Text

Alternative Source Term License Amendment Request
	ML16336A024
Person / Time
Site:	Farley
Issue date:	11/22/2016
From:	Pierce C; Southern Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NL-16-0388
	Download: ML16336A024 (232)
	v • d • e

4 Southern Nuclear Charles R. Pierce Regulatory Affairs Director 40 Inverness Center Parkway*.

Post Office Box 1295 Birmingham, AL 35242 .

205 992 7872 tel 205 992 7601 fax crpierce@southernco.com November 22, 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 (NRG) review and approval of proposed revisions to the licensing basis of FNP that support a full scope application of an Alternative Source Term (AST) methodology. Proposed TS changes, which are supported by the AST Design Basis Accident 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-A, "Control Room Habitability," Revision 3, and TSTF-312-A, Administrative Control of Containment Penetrations," Revision 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 ttie TS Bases (for information). provides the clean, retyped Operating License and TS pages. Enclosures 5 through 12 provide additional information in support of this LAR. Enclosure 13 provides the RADTRAD Input/Output files for the Loss-of-Coolant Accident and Fuel Handling Accident in CD format, as well as the ARCON96 input files used to develop the Reactor Water Storage Tank atmospheric dispersion factors.

SNC requests approval of the proposed LAR by November 30, 2017. 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 NRG commitments, as stated in Enclosure 14. If you have any questions, please contact Ken McElroy at 205.992.7369. O(

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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.

Res{~uf);;:,ed,

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C.R. Pierce ; "'*

Regulatory Affairs Director ,.. .

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Sworn to and subset'.* ed before me this J;l. day of A)~ , 2016.

My commission expires: /D *8- 201-1 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 Accident Analysis

9. Steam Generator Tube Rupture Accident Analysis

10. Control Rod Ejection Accident Analysis 11 . Locked Rotor Accident Analysis 12 .. FNP AST Accident Analysis Input Values Comparison Tables

13. FNP AST LAR Supporting Information

14. Summary of Regulatory Commitments 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

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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 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-A 2.3 TSTF-312-A 3.0 Technical Evaluation 3.1 Meteorology and Atmospheric Dispersion 3.2 Analytical Models 3.3 Loss of Coolant Accident 3.4 Fuel Handling Accident 3.5 Main Steam Line Break Accident 3.6 Steam Generator Tube Rupture Accident 3.7 Control Rod Ejection Accident 3.8 Locked Rotor Accident 3.9 Conclusions 3.1 O 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 1.0 Summary Description This evaluation supports a request to revise Operating License NPF-2 and NFP-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 support a selected scope application of an Alternative Source Term (AST) methodology. The proposed amendment also incorporates Revision 3 of Technical Specification Task Force (TSTF)

Traveler, TSTF-448-A, "Control Room Habitability," as well as TSTF-312-A, "Administrative Control of Containment Penetrations," into the FNP Technical Specifications (TS).

This application is made to improve two key parameters for which FNP currently has low margin: 1) Emergency Core Cooling System (ECCS) leakage outside containment, and

2) unfiltered in-leakage for the control room. This application will also extend the time for the critical operator action of initiating the pressurization and recirculation mode for the Control Room Emergency Filtration/Pressurization System (CREFS) in a Fuel Handling Accident (FHA).

2.0 Detailed Description 2.1 Background In December 1999, the NRC issued a new regulation, 1 O 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 (OBA) 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 Room (CR) radiological consequences of the FNP design basis accidents. This reanalysis involves several changes in selected analysis assumptions. 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 1p 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, General Design Criteria (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 lndustry/TSTF TS change TSTF-448 Revision 3. The availability of this TS improvement was published in the Federal Register on E1 - 1 to NL-16-0388 Basis for Proposed Change January 17, 2007 as part of the consolidated line item improvement process (CLllP) (Reference 2).

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

2 .3 TSTF-312 The proposed TS change adds a Note to the LCO for TS 3.9.3, "Containment Penetrations," allowing penetration flow path(s) that have direct access from the containment atmosphere to the outside atmosphere to be unisolated under administrative control.

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 the AST LOCA analysis, new and revised X/Q values for the CR have been developed to address potential leakage from the Reactor Water Storage Tank (RWST) vent and from the containment mini-purge system. The values resulting at the CR intakes are calculated using the NRG-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). Information used to develop the new X/Q values is included in Enclosure 13.

3.1.1 Meteorological Data In the AST analyses, two sets of meteorological data were used. The first was a continuous temporally representative four year period of hourly average data from the FNP meteorological tower (January 1, 2000, through December 31, 2003). The second data set included the same 2000 - 2003 information, but also contained an additional six months of derived data from 1999, resulting in a collection span of 4 Y2 years. Both sets of data were selected as they have been previously reviewed by the NRC as being of good quality (Reference 5). In Reference 3, the NRC concluded that X/Qs developed from either data set were acceptable if they produced the more limiting result. As shown in Tables 3.4 and 3.5, the 4-year data set generally provided more conservative X/Q results.

E1 - 2 to* NL-16-0388 Basis for Proposed Change 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 References 3 and 5.

The X/Q values for the EAB and the LPZ used in the radiological consequence analyses are shown in Table 3.1.

Table 3.1 - EAB and LPZ X/Q values (sec/m 3)

Location Time Period X/Q Value EAB 0-2 hours 7.6x10-4 LPZ o -2 hours 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 hours 1.ox10-0 24-96 hours 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 %-

year meteorological data described previously. New X/Q values were developed for the Containment mini-purge and RWST release points for the Loss-of-Coolant Accident (LOCA) based on the 4-year data set.

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 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 RG 1.194 Position 3.2.1.

Tables 3.2 and 3.3 provide information related to the relative elevations of the release-receptor combinations, the straight-line horizontal distance E1 - 3 to NL-16-0388 Basis for Proposed Change 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.

Figure 3.1 - Air Intake Locations and Release Points t

Unit 2 RWST G

. 9 e 10 8

1 Unitl RWST G

12 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 Control Room Emergency Intake 9 Unit 2 Control Room Emergency Intake 10 Normal Control Room Intake 11 Unit 2 Hatch Door 12 Unit 1 Hatch Door E1 - 4 to NL-16-0388 Basis for Proposed Change Table 3.2 - Distance and Geometry of Release and Receptor Locations Release Horizontal Vertical Horizontal Vertical Direction Receptor Point Distance (ft) Distance (ft) Distance (m) Distance (m) to Source Unit 1 Control Unit 1 Room Emergency RWST 372 3.5 113 1.07 101 Air Intake Unit 1 Control Unit2 Room Emergency RWST 114 1.07 78 373 3.5 Air Intake Unit 2 Control Unit 1 Room Emergency RWST 373.27 3.5 114 1.07 102 Air Intake Unit 2 Control Unit2 Room Emergency RWST 372.38 3.5 113.5 1.07 79 Air Intake Control Room Unit 1 Normal Air Intake Vent Reactor

. 154.04 121.5 46.9 37.0 135 111.46 0.0 33.9 0.0 159 Control Room Unit2 Normal Air Intake Vent Reactor

. 121.73 121.5 37.1 37.0 64 61.61 0.0 18.7 0.0 30

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

Table 3.3 - Elevations of Control Room Air Intakes and Release Points Location Elevation (ft) Elevation (m) Height above grade (ft) Height above grade (m)

Emergency Intake 192 58.5 37.5 11.4 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 provides the ARCON96 modeling outputs for releases originating at the reactor buildings, reactor vents, hatch doors, and RWST. These Control Room X/Q factors for LOCA, Main Steam Line Break (MSLB), Steam Generator Tube Rupture (SGTR}, Control Rod Ejection, and Locked Rotor Accidents are based on 4 years of meteorological data from years 2000 to 2003, as this data set provided the most limiting values. Table 3.4b provides the X/Q values based on 4 'Y2 years of data from 1999 to 2003 which was used in the FHA. These atmospheric dispersion factors were previously reviewed by the NRG in References 3 and 5.

E1 - 5 to NL-16-0388 Basis for Proposed Change Table 3.4a - X/Q Values at the Control Room Air Intakes (4 years meteorological data)

Release Point Receptor O - 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 8-24hours 1- 4days 4-30 days U1 Vent U1XCR 1.61 E-03 1.26E-03 5.65E-04 3.43E-04 2.32E-04 U1 Reactor U1 XCR 1.51 E-03 9.01 E-04 3.95E-04 2.75E-04 1.91 E-04 U1 Hatch Door U1XCR 8.39E-04 5.11 E-04 2.09E-04 1.47E-04 9.35E-05 U1 RWST U1XCR 4.97E-04 3.69E-04 1.53E-04 1.15E~04 7.98E-05 U2 Vent U1XCR 1.64E-03 1.37E-03 7.17E-04 5.41 E-04 3.60E-04 U2 Reactor U1 XCR 1.66E-03 1.36E-03 6.81 E-04 5.60E-04 4.21 E-04 U2 Hatch Door U1 XCR 7.95E-04 6.73E-04 3.35E-04 2.48E-04 1.87E-04 U2 RWST U1XCR 4.80E-04 3.82E-04 1.?0E-04 1.28E-04 9.98E-05 U1 Vent U2XCR 1.59E-03 1.25E-03 5.54E-04 3.34E-04 2.27E-04 U1 Reactor U2XCR 1.51 E-03 8.91 E-04 3.91 E-04 2.71 E-04 1.87E-04 U1 Hatch Door U2XCR 8.04E-04 4.90E-04 2.01 E-04 1.39E-04 9.04E-05 U1 RWST U2XCR 4.95E-04 3.66E-04 1.52E-04 1.14E-04 7.90E-05 U2Vent U2XCR 1.65E-03 1.38E-03 7.20E-04 5.47E-04 3.63E-04 U2 Reactor U2XCR 1.65E-03 1.34E-03 6.75E-04 5.40E-04 4.03E-04 U2 Hatch Door U2XCR 8.33E-04 6.98E-04 3.43E-04 2.57E-04 1.91 E-04 U2 RWST U2XCR 4.82E-04 3.82E-04 1.?0E-04 1.28E-04 1.00E-04 U1 Vent Normal CR 2.01 E-03 1.46E-03 6.0?E-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 3.11 E-03 1.38E-03 1.29E-03 1.04E-03 Table 3.4b - 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 8-24hours 1- 4days 4-30 days U1 Vent U1XCR 1.62E-03 1.21 E-03 5.37E-04 3.35E-04 2.32E-04 U1 Reactor U1 XCR 1.54E-03 9.62E-04 4.45E-04 2.61E-04 3.09E-04 U1 Hatch Door U1XCR 8.79E-04 5.89E-04 2.63E-04 1.57E-04 1.93E-04 U2 Vent U1XCR 1.59E-03 1.35E-03 7.08E-04 5.16E-04 3.59E-04 U2 Reactor U1 XCR 1.64E-03 1.34E-03 6.65E-04 5.45E-04 4.18E-04 U2 Hatch Door U1 XCR 7.83E-04 6.52E-04 3.22E-04 2.43E-04 1.85E-04 U1 Vent U2XCR 1.60E-03 1.19E-03 5.23E-04 3.24E-04 2.28E-04 U1 Reactor U2XCR 1.54E-03 9.50E-04 4.43E-04 2.?0E-04 3.11 E-04 U1 Hatch Door U2XCR 8.56E-04 5.67E-04 2.50E-04 1.52E-04 1.92E-04 U2Vent U2XCR 1.60E-03 1.37E-03 7.10E-04 5.21E-04 3.60E-04 U2 Reactor U2XCR 1.64E-03 1.32E-03 6.60E-04 5.27E-04 4.01E-04 U2 Hatch Door U2XCR 8.18E-04 6.77E-04 3.32E-04 2.49E-04 1.89E-04 Notes: U1 - Unit 1 U2 -Unit 2 XCR-EmergencyControl Room Intake Normal CR - Non-emergency Control Room Intake E1 - 6 to NL-16-0388 Basis for Proposed Change 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, Control Rod Ejection, and Locked Rotor Accidents using source term and X/Q inputs. This proprietary Bechtel software calculates radioactive isotope activities within 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 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)

Emergency Core Cooling System (ECCS) 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.

