62.4 lbm/ft 3 62.4 lbm/ft 3 No change Duration of Intact SG Tube Uncovery After Reactor Trip Not modeled Not modeled Not modeled Tube uncover does not occur with intact SGs 24 hrs 24 hrs 24 hrs No change Intact Steam Generator Iodine partition factor 10 10 100 RG 1.183 Appendix E Section 5.5.4 allows an iodine partition factor of 100 for the intact SG.

Intact Steam Generator Moisture Carryover Fraction Not modeled Not modeled 0.1% (Alkali Metal Partition Factor =1000) New AST value conservatively included. Table 5: SGTR Inputs and Assumptions Input/Assumption CLB Offsite Value CLB CRH AST Value New AST Value For Offsite and CRH Reason for Change Maximum Pre

-Accident Iodine Spike Concentration

-131 -131 -131 No change Concurrent Iodine Spike Appearance Rate RG 1.183 Appendix F Section 2.2 allows the 335 factor.

Initial Steam Generator Iodine -131 -131 -131 No change 2 to NL 0388 FNP AST Accident Analysis Input Values Comparison Tables E5 - 79 Table 5: SGTR Inputs and Assumptions Input/Assumption CLB Offsite Value CLB CRH AST Value New AST Value For Offsite and CRH Reason for Change Source Term Iodine Chemical Form Not provided Not provided 0% aerosol, 97% elemental, 3% organic Iodine chemical form is per RG 1.183 Appendix F Section 4.

Percentage of Fuel Rods Failed 0% 0% 1% No change. Note

- the SGTR does not result in failed fuel. This is a pre

-condition included for conservatism.

RCS Mass 441,000 lbm 441,000 lbm 440,900 lbm AST calculation more precisely accounts for CVCS mass.

Mass 105,000 lbm/SG 105,000 lbm/SG 105,000 lbm/SG No change Intact Steam Generator Steam Release 0 - 0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months

- 1133 lbm/second 0.09 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months - 422,000 lbm 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months - 934,000 lbm 0 - 0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months

- 1133 lbm/second 0.09 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months - 422,000 lbm 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months - 934,000 lbm No change Ruptured Steam Generator Steam Release 0 - 0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months

- 1133 lbm/second 0.09 - 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

- 79,000 lbm 0 - 0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months

- 1133 lbm/second 0.09 - 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

- 79,000 lbm 0 - 0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months

- 1133 lbm/second 0.09 - 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

- 79,000 lbm No change Feedwater Flow to Intact Steam Generators 0.09 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months - 327,000 lbm 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months - 981,000 lbm 0.09 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months - 327,000 lbm 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months - 981,000 lbm 0.09 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months - 327,000 lbm 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months - 981,000 lbm No change Time of Reactor Trip 0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months 0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months 0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months No change Primary-Secondary Leak Rate 1 gpm 1 gpm 1 gpm No change Density Used for Leakage Volume

-to-Mass Conversion 62.4 lbm/ft 3 62.4 lbm/ft 3 62.4 lbm/ft 3 No change Ruptured Tube Break Flow 0 - 0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months

- 21,600 lbm 0.09 - 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

- 79,000 lbm 0 - 0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months

- 21,600 lbm 0.09 - 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months - 79,000 lbm 0 - 0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months

- 21,600 lbm 0.09 - 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

- 79,000 lbm No change Duration of Ruptured Tube Break Flow 30 minutes 30 minutes 30 minutes No change Break Flow Flashing Fraction 0 - 0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months

- 21% 0.09 - 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

- 15% 0 - 0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months

- 21% 0.09 - 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

- 15% 0 - 0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months

- 21% 0.09 - 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

- 15% No change Duration of Intact SG Tube Uncovery After Reactor Trip 0 minutes 0 minutes 0 minutes No change 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months No change Intact Steam Generator Iodine Partition Coefficient 100 100 100 No change Intact Steam Generator Moisture Carryover Fraction Not provided Not provided 0.1% Carryover is provided for per RG 1.183 Appendix F Section 5.6.

2 to NL 0388 FNP AST Accident Analysis Input Values Comparison Tables E5 - 80 Table 6: CRE Inputs and Assumptions Input/Assumption CLB Offsite Value CLB CRH AST Value New AST Value For Offsite and CRH Reason for Change Fuel Rod Gap Fractions Iodine - 12% Kr 85- 10% Iodine - 12% Kr 85- 10% Iodines and noble gases

- 0.10 Other halogens - 0.05 Alkali metals

- 0.12 AST gap fractions are per RG 1.183 Table 3 Fuel Rod Peaking Factor Not provided Not provided 1.7 Radial peaking factor is applied per RG 1.183 Section 3.1.