E1 - 7 to NL-16-0388 Basis for Proposed Change 3.3.2 Radiological Dose Models The RADTRAD (Version 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 3.5a. 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.

Table 3.Sa - Parameters and Assumptions for the LOCA Parameter ECCS Leakage Initiation Time 20 minutes ECCS 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: 0% to 13.9%

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 Control Room 0-2 7.6E-4 2.80E-4 1.66E-03 2-8 1.10E-4 1.36E-03 8-24 1.00E-5 6.81 E-04 24-96 5.40E-6 5.60E-04 96-720 2;90E-6 4.21 E-04 Plant Vent:

Time (hr) EAB LPZ Control Room 0- 0.0167 7.6E-4 2.80E-4 2.79E-03 0.0167-2 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 E1 - 8 to NL-16-0388 Basis for Proposed Change RWST:

Time (hr) EAB LPZ Control Room 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 Control Room Parameters Parameter Volume 114,000 ft3 Ventilation System Normal Flow Rate 2340 cfm < 60 seconds Ventilation System Makeup Rate 375 cfm > 60 seconds Ventilation System Recirculation Flow Rate 2700 cfm > 60 seconds Ventilation System Charcoal Filter Efficiencies Pressurization Filters 98.5% all iodine species Recirculation Filters 94.5% elemental and organic 98.5% particulate Unfiltered In-leakage 325 cfm (includes 1O cfm for CR ingress/egress)

Breathing Rate 3.5E-4 m3/sec Occupancy Factors 0-24 hours 1.0 1 - 4 days 0.6 4-30 days 0.4 Table 3.Sb - Calculated LOCA Radiological Consequences TEDE (rem)

EAB LPZ Control Room Calculated results 13.2 6.0 4.7 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 area 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 from containment: one through the open Equipment Hatch directly to the environment and one through the open Personnel Airlock (PAL) to the Auxiliary Building and the Vent Stack. The release through the open Equipment Hatch is directly to the E1 - 9

_J to NL-16-0388 Basis for Proposed Change 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. Radioactivity released through the PAL mixes in a portion of the Auxiliary Building on the same level as the CR. Mixing in the Auxiliary Building volume is assumed to be instantaneous. This mixing establishes a pathway for the 1O CFM unfiltered ingress/egress CR in-leakage.

Upon detection of the FHA in containment, containment is evacuated and the penetrations open to containment are promptly closed.

Note that the release from containment to the Auxiliary Building through the PAL bounds the release through other containment penetrations into the Auxiliary Building. This is because the other penetrations are on elevations below the control room, and releases through those penetrations would have a tortuous path (through additional mixing volumes and up stairwells) to the area around the CR. Normal auxiliary building heating ventilating and air conditioning (HVAC) systems, which may be running at the time of the FHA, which ventilate the areas containing these other penetrations, and which might not be turned off in the course of the accident, exhaust to the normal auxiliary building plume. This plume is vented to the plant vent, and not to areas around the CR.

High radiation in the CR makeup air intake results in isolation of the CR.

Consistent with the FNP current licensing basis, manual operator action is taken to initiate the CREFS pressurization mode. For conservatism, this manual action was extended from 1O minutes to 20 minutes in the FHA analysis.

Releases which pass through the PAL can contaminate the Auxiliary Building at the level of the CR. Doses to operators from the ingress/egress of the CR through the door into the contaminated area have been evaluated and are included in the dose results.

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

The FHA 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 dose 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 Reactor power 2,831 MWt Fraction of Fission Product Inventory in Gap 1-131 0.08 Kr -85 0.10 Other Noble Gases 0.05 Other Halogens 0.05 E1 - 10 to NL-16-0388 Basis for Proposed Change 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 (OF) for Iodine 200 Chemical Form of Iodine Released from Pool Water Elemental 57%

Organic 43%

OF of Noble Gas 1 Duration of Release 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 Control Room 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 Control Room Parameters Parameter Volume 114,000 ft3 Ventilation System Makeup Rate 375 cfm Ventilation System Recirculation Flow Rate 2700 cfm Ventilation System Charcoal Filter Efficiencies Pressurization Filters 98.5% all iodine species Recirculation Filters 94.5% elemental and organic Unfiltered lnleakage 600 cfm (Isolation Mode) 325 cfm* (Pressurization Mode)

Breathing Rate 3.5E-4 m 3/sec Occupancy Factors 0-24 hours 1.0

For FHA in containment, 1O cfm for CR ingress/egress is modeled through the CR door from the contaminated auxiliary building. For FHA in the Spent Fuel Pool (SFP) area, 1O cfm for CR ingress/egress is included in the 325 cfm shown above.

Table 3.6b - Calculated FHA Radiological Consequences TEDE (rem)

EAB LPZ Control Room Calculated results 2.4 0.9 1.0 Dose acceptance criteria 6.3 6.3 5 E1 - 11 to NL-16-0388 Basis for Proposed Change 3.5 Main Steam Line Break Accident This event consists of a break in one main steam line outside of containment in which the faulted SG completely depressurizes and instantly releases the initial contents of the faulted 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 (1.0 gpm total). 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

µCi/gm. For the concurrent spike case, the initial primary iodine activity concentration is at the equilibrium TS value of 0.5 µCi/g Dose Equivalent Iodine.

This concurrent 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 as an initial condition, RCS activity includes consideration of 1% failed (leaking) fuel, consistent with the FNP current licensing basis.

Leakage from the 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 factor 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 Accident Parameter Steam Releases from Intact SG to Environment O - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 316,715 lbm 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 703,687 lbm 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 948,000 lbm E1 - 12 to NL-16-0388 Basis for Proposed Change Intact SG Liquid Iodine Partition Coefficient 100 Steam mass released from faulted SG to the Environment 439,145 lbm Faulted SG Dryout Time 322.8 seconds Atmospheric Dispersion Factors (sec/m 3 )

Time (hr) EAB LPZ Control Room 0-2 7.6E-4 2.80E-4 1.66E-3 2-8 1.1 OE-4 1.38E-3 8-24 1.00E-5 7.20E-4 Control Room Parameters Parameter Volume 114,000 ft 3 Ventilation System Makeup Rate 375 cfm Ventilation System Recirculation Flow Rate 2700 cfm Ventilation System Charcoal Filter Efficiencies Pressurization Filters 98.5% all iodine species Recirculation Filters 94.5% elemental and organic 98.5% particulate Unfiltered In-leakage 31 O cfm (includes 1O cfm CR ingress/egress)

Breathing Rate 3.5E-4 m3/sec Occupancy Factors 0-24 hours 1.0 1 - 4 days 0.6 4-30 days 0.4 Table 3. 7b - Calculated MSLB Accident Radiological Consequences TEDE (rem)

EAB LPZ Control Room Calculated results Pre-Incident Spike 0.94 0.37 0.23 Concurrent Iodine Spike 0.95 0.45 0.45 TEDE (rem)

EAB LPZ Control Room Dose acceptance criteria Fuel Damage or Pre-Incident Spike 25 25 5 Concurrent Iodine Spike 2.5 2.5 5 E1 - 13

Enclosure 1 to NL-16-0388 Basis for Proposed Change 3.6 Steam Generator Tube Rupture Accident 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 (1.0 gpm total).

Consistent with the FNP current licensing basis, 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 p.ortion 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 factor. A SG partition factor 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

µCi/gm. For the concurrent spike case, the initial primary iodine activity concentration is at the equilibrium TS value of 0.5 µCi/g Dose Equivalent Iodine.

This concurrent 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 as an initial condition, RCS activity includes consideration of 1% failed (leaking) fuel , consistent with the FNP 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 Reactor Coolant Activity (Initial)

Pre-Accident Iodine Spike 30 µCi/gm DE 1-131 Accident-Initiated Iodine Spike 0.5 µCi/gm DE 1-131 Noble Gas 1% failed (leaking) fuel Concurrent Iodine Spiking Factor 335 E1 - 14 to NL-16-0388 Basis for Proposed Change 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 0.1 µCi/gm DE 1-131 Atmospheric Dispersion Factors (sec/m 3)

Time (hr) EAB LPZ Control Room 0-2 7.6E-4 2.80E-4 1.66E-3 2-8 1.1 OE-4 1.38E-3 Control Room Parameters Parameter Volume 114,000 ft3 Ventilation System Makeup Rate 375 cfm Ventilation System Recirculation Flow Rate 2700 cfm Ventilation System Charcoal Filter Efficiencies Pressurization Filters 98.5% all iodine species Recirculation Filters 94.5% elemental and organic 98.5% particulate Unfiltered In-leakage 310 cfm (includes 10 cfm CR ingress/egress)

Breathing Rate 3.5E-4 m3/sec 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 Control Room Calculated results Pre-Incident Spike 2.4 0.92 0.48 Concurrent Iodine Spike 0.82 0.34 0.17 TEDE (rem)

EAB LPZ Control Room Dose acceptance criteria Fuel Damage or Pre-Incident Spike 25 25 5 Concurrent Iodine Spike 2.5 2.5 5 E1 - 15 to NL-16-0388 Basis for Proposed Change 3.7 Control Rod Ejection Accident 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 iodine isotopes 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 iodine isotopes 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 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 iodine isotopes in the melted fuel are 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 from the RCS 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 Control Rod Ejection 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 Control Rod Ejection. 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.

E1 - 16 to NL-16-0388 Basis for Proposed Change Table 3.9a - Parameters and Assumptions for the Control Rod Ejection Accident Parameter Reactor power 2831 MWt Post-accident damaged 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 Atmospheric Dispersion Factors (sec/m 3 )

Time (hr) EAB LPZ Control Room 0-2 7.6E-4 2.80E-4 1.66E-3 2-8 1.1 OE-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 4.21 E-4 Control Room Parameters Parameter Volume 114,000 ft 3 Ventilation System Makeup Rate 375 cfm Ventilation System Recirculation Flow Rate 2700 cfm Ventilation System Charcoal Filter Efficiencies Pressurization Filters 98.5% all iodine species Recirculation Filters 94.5% elemental and organic 98.5% particulate Unfiltered In-leakage 310 cfm (includes 1O cfm CR ingress/egress)

Breathing Rate 3.5E-4 m3/sec E1 - 17 to NL-16-0388 Basis for Proposed Change Occupancy Factors 0-24 hours 1.0 1 - 4 days 0.6 4 -30 days 0.4 Table 3.9b - Calculated Control Rod Ejection Accident Radiological Consequences TEDE (rem)

EAB LPZ Control Room Calculated results 3.8 2.7 3.7 Dose acceptance criteria 6.3 6.3 5 3.8 Locked Rotor Accident The Locked Rotor Accident 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 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. For conservatism, an assessment is being performed for a delayed manual CREFS initiation.

Results of this assessment are expected to be within 5 rem TEDE.

The analysis used assumptions and inputs that follow the guidance in RG 1.183.

The key parameters and assumptions are listed in Table 3.1 Oa. The calculated dose results are given in Table 3.1 Ob. 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 3.1 Oa - Parameters and Assumptions for the Locked Rotor Accident Parameter Reactor power 2831 MWt Post-Locked Rotor Accident 20%

Secondary Coolant Iodine Specific Activity 0.1 µCi/gm DE 1-131 E1 - 18

Enclosure 1 to NL-16-0388 Basis for Proposed Change 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 lbm 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 833,221 lbm Atmospheric Dispersion Factors (sec/m 3 )

Time (hr) EAB LPZ Control Room 0-2 7.6E-4 2.80E-4 1.66E-3 2-8 1.1 OE-4 1.38E-3 Control Room Parameters Parameter Value Volume 114,000 ft3 Ventilation System Makeup Rate 375 cfm Ventilation System Recirculation Flow Rate 2700 cfm Ventilation System Charcoal Filter Efficiencies Pressurization Filters 98.5% all iodine species Recirculation Filters 94.5% elemental and organic 98.5% particulate Unfiltered In-leakage 31 O cfm (includes 1O cfm CR ingress/egress)

Breathing Rate 3.5E-4 m3/sec Occupancy Factors 0-24 hours 1.0 1 - 4 days 0.6 4-30 days 0.4 Table 3.1 Ob - Calculated Locked Rotor Accident Radiological Consequences TEDE (rem)

EAB LPZ Control Room 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.

E1 - 19 to NL-16-0388 Basis for Proposed Change 3.9 Conclusions The proposed changes provide a source term for FNP that will result in a more accurate assessment of the OBA 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.

SNC has assessed the dose to CR occupants during CR ingress/egress, such as with shift turnover activities. Consistent with Standard Review Plan 6.4, 1O CFM of unfiltered in-leakage is included for CR ingress/egress in the dose analyses to account for this. However, SNC recognizes the Environmental Protection Agency (EPA) guidance which limits the dose to an emergency worker during nuclear incidents to 5 rem (which includes shift turnover transit time from the EAB to the CR). Consistent with the FNP current licensing basis, controls are in place during radiological events to limit dose to emergency workers. In general, for any accident where a release is in progress, radiological protection personnel are monitoring the dose rates onsite in various areas, including offsite and in the control room. The Emergency Response Organization has specific members that are monitoring the release and the direction of the plume. It is not permitted for any relief workers to enter the plant from the direction of the plume, and especially not in their own vehicles. SNC radiological protection staff will bring the relief workers to the plant in buses/vans, per the Emergency Plan. Masks and or respirators will be provided, as needed, to assure workers dose is as low as reasonably achievable. For additional conservatism, CR occupants will ingress/egress via the secondary CR door during an FHA. Given these controls and protection factors afforded by emergency equipment, the doses to CR occupants during shift turnover ingress and egress will be within the EPA dose guidance for emergency workers.

3.1 O TS Discussion 3.10.1 TSTF-448 With License Amendment 166/158 for FNP Units 1/2 (Reference 3), 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 (SE) for that amendment:

"The CRI P represents the manner in which the licensee will demonstrate CRE [Control Room Envelope] integrity. SNC has proposed a GRIP 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 GRIP 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 License Amendment 166/158 (September 30, 2004) predated NRC approval of TSTF-448 Revision 3 (January 17, 2007).

E1 - 20 to NL-16-0388 Basis for Proposed Change Accordingly, the proposed adoption of TSTF-448 will supersede the current licensing basis for the CRIP.

3.10.1.1 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.

3.10.1.2 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 optional statements and evaluations provided on Tables 3.11 a and 3.11 b below are applicable.

E1 - 21 to NL-16-0388 Basis for Proposed Change Table 3.11 a - Options With the Model Safety Evaluation

Number, Model SE SE Statement' FNP Option* Justification

.,, Location ,

1 Throughout SE Model SE uses the term "[CREEVS]" The FNP term is "CREFS" which This is an option defined as "[Control Room Envelope is defined as the "Control Room authorized by the Emergency Ventilation System]" to Emergency Filtration/ SE.

describe the CR ventilation system. Pressurization System."

2 Throughout SE Model SE offer~ the term "[trains]" or FNP uses "trains" when referring This is an option

"[subsystem]" when referring to to CREFS. authorized by the CREEVS. SE.

3 Throughout SE Model SE describes the CRE SNC confirms that bracketed This is an SE administrative control program as designation. confirmation.

"[Specification 5.5.18, "CRE Habitability ProQram]"

4 Section 1.0, first Model SE includes the term "[Name of The facility name is "Farley This information paragraph the Licensee]." Nuclear Plant, Units 1 and 2." authorized by the SE.

5 Section 2.2, Model SE provides the options of "[5 With this AST LAR, FNP will be This is an option second rem whole body dose or its equivalent using 5 rem TEDE. authorized by the paragraph to any part of the body]" or "[5 rem total SE.

effective dose equivalent (TEDE)]."

6 Section 2.3, first Model SE provides an option for FNP was licensed under the 1O This option is not paragraph facilities not licensed under the 10 CFR CFR 50 General Design Criteria. being used by FNP.

50 General Design Criteria.

7 Section 3.0 Model SE provides the following The FNP CREFS uses This is an option bracketed statement: "[The emergency pressurization. The entire authorized by the operational mode of the [CREEVS] at statement should read: "The SE.

[facility name][pressurizes] [isolates but emergency operational mode of does not pressurize] the CRE to the CREFS at Farley Nuclear minimize unfiltered air in-leakage]. Plant, Units 1 and 2 pressurizes the CRE to minimize unfiltered air in-leakaoe.

E1 - 22 to NL-16-0388 Basis for Proposed Change

'\

~umber Model SE* .$E Statement * , FNP*Option Justification . '

,_ Location . , " *c< ' '

8 Section 3.1, first Model SE states: "[Except for plant There are no plant-specific This option is not paragraph specific differences, all of] these exceptions being taken from the being used by FNP.

changes are consistent with STS as Standard Technical revised by TSTF-448, Revision 3." Specifications text, as revised by TSTF-448.

9 Section 3.1, Model SE describes the TS Bases The FNP TS Bases Control This is an option second Control Program as being "TS 5.5.[11 ]." Program is described in TS authorized by the paragraph 5.5.14, not TS 5.5.11. SE.

10 Section 3.2 Model SE describes an option to FNP TS 3.7.10 Condition E does This option is not correct a typographical error by not contain this typographical being used by FNP.

replacing "irradiate" with "irradiated." error, and so is not utilizing this option.

11 Section 3.3 Model SE provides six evaluation SNC has selected Evaluation 1 This is an option options as that which is most applicable authorized by the to the FNP technical SE.

specifications.

12 Section 3.3, Model SE describes an option of not SNC proposes to follow the air This option is not paragraph following the ASTM E741 testing in-leakage testing methodology being used by FNP.

below end of methodology, and providing an described in that paragraph.

Evaluation 6 alternative methodology that has been accepted by the NRC.

13 Section 3.4, first Model SE provides the option radiation With this AST LAA, radiation This is an option paragraph exposures in units of "whole body or its exposures will be in units of authorized by the equivalent to any part of the body" or TEDE. SE.

"total effective dose equivalent."

14 Section 3.4, fifth Model SE provides an option for taking SNC is taking no exceptions to This option is not paragraph exception(s) to Sections C.1 and C.2 of Section C.1 and C.2 of RG being used by FNP.

RG 1.197. 1.197.

15 Section 3.4, Model SE provides a bracketed FNP is on a 24-month fuel cycle, This is an option sixth paragraph frequency interval of 18 months for so the frequency interval will be authorized by the CRE pressure testinQ. 24 months. SE.

16 Section 4 Model SE provides a bracket to "Alabama" is the applicable This is an option describe the State applicable to the State for the State official. authorized by the State official. SE.

E1 - 23

Enclosure 1 to NL-16-0388 Basis for Proposed Change Table 3.11 b - Variances With the Model Safety Evaluation

Number Model SE'* SE Statement FNP Variation Justification

.Location-* .. *~'

1 Section 1.0, last Model SE describes TS The FNP title of TS 3.7.10 is With License Amendments paragraph, and 3.7.1 Oas "Control Room "Control Room" which is being 166/158, the title of TS 3.7.10 Section 3.1, first Envelope Emergency changed to "Control Room was changed from "CREFS" to paragraph Ventilation System Emergency Filtration/ "Control Room" and established (CREEVS)." Pressurization System (CREFS). CRE operability as distinct from This is a change not specifically CREFS operability. This LAA is addressed by the TSTF. restoring the previous title and establishing the CRE as a subsystem for CREFS operability. The title change itself is an administrative change.

2 Section 1.0, last Model SE describes TS FNP has an existing TS 5.5.18 The existing TS 5.5.18 for GRIP paragraph, and 5.5.18 as a new Control Room Integrity Program was introduced with License Section 3.1, first administrative controls (GRIP) that is being replaced in its Amendment 166 and 158 for FNP paragraph program. entirety by the TSTF-448 Control Units 1 and 2 respectively. These Room Envelope Habitability amendments for issued on program. Therefore, the model SE September 30, 2004, well before should change the word "new" to NRG acceptance of TSTF-448

"revised." Revision 3 as a basis for Control Room habitability.

The GRIP requirements are comparable to the CRE Habitability program, except the GRIP is more prescriptive in the testing requirements for CR in-leakage, and cites specific acceptance criteria, rather than referring to the values included in the licensing basis OBA analyses. The GRIP also more strinqentlv defines the CRE E1 - 24

Enclosure 1 to NL-16-0388 Basis for Proposed Change

... ,,_ *' ,", ~

'.Numb*er *FNP* vB,riation f

.. .. .. :.Model SE/ .. si= statement . Justification *J

. \. ..Location , * ... I* ... *\: . .. ( "

. ~-* <~' ' :r* .. * *.* . ,.

configuration control program.

The TSTF-448 CRE Habitability Program provides a technically justified alternative to the existing GRIP and provides reasonable assurance of public health and safety.

3 Section 3. 1, N/A SNC is making several changes to FNP Amendment 166/158 Proposed TS 3.7.10 to make it comport to established the CRE as a Changes the Westinghouse Owners Group separate entity from CREFS with (WOG) STS, as revised by TSTF- its own operability requirements.

448, but that are not specifically The proposed change to treat the addressed by the TSTF: CRE as a subsystem for CREFS operability, consistent with the

1) Revising LCO 3.7.1 O from Two WOG STS, has no adverse Control Room Emergency impact on how these SSCs are Filtration/Pressurization System controlled.

(CREFS) trains and the Control Room Envelope (CRE) shall be The changes to Required Actions OPERABLE" to Two CREFS B.2 effectively eliminate the trains shall be OPERABLE." option of restoring CRE to operability in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , in lieu of

2) Revise the Required Actions of verifying the effectiveness of the B.2.1 from B.2.1 (Restore CRE mitigating actions. Therefore this to OPERABLE status) OR change is more restrictive.