Percentage of Fuel Rods Failed 10% 10% 10% No change Percentage of Fuel That Experiences Melting 0.25% 0.25% 0.25% No change Number of rods exceeding 6.3 kw/ft above 54 GWD/MTU 0 0 0 No change Initial Steam Generator Iodine Source Term No change Iodine Chemical Form

- Secondary Release Not Provided Not provided 0% aerosol, 97% elemental, 3% organic Iodine chemical form is in accordance with RG 1.183 Appendix H Section 5 Iodine Chemical Form

- Containment Release Not provided Not provided 95% aerosol, 4.85% elemental, 0.15% organic Iodine chemical form is in accordance with RG 1.183 Appendix H Section 4 Containment Volume 2.03E6 ft 3 2.03E6 ft 3 2.03E6 ft 3 No change Containment Leakage Rate 0 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 hours to 30 days

0.15% 0.075%

0.15% 0.075% No change Containment Leakage Filtration 0% 0% 0% No change Natural Deposition in Containment 50% plateout of RCS release 50% plateout of RCS release Elemental iodine

- None Aerosols - 2.74E-2 hr-1 Natural deposition is credited per RG 1.183 Appendix H Section 6.1.

Iodine/Particulate Removal by Containment Sprays Not provided Not provided 5.0 hr-1 Particulate removal by Containment Spray is per RG 1.183 Appendix H Section 6.1.

RCS Mass 435,000 lbm 435,000 lbm 440,900 lbm The AST calculation more accurately cites the RCS mass.

Steam Generator Mass 168,000 lbm 168,000 lbm 168,000 lbm No change Primary-Secondary Leak Rate 150 gallons per day per SG 150 gallons per day per SG 1 gpm total AST value increased for additional conservatism.

2 to NL 0388 FNP AST Accident Analysis Input Values Comparison Tables E5 - 81 Table 6: CRE Inputs and Assumptions Input/Assumption CLB Offsite Value CLB CRH AST Value New AST Value For Offsite and CRH Reason for Change Density Used for Leakage Volume

-to-Mass Conversion 62.4 lbm/ft 3 62.4 lbm/ft 3 62.4 lbm/ft 3 No change Secondary Steam Release 426,000 lbm 426,000 lbm 426,000 lbm No change Time until Primary and Secondary 2500 seconds 2500 seconds 2500 seconds No change Duration of SG Tube Uncovery Following Reactor Trip 0 minutes 0 minutes 0 minutes No change Steam Generator Iodine Partition Coefficient 10 10 100 RG 1.183 Appendix H Section 7.4 allows an iodine partition factor of 100 for SG releases.

Steam Generator Moisture Carryover Fraction Not provided Not provided 0.1% Carryover is provided for per RG 1.183 Appendix H Section 7.4

. Table 7: LRA Inputs and Assumptions Input/Assumption CLB Offsite Value CLB CRH AST Value New AST Value For Offsite and CRH Reason for Change Fuel Rod Gap Fractions Iodine - 12% Iodine - 12% I-131 - 0.08 Kr 0.10 Other Halogens and Noble Gases - 0.05 Alkali Metals

- 0.12 AST gap fractions are per RG 1.183 Table 3 Fuel Rod Peaking Factor Not provided Not provided 1.7 Radial peaking factor is applied per RG 1.183 Section 3.1.

Number of rods exceeding 6.3 kw/ft above 54 GWD/MTU 0 0 0 0 Initial Steam Generator Iodine Source Term No change Iodine Chemical Form Not provided Not provided 95% particulate 4.85% elemental 0.15% organic Iodine chemical form is in accordance with RG 1.183 Appendix G Section 5.6.

RCS Mass 441,000 lbm 441,000 lbm 440,900 lbm AST calculation more precisely accounts for CVCS mass.

Mass 168,000 lbm 168,000 lbm 168,000 lbm No change 2 to NL 0388 FNP AST Accident Analysis Input Values Comparison Tables E5 - 82 Table 7: LRA Inputs and Assumptions Input/Assumption CLB Offsite Value CLB CRH AST Value New AST Value For Offsite and CRH Reason for Change Primary-Secondary Leak Rate 150 gallons per day per SG 150 gallons per day per SG 1 gpm total AST value increased for additional conservatism.

Density Used for Leakage Volume

-to-Mass Conversion 62.4 lbm/ft 3 62.4 lbm/ft 3 62.4 lbm/ft 3 No change Secondary Steam Release 426,000 lbm 426,000 lbm 426,000 lbm No change Time until Primary and Secondary 2500 seconds 2500 seconds 2500 seconds No change Duration of SG Tube Uncovery Following Reactor Trip 0 minutes 0 minutes 0 minutes No change Steam Generator Iodine Partition Coefficient 10 10 100 RG 1.183 Appendix G Section 5.6 allows an iodine partition factor of 100 for SG releases.

Intact Steam Generator Moisture Carryover Fraction Not provided Not provided 0.1% Carryover is provided for per RG 1.183 Appendix G Section 5.6.

NL-16-0388, Draft License Amendment Request Regarding Alternative Source Team

Text

Enclosures:

6.0 References

3.0 TECHNICAL EVALUATION

3.9 Conclusions

4.4 Conclusions

6.0 References

Attachment:

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