B.2.2.1 (Verify GOG 19 is met)

AND B.2.2.2 (Restore CRE to Specific SRs to .verifying CRE ~p OPERABLE status), to B.2 is within limits and require CRE (Verify mitigating actions) AND integrity testing is not necessary, B.3 (Restore CRE boundary to as these provisions are in the TS OPERABLE status). 5.5.18 CRE Habitability Program.

E1 - 25 to NL-16-0388 Basis for Proposed Change Number Model SE SE Statement FNP Variation Justification

'~ <

., Location

3) Revising SR 3.7.10.4 from "Verify CRE b.p within limits in the CRIP" to the new wording of the TSTF.

4) Deleting SR 3.7.10.5 which verifies CRE integrity in accordance with the CRIP.

4 Section 3.3, first Model SE describes FNP has not adopted TSTF-287, Of the six Evaluation options, sentence of Evaluation 1 as being but has a Note that is similar, Evaluation 1 is the most Evaluation 1 for facilities that have although Action B is dissimilar. applicable to the FNP TS.

adopted the [CREEVS]

TS LCO Note and Action B of TSTF-287, Rev. 5.

5 Section 3.3, The model SE states: TS 3.7.10 for FNP is organized As discussed in Item 3 above, Evaluation 1, "The licensee propose differently than the WOG STS Amendment 166/158 established first paragraph to revise the action described in TSTF-448. The separate operability requirements requirements of TS model SE should read: "The for the CRE and for CREFS,

[3. 7.10, "CREEVS,"] to licensee propose to revise the which resulted in a Condition B acknowledge that an action requirements of TS 3.7.1 O that only covered an inoperable inoperable CRE to establish that one or more CRE. Consistent with the WOG boundary, depending CREFS trains will be inoperable STS, as revised by TSTF-448, upon the location of the due to an inoperable CRE the CRE will be a subsystem associated degradation, boundary." required for CREFS operability.

could cause just one, This has no adverse impact on instead of both how these SSCs are controlled.

[CREEVS] [trains] to be inoperable."

E1 - 26

Enclosure 1 to NL-16-0388 Basis for Proposed Change

Number

'""'*J!'"o;- '>e *

ModelSE- ., SE Statement :FNP Variation:' J1,1i:;tification

- ~

~<.

, ,_:; t:otation '.:.. :* .*

6 Section 3.3, The model SE state~: As stated in Item 5 above, TS As discussed in Item 3 above, Evaluation 1, "This change clarifies 3.7.10 for FNP is organized Amendment 166/158 established second how to apply the action differently than the WOG STS separate operability requirements paragraph requirements in the described in TSTF-448. The for the CRE and for CREFS, event just one CREFS model SE should read: This which resulted in a Condition B train is unable to ensure change clarifies how to apply the that only covered an inoperable CRE occupant action requirements in the event CRE. Consistent with the WOG safety ... because of an that one or both CREFS trains are STS, as revised by TSTF-448, inoperable CRE unable to ensure CRE occupant the CRE will be a subsystem boundary." safety ... because of an inoperable required for CREFS operability.

CRE boundary." This has no adverse impact on how these SSCs are controlled.

7 Section 3.3, The model SE states: Should read: "One or more For FNP, TS 3.7.10 is applicable Evaluation 1 , "One or more CREFS trains inoperable due to in MODES 1, 2, 3, and 4, during first paragraph, [CREEVS][trains] inoperable CRE boundary." movement of irradiated fuel Condition B inoperable due to assemblies, and during Core bullet inoperable CRE Alterations." Condition B remains boundary in MODE 1, 2, applicable in all those operating

[or] 3[, or 4]." states.

8 Section 3.3, The model SE states: TS 3.7.10 for FNP is organized This is considered to be an Evaluation 1 , "The licensee proposes differently than the WOG STS editorial change.

third paragraph to replace existing described in TSTF-448. The Required Action B.1, model SE should read: The

'Restore control room licensee proposes to replace boundary to OPERABLE existing Required Action B.1, status,' which has a 24- 'Initiate mitigating actions,' which hour Completion has an immediate Completion Time ... " Time ...

E1 - 27 to NL-16-0388 Basis for Proposed Change 3.10.1.3 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 TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air in-leakage 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 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 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.

3.10.2 TSTF-312 3.10.2.1 Summary of the Approved Traveler Justification The proposed TS change adds a Note to the Limiting Condition for Operation (LCO) for TS 3.9.3, "Containment Penetrations," allowing "Penetration flow path(s) that have direct access from the containment atmosphere to the outside atmosphere to be unisolated under administrative control." The Applicability for LCO 3.9.3 is during core alterations, and during movement or irradiated fuel assemblies within containment.

E1 - 28 to NL-16-0388 Basis for Proposed Change The changes proposed in TSTF-312-A, Revision 1, are consistent with those in Specification 3.6.3, "Containment Isolation Valves." TS 3.6.3, Actions Note 1, allows penetration flow path(s) (except for the 24 inch purge valves) to be unisolated intermittently under administrative control, and is Applicable in MODES 1, 2, 3, and 4.

Under the applicable conditions for LCO 3.6.3, the accident analyses credit the primary containment as a release barrier. The proposed change to LCO 3.9.3 would be Applicable under significantly lower energy conditions than those that apply for LCO 3.6.3, and is therefore less risk significant. Adoption of this change is proposed to provide a consistent approach to containment boundary issues that utilizes previously approved and acceptable compensatory measures.

The proposed change also includes the addition of text to the LCO discussion in Bases 3.9.3 stipulating that the administrative controls that are put in place when penetrations flow path(s) are unisolated ensure that: 1) appropriate personnel are aware of the open status of the penetration flow path during core alterations or movement of irradiated fuel assemblies within the containment, and 2) specified individuals are designated and readily available to isolate the flow path in the event of an FHA.

TSTF-312-A includes a Reviewer's Note that identifies the need for a confirmatory FHA dose calculation that has been accepted by the NRC staff, and that indicates acceptable radiological consequences.

This TSTF is predicated upon NRC acceptance of this AST LAA for the FHA dose consequences.

The Reviewer's Note identifies licensee commitments to implement administrative procedures that ensure the open containment airlock can be promptly closed in the event of an FHA following personnel evacuation, and that open penetration flow path(s) can be promptly closed. The Reviewer's Note also identifies that the time to close such penetrations, or combination of penetrations, should be included in the confirmatory dose calculations.

The Farley FHA dose calculation analyzes offsite and control room doses for FHA events within containment, and evaluates scenarios where the equipment hatch and/or personnel airlocks are open.

Doses are calculated for the 0-2 hour period, and essentially all of the activity that is released into containment by the FHA event is released from containment during these 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . There is fundamentally no contribution to calculated doses after the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the event.

Due to the fact that essentially all of the activity that is released into containment during the inside containment FHA is assumed to be released during the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the event, calculated doses for a release through an open equipment hatch and/or open personnel airlock doors are very nearly the same. Although prompt closure of E1 - 29 to NL-16-0388 Basis for Proposed Change the equipment hatch and the personnel airlock can be achieved, no credit for these actions are taken in the dose calculation.

Similarly, offsite and control room doses resulting from either a simultaneous or individual release through one or more open containment penetrations, an open equipment hatch, or open personnel airlock doors will be very nearly the same as those calculated for an open equipment hatch and/or open personnel airlock doors, and will also be within acceptance limits without assuming that any leak paths are isolated. Consequently, it is not necessary to provide the time to close unisolated containment penetrations(s) in the FHA dose calculations.

Allowing penetration flow paths to be unisolated during core alterations or movement of irradiated fuel will not invalidate the conclusion that he potential dose consequences from a FHA are within 10 CFR 50.67 limits.

SNC will establish administrative controls to ensure: 1) appropriate personnel are aware of the open status of the penetration flow path(s) during core alterations or movement of irradiated fuel assemblies within the containment, and 2) specified individuals are designated and readily available to isolate any open penetration flow path(s) in the event of an FHA inside containment.

3.10.2.2 Differences Between the Proposed Change and the Approved Traveler The Farley Section 3.9 specification numbers are different from the Improved Standard Technical Specification (ISTS) Section 3.9 specification numbers. Farley Specification 3.9.3, "Containment Penetrations," is equivalent to Specification 3.9.4 in the ISTS. This has no effect on the requested change.

Farley LCO 3.9.3.b was previously amended to allow the personnel and equipment airlocks to remain open during core alterations or movement of irradiated fuel assemblies within the containment, provided one airlock door was available and a designated individual was available to close the open airlock door(s) if needed. The scope of this previous amendment overlaps the scope of TSTF-312-A, and as a result LCO 3.9.3 and its associated Bases differ from that presented in TSTF-312-A. The Note for LCO 3.9.3 and the supplemental LCO text for Bases 3.9.3 are incorporated without change from TSTF-312-A. No additional changes to the LCO and Bases were necessary of made as a result of the existing allowance for the personnel and equipment airlock.

E1 - 30 to NL-16-0388 Basis for Proposed Change 4.0 Regulatory Safety Analysis 4.1 Applicable Regulatory Requirements/Criteria Title 1O 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 NRG Notice of Availability published in Reference 2, and TSTF-448, Revision 3.

Title 1O Code of Federal Regulations Section 50.67. "Accident Source Term" On December 23, 1999, the NRG 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 OBA 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.

4.1.1 Additional Applicable Regulatory Criteria for TSTF-312 Appendix A to Title 1O of the Code of Federal Regulations (1 O CFR), Part 50, "General Design Criteria for Nuclear Power Plants," contains the following pertinent criteria:

Criterion 56, Primary containment isolation, states:

Each line that connects directly to the containment atmosphere and penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lies, are acceptable on some other defined basis:

(1) One locked closed isolation valve inside and one locked closed isolation valve outside containment; or (2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or (3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or (4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as he automatic isolation valve outside containment.

Isolation valves outside containment shall be located as close to the containment as practical and upon loss of actuating power, automatic E1 - 31 to NL-16-0388 Basis for Proposed Change isolation valves shall be designed to take the position that provides greater safety.

The proposed change to LCO 3.9.3 will allow containment penetration flow path(s) to be open during refueling operations under administrative control. This change does not significantly change how the plant would mitigate an accident previously evaluated, and is bounded by existing FHA accident analysis.

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 approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

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.

TSTF-312, Revision 1, was approved by the NRC as documented in a letter from William Beckner (NRC) to James Davis (NEI), dated August 16, 1999 (CAN 9908250220). TSTF-312-A, Revision 1, has been adopted by many plants as part of a complete conversion to the ISTS, such as North Anna Power Station (CAN ML0212110540). An example of a plant-specific NRC approval of the changes in TSTF-312-A, Revision 1, is Arkansas Nuclear One, Unit 1, Amendment Number 245 dated August 10, 2011 (ACN ML111940085).

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.

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 (CLllP). 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 E1 - 32 to NL-16-0388 Basis for Proposed Change Alternative Source Term (AST) and the new atmospheric dispersion factors have no impact on the probability for initiation of any Design Basis Accidents (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 15 of the FSAR. Plant-specific radiological analyses have been performed using the AST methodology and new atmospheric dispersion factors (X/Qs) have been established.

Based on the results of these analyses, it has been demonstrated that the Control Room 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.

Regarding TSTF-312-A, the proposed change would allow containment penetrations to be unisolated under administrative controls during core alterations or movement of irradiated fuel assemblies within containment.

The status of containment penetration flow paths (i.e., open or closed) is not an initiator for any design basis accident or event, and therefore the proposed change does not increase the probability of any accident previously evaluated. The proposed change does not affect the design of the primary containment, or alter plant operating practices such that the probability of an accident previously evaluated would be significantly increased. The proposed change does not significantly change how the plant would mitigate an accident previously evaluated, and is bounded by the fuel handling accident (FHA) analysis.

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 modes 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 affect the design ful)ction or modes of operation of structures, systems E1 - 33 to NL-16-0388 Basis for Proposed Change 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.

Regarding TSTF-312-A, allowing penetration flow paths to be open is not an initiator for any accident. The proposed change to allow open penetration flow paths will not affect plant safety functions or plant operating practices such that a new or different accident could be created. There are no design changes associated with the proposed changes, and the change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). The change does not alter assumptions made in the safety analysis, and is consistent with the safety analysis assumptions and current plant operating practice. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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.

Regarding TSTF-312-A, TS 3.9.3 provides measures to ensure that the dose consequences of a postulated FHA inside containment are minimized. The proposed change to LCO 3.9.3 will allow penetration flow path(s) to be open during refueling operations under administrative control. These administrative controls will provide assurance that prompt closure of open penetrations flow paths can and will be achieved in the event of an FHA inside containment, and will minimize dose consequences. The proposed change does not affect the safety analysis acceptance criteria for ay analyzed event, nor is there a change to any safety analysis limit. The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are deterred, not is here any adverse effect on those plant systems necessary to assure the accomplishment of protective functions.

The proposed change will not result in plant operation in a configuration outside the design basis.

E1 - 34 to NL-16-0388 Basis for Proposed Change 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 the proposed amendment.

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 CLllP. 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 (NRG) 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)." [ADAMS Accession Number ML042780424]

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

E1 - 35 to NL-16-0388 Basis for Proposed Change

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.

MC0625 and MC0626)." [ADAMS Accession Number ML042820368]

E1 - 36

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

I

(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 NRG in writing when implementation of these activities is complete and can be verified by NRG inspection.

(6) Reactor Vessel Material Surveillance Capsules All capsules in the reactor vessel that are removed and tested must U1/U2 meet the test procedures and reporting requirements of American Operating Society for Testing and Materials (ASTM) E 185-82 to the extent License practicable for the configuration of the specimens in the capsule. Any Condition 7 changes to the capsule withdrawal schedule, including spare capsules, must be approved by the NRG 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. 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 1 Renewed License No. NPF-2 Amendment No. +99

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-14-1273, 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, U1/U2 Table S-3, "Implementation Items," of SNC letter NL-14-1273, dated Operating August 29, 2014, within 180 days after NRC approval, except for items 30 and 32.

License "'-

Condition 7 I ~

Items 30 and 32 shall be implemented by February 6, 2018.

(7) 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.

Farley - Unit 2 Renewed License No. NPF-8 Amendment No. ~

U1/U2 Operating License Condition 7 Insert (7) Upon implementation of Amendment No. xxx adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air in-leakage 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 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 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.

Control RoomCREFS 3.7.10

3. 7 PLANT SYSTEMS

3. 7 .1 O 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.

N()TE ---------------------------------------------

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 ALTERATl()NS.

ACTl()NS CONDITl()N REQUIRED ACTl()N COMPLETION TIME A. One CREFS train A.1 Restore CREFS train to 7 days inoperable for reasons OPERABLE status.

other than Condition B.

B. One or more CREFS B.1 Initiate action to Immediately trains inoQerable due to imQlement mitigating inoQerable CRE actions.

inoperableboundary.

AND B.2.~ Restore CRE to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE status OR B.2~ General Qesign Criteria 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (GQC) ~ 9 met 1o1sing Verify mitigating actions

/ , :~:aiiRg astions iAB.1.

ensure CRE occuQant exQosures to radiological 1 chemical 1 and smoke v

hazards will not exceed B.~~ Restore CRE boundary to dG90 days limits. OPERABLE status.

Farley Units 1 and 2 3.7.10-1 Amendment No. +ee (Unit 1)

Amendment No. +as (Unit 2)

Control RoomCREFS 3.7.10 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

c. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or B AND not met in MODE 1, 2, 3, or 4. C.2 ------------NOTE-----------

LCO 3.0.4.a is not applicable when entering MODE4.

Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D. Two CREFS trains D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months inoperable in MODE 1, 2, 3,0R4.

AND D.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months E. Required Action and E.1 Place OPERABLE CREFS Immediately associated Completion train in emergency Time of Condition A not recirculation mode.

met during movement of irradiated fuel assemblies OR or during CORE ALTERATIONS. E.2.1 Suspend CORE Immediately AL TERATIO NS.

AND E.2.2 Suspend movement of Immediately irradiated fuel assemblies.

Farley Units 1 and 2 3.7.10-2 Amendment No. +ee (Unit 1)

Amendment No. +a8 (Unit 2)

Control RoomCREFS 3.7.10 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F. Required Action and F.1 Suspend CORE Immediately associated Completion ALTERATIONS.

Time of Condition B not met during movement of AND irradiated fuel assemblies or during CORE F.2 Suspend movement of Immediately ALTERATIONS. irradiated fuel assemblies.

OR OR Two CREFS trains One or more CREFS inoperable during trains ino12erable due to an movement of irradiated fuel ino12erable CRE bounda~

assemblies or during during movement of COR~

irradiated fuel assemblies or during CORE ALTERATIONS.

SURVEILLANCE REQUREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.1 Operate each CREFS Pressurization train with the In accordance with heaters operating and each CREFS Recirculation and the Surveillance Filtration train for ~ 15 minutes. Frequency Control Program SR 3.7.10.2 Perform required CREFS filter testing in accordance In accordance with with \ the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.7.10.3 ----------------------------l\IOTE--------------------------------- In accordance with 1\lot required to be performed in MODES 5 and 6. the Surveillance Frequency Control Program Verify each CREFS train actuates on an actual or simulated actuation signal.

Farley Units 1 and 2 3.7.10-3 Amendment l\lo. +se (Unit 1)

Amendment l\lo. +w (Unit 2)

Control RoomCREFS 3.7.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.4 Perform required CRE unfiltered air in-leakage testing in In accordance accordance with the Control Room Envelope Habitability with the Control ProgramVerify CRE Ap within limits in the Control Room Room Envelope Integrity Program (GRIP). Habitability Program24 months on a STAGGERED TEST BASIS SR 3.7.10.5 Verify CRE integrity in aooordanoe with the GRIP. In aooordanoe

'Nith the GRIP Farley Units 1 and 2 3.7.10-4 Amendment No. +es (Unit 1)

Amendment No. +as (Unit 2)

Containment Penetrations 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Containment Penetrations LCO 3.9.3 The containment penetrations shall be in the following status:

a. The equipment hatch is capable of being closed and held in place by four bolts;

b. One door in each air lock is capable of being closed; and

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:

1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or

2. capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System.

f\l()"f E:

Penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative control.

APPLICABILITY: During CORE ALTERATIONS, During movement of irradiated fuel assemblies within containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more containment A.1 Suspend CORE Immediately penetrations not in ALTERATIONS.

required status.

AND A.2 Suspend movement of Immediately irradiated fuel assemblies within containment.

Farley Units 1 and 2 3.9.3-1 Amendment No. +78 (Unit 1)

Amendment No. +++ (Unit 2)

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

A Control Room Integrity Program (GRIP) shall be established and implemented to ensure that the control room integrity is maintained such that a radiological event, hazarqous chemicals, or a fire challenge (e.g., fire byproducts, halon, eta.)

Control Room will not prevent the control room operators from eontrolling the reactor during Envelope normal or aooident conditions. The program shall require testing as outlined Habitability below. Testing should be performed \"Jhen changes are made to structures, Program Insert systems and components whioh oould impact Control Room Impact (CRE) integrity. Those structures, systems and components may be internal or external to the CRE. Testing should also be conducted follov1ing a modification or a repair that oould affect CRE inleakage. Testing should also be performed if the conditions assoeiated with a particular ehallenge result in a change in operating mode, system alignment or system response that oould result in a new limiting condition. Testing should be commensurate with the type and degree of modification or repair. Testing should be oonduoted in the alignment that results in the greatest consequence to the operators.

A GRIP 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 belmv values. The values listed below do not inelude 1O ofm assumed in accident analysis for ingress I egress.

i) 4a cfm *.vhen the control room ventilation systems are aligned in the emergency reoirculation mode of operation, ii) 600 efm 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 oonfiguration control program will determine 'Nhether the i) CRE differential pressure relative to acijaoent areas and ii) the control room ventilation system flow rates, as determined in accordance with ASME N51 O 1989 or ASTM E2029 99, are consistent *.vith 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 eharacteristies of the Farley Units 1 and 2 5.5-15 Amendment No. +92 (Unit 1)

Amendment No. +sa (Unit 2)

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.18 Control Room Integrity Program (GRIP) (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.

5.5.19 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.

Farley Units 1 and 2 5.5-16 Amendment No. +gg (Unit 1)

Amendment No. 88 (Unit 2)

Control Room Envelope Habitability Program Insert 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 (OBA) 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 OBA 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 82.1.2 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 Ill 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)

Farley Units 1 and 2 B 2.1.2-1 Revision O

RCS Pressure SL B2.1.2 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 Ill, Article NB-7000.

3. ASME, Boiler and Pressure Vessel Code, Section XI, Article IWX-5000.

4. 10 CFR 50.67.+oo.

5. FSAR. Section 7.2.

Farley Units 1 and 2 B 2.1.2-3 Revision O

SOM B 3.1.1 BASES APPLICABLE SOM satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). Even though SAFETY ANALYSES it is not directly observed from the control room, SOM 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 SOM 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 SOM 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, "Reastor 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 kett < 1.0 and in MODES 3, 4, and 5, the SOM 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, SOM is ensured by complying with LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits."

ACTIONS If the SOM 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)

Farley Units 1 and 2 B 3.1.1-4 Revision O

SOM B 3.1.1 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 50.67-+00.

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.

Farley Units 1 and 2 B 3.1.1-6 Revision O

RTS Instrumentation B 3.3.1 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 50.67.+-0G limits. Different accident categories are allowed a different fraction of these limits, based on probability of (continued)

Farley Units 1 and 2 B 3.3.1-1 Revision 0

Containment Purge and Exhaust Isolation Instrumentation B 3.3.6 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 50.67-WQ (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 ES FAS Function 1.b (Paragraph 1), SI, and ES FAS Function 3.a, (continued)

Farley Units 1 and 2 B 3.3.6-2 Revision O

Containment Purge and Exhaust Isolation Instrumentation B 3.3.6 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 mea.sured 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, ES FAS.

REFERENCES 1. 10 CFR 50.67100.11.

2. Not used.

3. Not used.

4. FSAR Table 7,3-16 Farley Units 1 and 2 B 3.3.6-9 Revision O

PRF Actuation Instrumentation B 3.3.8 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 50.67.+oo (Ref. 1).

(continued)

Farley Units 1 and 2 B 3.3.8-1 Revision 18

PRF Actuation Instrumentation B 3.3.8 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.

Farley Units 1 and 2 B 3.3.8-9 Revision 52

RCS Operational LEAKAGE B 3.4.13 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 9.Q.!D.470 gpd primary to secondary LEAKAGE in the faulted and both intact steam generators respectively as an initial condition (1 gpm total).

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 50.67.+GQ.. The RCS specific activity assumed was 0.5 µCi/gm DOSE EQUIVALENT 1-131 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).

Farley Units 1 and 2 B 3.4.13-2 Revision 24

RCS Specific Activity B 3.4.16 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 50.67.+QG (Ref. 1). The limits on specific activity ensure that the doses are held to an appropriate fraction of the 10 CFR 50.67.:t-OG limits (i.e., a small fraction of or well within the 10 CFR 50.67.+QG 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 1-131 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 1 O CFR 50.67.+QG 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 sho*11ed the potential offsite dose levels for a SGTR or MSLB aooident *.vere an appropriately small fraction of the 10 CFR 100 dose guideline limits. Eash evaluation assumes a broad range of site applioable atmospherio dispersion faotors in a parametrio 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 50.67.:t-OO 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 µCi/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)

Farley Units 1 and 2 B 3.4.16-1 Revision 15

RCS Specific Activity B 3.4.16 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 14 4 0 gpd. Theseis analysgis resulted in offsite doses bounded by a small fraction (i.e., 10%) of the 10 CFR 50.67-1-00 guidelines using FGR No. 11 and 121CRP 30 Dose Conversion Factors (DCFs). The initial RCS specific activity assumed was 0.5 µCi/gm DOSE EQUIVALENT 1-131 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 µCi/gm DOSE EQUIVALENT 1-131 from LCO 3.7.16, "Secondary Specific Activity."

The analysis for the SGTR and MSLB accident§. establishes the acceptance limits for RCS specific activity. Reference to theseis analyseis areis 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 µCi/gm DOSE EQUIVALENT I 131 at a conservatively high letdown flov,i of 145 *

@ffi;. The SGTR and MSLB analyseis considers two cases of reactor coolant specific activity. One case assumes specific activity at 0.5 µCi/gm DOSE EQUIVALENT 1-131 at a conservatively high letdown flow of 145 gpm with an accident initiated iodine spike that increases the 1-131 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 µCi/gm DOSE EQUIVALENT 1-131 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 equals the LCO limit of 100/E µCi/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 SG discharges radioactively contaminated steam to the atmosphere (continued)

Farley Units 1 and 2 B 3.4.16-2 Revision 15

RCS Specific Activity B 3.4.16 BASES APPLICABLE through the SG power operated relief valves and the main steam SAFETY ANALYSES safety valves. The unaffected SGs remove core decay heat by (continued) 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 µCi/gm DOSE EQUIVALENT 1-131.

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 50.67.:t-OO 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 µCi/gm DOSE EQUIVALENT 1-131 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

µCi/gm equal to 100 divided by E (average disintegration energy of the sum of the average beta and gamma energies of the coolant nuclides). The limit on DOSE EQUIVALENT 1-131 ensures the thyroid dose to an individual during the Design Basis Accident (OBA) will be (continued)

Farley Units 1 and 2 B 3.4.16-3 Revision 15

RCS Specific Activity B 3.4.16 BASES LCO an appropriate fraction of the allowed thyroid dose. The limit on gross (continued) 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 OBA will be a small fraction of the allowed 1,vhole 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 1-131 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 1-131 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 1-131 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 previsions 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)

Farley Units 1 and 2 B 3.4.16-4 Revision 33

RCS Specific Activity B 3.4.16 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power change ~ 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 E determination is required with the plant operating in MODE 1 equilibrium conditions. The E determination directly relates to the LCO and is required to verify plant operation within the specified gross activity LCO limit. The analysis for E is a measurement of the average energies per disintegration for isotopes with half lives longer than 15 minutes, excluding. 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 E 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.

(continued)

Farley Units 1 and 2 B 3.4.16-6 Revision 52

SG Tube Integrity B 3.4.17 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 l..912.rntAe 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 DOS~ EQUIVALENT 1-131 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 GOG 19 (Ref.

2fr 1o CFR 50.67.+oo (Ref. 3) or the NRG 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)

Farley Units 1 and 2 B 3.4.17-2 Revision 60

SG Tube Integrity B 3.4.17 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, GOG 19Not used.

3. 10 CFR 50.67-1-00.

4. ASME Boiler and Pressure Vessel Code, Section Ill, 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."

Farley Units 1 and 2 B 3.4.17-7 Revision 60

EGGS Recirculation Fluid pH Control System B 3.5.6 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 EGGS 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 ft3 )

and 12,900 pounds (215 ft3 ) 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 OBA could cause the release of radioactive material in containment requiring the operation of the EGGS Recirculation Fluid pH Control System. The EGGS Recirculation Fluid pH Control System assists in reducing the amount of radioactive material available for release to the outside atmosphere after a OBA.

(continued)

Farley Units 1 and 2 B 3.5.6-3 Revision O

MSIVs B 3.7.2 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 50.67.:t-oo (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 MS IVs are closed, they are already performing the safety function.

In MODE 4, normally most of the MSIVs are closed, and the steam 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.

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)

Farley Units 1 and 2 B 3.7.2-4 Revision 0

MSIVs B 3.7.2 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 lnservice Testing Program.

Operating experience has shown that these components usually pass the Surveillance when performed in accordance with the lnservice 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.

Farley Units 1 and 2 B 3.7.2-7 Revision 19 l_

A RVs B 3.7.4 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 50.67-WG and for the case of a pre-accident Iodine spike, the doses from the accident are within the limits defined in 10 CFR 50.67-WG.

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 O

Control RoomCREFS B 3.7.10 B 3.7 PLANT SYSTEMS B 3.7.10 Control RoomCREFS BASES BACKGROUND The control room provides a protected environment from which operators occupants can control the unit following an uncontrolled release of radioactivity, hazardous chemicals, or smoke toxic gas.

This environment is protected by the integrity of the Control Room Envelope (CRE) and the operation of the Control Room Emergency Filtration/Pressurization System (CREFS). The Unit 1 and 2 control room is a common room served by a shared CREFS.

The control room boundary is the oombination of 1.valls, floor, roof, dusting, valves or dampers, ESF HVAC equipment housings, doors, penetrations and equipment that physioally form the CRE. The CRE is the area *11ithin the confines of the control room boundary that contains the spaces that oontrol room operators inhabit to control the plant. This spaoe is protected for normal operation, natural events, and acoident oonditions.

Maintaining the integrity of the CRE minimizes the infiltration of unfiltered air from areas adjacent to the CRE, thereby minimizing the possibility that the effects of a radiological challenge would result in a radiological dose which exceeds General Design Criteria (GDC) 19. It also minimizes the possibility that a fire challenge would result in a condition where the operator would be disabled or impaired such that the reactor could not be controlled from the control room or the hot shutdown panels. In addition, the CRE minimizes the possibility that a hazardous chemical challenge would result in a condition where the operator would be disabled or impaired such that the reactor could not be controlled from the control room. While the CRE provides a boundary for the CREFS to operate in, the CRE is independent from the CREFS and its OPERABILITY requirements are separate from the CREFS.

The CREFS consists of two independent, redundant trains that recirculate and filter the air in the CRE control room air in conjunction with the CRAGS, aAG--two independent, redundant trains that pressurize the control room with filtered outside air, and a CRE boundary that limits the inleakage of unfiltered air. Each filter unit CREFS train consists of a prefilter, a high efficiency particulate air (HEPA) filter, and an activated charcoal adsorber section for removal of gaseous activity (principally iodine). Each pressurization filter also contains a heater. Each train contains filter units, fans, and instrumentation which form the system.

(continued)

Farley Units 1 and 2 B 3.7.10-1 Revision 27

Control RoomCREFS B 3.7.10 BASES BACKGROUND The CRE is the area within the confines of the CRE boundary that (continued) contains the spaces that control room occupants inhabit to control the unit during normal and accident conditions. This area encompasses the control room. and may encompass other non-critical areas to which frequent personnel access or continuous occupancy is not necessary in the event of an accident. The CRE is protected during normal operation. natural events. and accident conditions. The CRE boundary is the combination of walls, floor. roof. ducting, doors.

penetrations and equipment that physically form the CRE. The OPERABILITY of the CRE boundary must be maintained to ensure that the inleakage of unfiltered air into the CRE will not exceed the inleakage assumed in the licensing basis analysis of design basis accident (OBA) consequences to CRE occupants. The CRE and its boundary are defined in the Control Room Envelope Habitability Program.

The CREFS is an emergency system, parts of which may also operate during normal unit operations in the standby mode of operation. Upon receipt of the actuating signal(s), normal air supply to the CRE control room is isolated, and the stream of ventilation air is recirculated through the system filter trains. The prefilters remove any large particles in the air to prevent excessive loading of the HEPA filters and charcoal adsorbers. Operation of each pressurization train for at least 15 minutes per month, with the heaters energized, justifies their OPERABILITY. During operation, the heaters reduce moisture buildup on the HEPA filters and adsorbers. The heater is important to the effectiveness of the charcoal adsorbers.

Actuation of the CREFS places the system in the emergency recirculation mode of operation. Actuation of the system to the emergency recirculation mode of operation, closes the unfiltered outside air intake and unfiltered exhaust dampers, and aligns the system for recirculation of the control room air within the CRE through the redundant trains of HEPA and the charcoal filters. The emergency recirculation mode of operation also initiates pressurization and filtered ventilation of the air supply to the CRE control room.

The normal outside air supply is filtered, diluted with building air from the computer rooms, and added to the control room. The air entering 1 the CRE control room is continuously monitored by radiation detectors. One detector output above the setpoint will cause the control room ventilation to be isolated. The CREFS is then started manually.

(continued)

Farley Units 1 and 2 B3.7.10-2 Revision 27

Control RoomCREFS B3.7.10 BASES BACKGROUND A single CREFS train provides makeup air flow and radiological dose (continued) cleanup for the control room. The CREFS operation in maintaining the control room habitable is discussed in the FSAR, Section 6.4 (Ref. 1).

Redundant supply and recirculation trains provide the required filtration should an excessive pressure drop develop across the other filter train. Normally open isolation dampers are arranged in series pairs so that the failure of one damper to shut will not result in a breach of isolation. The CREFS is designed in accordance with Seismic Category I requirements.

The CREFS is designed to maintain a habitable environment in the CRE the oontrol room environment for 30 days of continuous occupancy after a Design Basis Accident (OBA) without exceeding a 5 rem total effective dose equivalent (TEDE).

An inoperable CRE does not render the CREFS inoperable or vise versa. The OPERABILITY of the CREFS and the CRE are determined separately and both are required to be OPERABLE.

APPLICABLE The CREFS components are arranged in redundant, safety related SAFETY ANALYSES ventilation trains. The location of components within the CRE and ducting of the CRE ensure an adequate supply of filtered air to all areas requiring access. The CREFS provides airborne radiological protection for the CRE occupants oontrol room operators, as

' demonstrated by the CRE oontrol room aooident dose occupant dose analyses for the most limiting design basis loss of ooolant accident, fission product release presented in the FSAR, Chapter 15 (Ref. 2).

Maintaining the integrity of the CRE limits the quantity of contaminants allowed into the CRE so that the radiological dose criteria of GDC 19 are met. The analysis of toxic gas releases demonstrates that the toxicity limits are not exceeded in the control room following a toxic chemical release. The evaluation of a smoke challenge demonstrates that it will not result in the inability of Maintaining the integrity of tho CRE helps to ensure that the CRE occupants oontrol room operators may !Q_maintain reactor control either from the control room af\G maintain separation between the oontrol room and or from the hot shutdown panels.

The worst case single active failure of a component of the CREFS, assuming a loss of offsite power, does not impair the ability of the system to perform its design function.

(continued)

Farley Units 1 and 2 B3.7.10-3 Revision 27 I

L__

Control RoomCREFS B3.7.10 BASES APPLICABLE The CREFS satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

SAFETY ANALYSES (continued)

LCO Two independent and redundant CREFS trains are required to be OPERABLE to ensure that at least one is available assuming jf_a single active failure disables the other train. Total system failure, such as from a loss of both ventilation trains or from an inoperable CRE boundary, could result in exceeding a dose of 5 rem TEDE to the CRE occupants control room operator in the event of a large radioactive release.

+AeEach CREFS train is considered OPERABLE when the individual components necessary to limit CRE occupant operator exposure are OPERABLE in both trains. A CREFS train is OPERABLE when the associated:

a. Fans are OPERABLE; (recirculation, filtration, Pressurization, and CRAGS Fans)

b. HEPA filters and charcoal adsorbers are not excessively restricting flow, and are capable of performing their filtration functions; and

c. Heater is OPERABLE and air circulation can be maintained.

In addition, the CRE must be maintained OPERABLE, including the integrity of the *.valls, floors, ceilings, ductwork, valves and dampers, ESF HVAC equipment housings, and access doors. lnleakage must also be minimized such that operator exposure limits are not exceeded.

In order for the CREFS trains to be considered OPERABLE, the CRE boundary must be maintained such that the CRE occupant dose from a large radioactive release does not exceed the calculated dose in the licensing basis consequence analyses for DBAs, and that CRE occupants are protected from hazardous chemicals and smoke.

An inoperable CRE does not render the CREFS inoperable or vice versa. The OPERABILITY of the CREFS and the CRE are determined separately and both are required to be OPERABLE.

The LCO is modified by a Note allowing the CRE to be opened intermittently under administrative controls without requiring entry into (continued)

Farley Units 1 and 2 B3.7.10-4 Revision 27

Control RoomCREFS B3.7.10 BASES LCO Condition B for an inoperable CRE. This Note only applies to opening (continued) in the CRE boundary that can be rapidly restored to the design condition, such as doors, hatches. floor plugs, and access panels.

For entry and exit through doors, the administrative control of the opening is performed by the person(s) entering or exiting the area.

For maintenance access openings, such as hatches and test ports, the administrative control of the opening is performed by the attendant person(s) performing the maintenance. For other openings, these controls should be proceduralized and consist of stationing a dedicated individual at the opening who is in continuous communication with the operators in the CREoontrol room. This individual will have a method to rapidly close the opening and to restore the CRE boundary to a condition equivalent to the design condition when a need for CREcontrol room integrity is indicated.

APPLICABILITY With either unit in MODES 1, 2, 3, or 4 or during movement of irradiated fuel assemblies or during CORE ALTERATIONS, the CREFS and the CRE must be OPERABLE to ensure that the CRE will remain habitable oontrol operator exposure during and following a OBA.

During movement of irradiated fuel assemblies and CORE ALTERATIONS, the CREFS and the CRE must be OPERABLE to cope with the release from a fuel handling accident.

ACTIONS With one CREFS train inoperable, for reasons other than an inoperable CRE boundary, action must be taken to restore it to OPERABLE status within 7 days. In this Condition, the remaining OPERABLE CREFS train is adequate to perform the CRE occupant oontrol room protection function. However, the overall reliability is reduced because a single failure in the OPERABLE CREFS train could result in loss of CREFS function. The 7 day Completion Time is based on the. low probability of a OBA occurring during this time period, and ability of the remaining train to provide the required capability.

(continued)

Farley Units 1 and 2 B 3.7.10-5 Revision 27

Control RoomCREFS B3.7.10 BASES ACTIONS B.1, B.2.1, B.2.2.1, and B.32-.2-.2 (continued)

If the unfiltered inleakage of potentially contaminated air past the CRE boundary and into the CRE can result in CRE occupant radiological dose greater than the calculated dose of the licensing basis analyses of OBA consequences (allowed to be up to 5 rem TEDE), or inadequate protection of CRE occupants from hazardous chemicals or smoke. the CRE boundary is inoperable. Actions must be taken to restore an OPERABLE CRE boundary within 90 days.

During the period that the CRE boundary is considered inoperable, action must be initiated to implement mitigating actions to lessen the effect on CRE occupants from the potential hazards of a radiological or chemical event or a challenge from smoke. Actions must be taken within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to verify that in the event of a OBA. the mitigating actions will ensure that CRE occupant radiological exposures will not exceed the calculated dose of the licensing basis analyses of OBA consequences, and that CRE occupants are protected from hazardous chemicals and smoke. These mitigating actions (i.e.,

actions that are taken to offset the consequences of the inoperable CRE boundary) should be preplanned for implementation upon entry into the condition, regardless of whether entry is intentional or unintentional. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is reasonable based on the low probability of a OBA occurring during this time period, and the use of mitigating actions. The 90 day Completion Time is reasonable based on the determination that the mitigating actions will ensure protection of CRE occupants within analyzed limits while limiting the probability that CRE occupants will have to implement protective measures that may adversely affect their ability to control the reactor and maintain it in a safe shutdown condition in the event of a OBA. In addition, the 90 day Completion Time is a reasonable time to diagnose, plan and possibly repair, and test most problems with the CRE boundary.

If the CRE is inoperable, the operator protection analyses assumption of inleal<age may be e>roeeded. During the period that the CRE is inoperable, mitigating actions must be initiated to protect control room operators from potential hazards. These mitigating actions (i.e.,

actions that are taken to offset the consequences of the inoperable CRE) should be preplanned for initiation upon entry into the condition.

'Nithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of entry into Condition B, Actions must be taken to restore the CRE to OPERABLE status or to verify that the requirements of GOG 19 are met for the facility. GOG 19 is verified to be met by limiting dose from radioactive gas and particulates, and exposure to toxic gas and smoke, to levels that support control room (continued)

Farley Units 1 and 2 - B 3.7.10-6 Revision 27

Control RoomCREFS B3.7.10 BASES ACTIONS habitability, crediting, as necessary, the mitigating actions required by (continued) Required Action B.1. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is reasonable based

on the lmv probability of a DBA oeeurring d-uring this time period, the use of mitigating actions, and the time necessary to perform an assessment.

If it is determined that the requirements of GOG 19 are met crediting, as necessary, the mitigating actions required by Required Action B.1, 30 days are provided to return the CRE to OPERABLE status. The 30 day Completion Time is a reasonable time to diagnose, plan, and repair most problems 'Nith the CRE.

C.1 and C.2 In MODE 1, 2, 3, or 4, if an inoperable CREFS train or CRE cannot be restored to OPERABLE status within the required Completion Time, the unit must be placed in a MODE in which overall plant risk is reduced. 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 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Remaining within the applicability of the LCO is acceptable to accomplish short duration repairs to restore inoperable equipment because the plant risk in MODE 4 is similar to or lower than MODE 5 (Ref. 5). In MODE 4 the Steam Generators and Residual Heat Removal System are available to remove decay heat, which provides diversity and defense in depth. As stated in Reference 5, the steam turbine driven Auxiliary Feedwater Pump must be available to remain in MODE 4. Should Steam Generator cooling be lost while relying on this Required Action, there are preplanned actions to ensure long-term decay heat removal.

Voluntary entry into MODE 5 may be made as it is also acceptable from a risk perspective.

Required Action C.2 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 4. This Note prohibits the use of LCO 3.0.4.a to enter MODE 4 during startup with the LCO not met.

However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 4, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. 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.

(continued)

I Farley Units 1 and 2 B3.7.10-7 Revision 52 I*

Control RoomCREFS B 3.7.10 BASES ACTIONS D.1 and D.2 (continued)

If two CREFS trains are inoperable in MODE 1, 2, 3, or 4, the unit must be placed in a MODE that minimizes accident risk. 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 Time is reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

E.1, E.2.1, and E.2.2 During movement of irradiated fuel assemblies or during CORE ALTERATIONS, if an inoperable CREFS train cannot be restored to OPERABLE status within the required Completion Time, action must be taken to immediately place the OPERABLE CREFS train in the emergency recirculation mode. This action ensures that the remaining train is OPERABLE, that no failures preventing automatic actuation will occur, and that any active failure would be readily detected.

An alternative to Required Action D.1 is to immediately suspend activities that could result in a release of radioactivity that might require isolation of the CREoontrol room. This places the unit in a condition that minimizes the accident risk. This does not preclude the movement of fuel to a safe position.

F.1 and F.2 During movement of irradiated fuel assemblies or during CORE ALTERATIONS, if an inoperable CRE oannot be restored to OPERABLE status within the required Completion Time or with two CREFS trains inoperable or with one or more CREFS trains inoperable due to an inoperable CRE boundary, action must be taken to immediately suspend activities that could result in a release of radioactivity that might require isolation of the CREoontrol room. This places the unit in a condition that minimizes accident risk. This does not preclude the movement of fuel to a safe position.

(continued)

Farley Units 1 and 2 B3.7.10-8 Revision 52

Control RoomCREFS B3.7.10 BASES SURVEILLANCE SR 3.7.10.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. As the environment and normal operating conditions on this system are not severe, testing each train (CREFS and Pressurization) once every month provides an adequate check of this system. The CREFS trains are initiated from the control room with flow through the HEPA and charcoal filters. Systems must be operated for~

15 minutes to demonstrate the function of the system (Ref. 3). Systems with heaters must be operated with the heaters energized. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.7.10.2 This SR verifies that the required CREFS testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The CREFS filter tests are in accordance with ASME N510-1989 (Ref. 4).

The VFTP includes testing the performance of the HEPA filter, charcoal adsorber efficiency, flow rate, and the physical properties of the activated charcoal. Specific test Frequencies and additional information are discussed in detail in the VFTP.

SR 3.7.10.3 This SR verifies that each CREFS train starts and operates on an actual or simulated Safety Injection (SI) actuation signal. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. This SR is modified by a note which provides an exception to the requirement to meet this SR in MODES 5 and 6. This is acceptable since the automatic SI actuation function is not required in these MODES.

SR 3.7.10.4 This SR verifies the OPERABILITY of the CRE boundary by testing for unfiltered air inleakage past the CRE boundary and into the CRE. The details of the testing are specified in the Control Room Envelope Habitability Program.

The CRE is considered habitable when the radiological dose to CRE occupants calculated in the licensing basis analyses of OBA consequences is no more than 5 rem TEDE and the CRE occupants are protected from hazardous chemicals and smoke. This SR verifies that the unfiltered air inleakage into the CRE is no greater than the flow rate (continued)

Farley Units 1 and 2 B 3.7.10-9 Revision 52

Control RoomCREFS B 3..7.10 BASES SURVEILLANCE SR 3.7.10.4 (continued)

REQUIREMENTS assumed in the licensing basis analyses of OBA consequences. When unfiltered air inleakage is greater than the assumed flow rate, Condition B must be entered. Required Action B.3 allows time to restore the CRE boundary to OPERABLE status provided mitigating actions can ensure that the CRE remains within the licensing basis habitability limits for the occupants following an accident. Compensatory measures are discussed in Regulatory Guide 1.196, Section C.2.7.3, (Ref. 6) which endorses, with exceptions, NEI 99-03, Section 8.4 and Appendix F (Ref. 7). These compensatory measures may also be used as mitigating actions as required by Required Action B.2. Temporary analytical methods may also be used as compensatory measures to restore OPERABILITY (Ref. 8). Options for restoring the CRE boundary to OPERABLE status include changing the licensing basis OBA consequence analysis, repairing the CRE boundary, or a combination of these actions. Depending upon the nature of the problem and the corrective action, a full scope inleakage test may not be necessary to establish that the CRE boundary has been restored to OPERABLE status.

This SR verifies that the CRE Ap can be maintained 'J'Jithin limits defined in the Control Room Integrity Program (GRIP) with one CREFS train in operation. If the requirements of this SR cannot be met, a determination must be made as to the cause of the failure. Once identified, the appropriate Condition (for eit~er the CREFS or the CRE) must be entered. For example, if the failure is due to a breach in the integrity of the CRE, the Condition for an inoperable CRE would be entered but the Condition for an inoperable CREFS would not be entered.

An inoperable CRE does not render the CREFS inoperable or vice versa. The frequency of 24 months on a STAGGERED TEST BASIS is adequate and has been sho'J':n to be acceptable by operating experience.

Any change in the components being tested by this SR will require reevaluation of STI Evaluation Number 558904 in aecordance v:ith the Surveillance Frequency Control Program.

SR 3.7.10.5 This SR verifies the integrity of the CRE by requiring testing for control room inleakage. The details of the inleakage testing are contained in the GRIP.

(continued)

Farley Units 1 and 2 B 3.7.10-10 Revision 52

Control RoomCREFS 83.7.10 BASES REFERENCES 1. FSAR, Section 6.4.

2. FSAR, Chapter 15.

3. Regulatory Guide 1.52, Rev. 3.

4. ASME N510-1989.

5. WCAP-16294-NP-A, Rev. 1, "Risk-Informed Evaluation of Changes*

to Technical Specification Required Action Endstates for Westinghouse NSSS PWRs," June 2010.

6. Regulatory Guide 1.196

7. NEI 99-03, "Control Room Habitability Assessment." June 2001

8. Letter from Eric J. Leeds (NRC) to James W. Davis (NEI) dated January 30, 2004, "NEI Draft White Paper. Use of Generic Letter 91-18 Process and Alternative Source Terms in the Context of Control Room Habitability." (ADAMS Accession No. ML040300694)

Farley Units 1 and 2 83.7.10-11 Revision 52

PRF B 3.7.12 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 OBA 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 1.1.1832:§ (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 50.67.+oo (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)

Farley Units 1 and 2 B 3.7.12-2 Revision O

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PRF 8 3.7.12 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 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 1.1832§.

5. 10 CFR 50.67.+oo.

6. ASME N510-1989.

Farley Units 1 and 2 8 3.7.12-7 Revision 52

Fuel Storage Pool Water Level B3.7.13 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 1.1832§ (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 50.67.:t-OG (Ref. 5) limits.

According to Reference 4, there is 23 ft of water between the top of the damaged fuel bundle and the fuel poof 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).

Farley Units 1 and 2 B 3.7.13-1 Revision O

Fuel Storage Pool Water Level 83.7.13 BASES SURVEILLANCE SR 3.7.13.1 (continued)

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 1.1832§, Rev. 0.

5. 10 CFR 50.67100.11.

Farley Units 1 and 2 B 3.7.13-3 Revision 52

Secondary Specific Activity B3.7.16 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, 1-131 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 µCi/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 50.67.:t.OO (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 µCi/gm DOSE EQUIVALENT 1-131. This assumption is used in the analysis for determining the radiological (continued)

Farley Units 1 and 2 B 3.7.16-1 Revision 5

Secondary Specific Activity B3.7.16 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 cool down.

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 µCi/gm DOSE EQUIVALENT 1-131 to limit the radiological consequences of a Design Basis Accident (OBA) 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 OBA.

Farley Units 1 and 2 B 3.7.16-2 Revision 0

_J

Secondary Specific Activity B 3.7.16 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 1-131 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 1-131, 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.

Farley Units 1 and 2 B 3.7.16-3 Revision 52

Containment Penetrations B 3.9.3 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 50.67.+GG. 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)

Farley Units 1 and 2 B 3.9.3-1 Revision 44

Containment Penetrations B 3.9.3 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 thanv1ell *.vithin the dose limitsguideline values specified in 10 CFR 50.67-+00-, and the more restrictive offsite exposure criteria of..-- Standard Review Plan, Section 15.0.17.4, Rev. 1 (Ref. 3), defines '\veil within" 10 CFR 100 to be 25% or less of the 10 CFR 100 values. The aooeptanoe 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)

Farley Units 1 and 2 B3.9.3-6 Revision 52

Containment Penetrations B 3.9.3 BASES LCO are terminated, such that radiological doses are within the acceptance (continued) limit.

The equipment hatch and personnel air locks are considered isolable when the following criteria are satisfied:

1. the necessary equipment required to close the hatch and personnel air locks is available,

2. at least 23 feet of water is maintained over the top of the reactor vessel flange in accordance with Specification 3.9.6,

3. a designated trained closure crew is available.

The equipment hatch and personnel air locks door openings must be capable of being cleared of any obstruction so that closure can be achieved as soon as possible.

The containment personnel air lock and emergency personnel air lock doors may be open during movement of irradiated fuel in the containment and during CORE ALTERATIONS provided that one door in each air lock is capable of being closed in the event of a fuel handling accident. Should a fuel handling accident occur inside containment, one door in each personnel air lock will be closed following an evacuation of containment.

The closure of the equipment hatch and the personnel air locks will be completed promptly following a fuel handling accident within containment.

The LCO is modified by a Note allowing penetration flow paths with direct access from the containment atmosphere to the outside atmosphere to be unisolated under administrative controls.

Administrative controls ensure 1) appropriate personnel are aware of the open status of the penetration flow path during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, and 2) special individuals are designated and readily available to isolate the flow path in the event of a fuel handling accident.

APPL! GABI LITY The containment penetration requirements are applicable during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment because this is when there is a potential for a fuel handling accident. In MODES 1, 2, 3, and 4, containment penetration requirements are addressed by LCO 3.6.1. In MODES 5 and 6, when CORE ALTERATIONS or movement of irradiated fuel assemblies within containment are not being conducted, the potential for a fuel Farley Units 1 and 2 B 3.9.3-21 Revision 44

Containment Penetrations B 3.9.3 BASES SURVEILLANCE SR 3.9.3.2 (continued)

REQUIREMENTS isolation time of each valve is in accordance with the lnservice 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. Q+, July 2000+98-+.

4. Regulatory Guide 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," May 2003.

Farley Units 1 and 2 B 3.9.3-22 Revision 44

Refueling Cavity Water Level B 3.9.6 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 50.67.:tW 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 1.1832§ (Ref. 1). A minimum water level of 23 ft (Regulatory Position C.1.c of Ref. 1) allows a decontamination factor of g+oo (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 oavity 'Nater.

The fuel pellet to oladding gap is assumed to oontain 10% (exoept 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).

Farley Units 1 and 2 B 3.9.6-23 Revision O

Refueling Cavity Water Level 8 3.9.6 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-A.

4. 10 CFR 50.67100.10.

5. Malinmvski, 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 8 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 Units 1 and 2 8 3.9.6-3 Revision 52

(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 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.

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

(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 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.

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

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 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 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.

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

I_

(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 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.

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.

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.

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

I 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.

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 1 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-14-1273, 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. 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 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.

Farley - Unit 2 Renewed License No. NPF-8 Amendment No.

(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 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.

(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.

Farley - Unit 2 Renewed License No. NPF-8 Amendment No.

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.

(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.

Farley - Unit 2 Renewed License No. NPF-8 Amendment No.

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

Farley - Unit 2 Renewed License No. NPF-8 Amendment No.

- - ~

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

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 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-8 Amendment No.

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. 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 fife 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 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 Farley - Unit 2 Renewed License No. NPF-8 Amendment No.

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.

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 Farley - Unit 2 Renewed License No. NPF-8 Amendment No.

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

LCO 3.7.10 Two CREFS trains shall be OPERABLE.

N 0 TE -----------------------------------------------

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 A.1 Restore CREFS train to 7 days inoperable for reasons OPERABLE status.

other than Condition B.

B. CRE or more CREFS B.1 Initiate action to implement Immediately trains inoperable due to mitigating actions.

inoperable CRE boundary. AND B.2. Verify mitigating actions 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ensure CRE occupant exposures to radiological, chemical, and smoke hazards will not exceed limits.

AND B.3 Restore CRE boundary to 90 days OPERABLE status.

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

Amendment No. (Unit 2)

CREFS 3.7.10 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or B AND not met in MODE 1, 2, 3, or4.

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

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

Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D. Two CREFS trains D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months inoperable in MODE 1, 2, 3, OR4.

AND D.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months E. Required Action and E.1 Place OPERABLE CREFS Immediately associated Completion train in emergency Time of Condition A not recirculation mode.

met during movement of irradiated fuel assemblies OR or during CORE ALTERATIONS. E.2.1 Suspend CORE Immediately ALTERATIONS.

AND E.2.2 Suspend movement of Immediately irradiated fuel assemblies.

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

Amendment No. (Unit 2)

CREFS 3.7.10 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F. Required Action and F.1 Suspend CORE Immediately associated Completion ALTERATIONS.

Time of Condition B not met during movement of AND irradiated fuel assemblies or during CORE F.2 Suspend movement of Immediately ALTERATIONS. irradiated fuel assemblies.

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

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

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.1 Operate each CREFS Pressurization train with the In accordance with heaters operating and each CREFS Recirculation and the Surveillance Filtration train for~ 15 minutes. Frequency Control Program SR 3.7.10.2 Perform required CREFS filter testing in accordance In accordance with with the Ventilation Filter Testing Program (VFTP). the VFTP Farley Units 1 and 2 3.7.10-3 Amendment No. (Unit 1)

Amendment No. (Unit 2)

CREFS 3.7.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.3 -----------------------------N 0 TE-------------------------------- In accordance N ot required to be performed in MODES 5 and 6. with the Surveillance Verify each CREFS train actuates on an actual or Frequency simulated actuation signal. Control Program SR 3.7.10.4 Perform required CRE unfiltered air inleakage testing in In accordance accordance with the Control Room Envelope with the Control Habitability Program. Room Envelope Habitability Program Farley Units 1 and 2 3.7.10-4 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Containment Penetrations 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Containment Penetrations LCO 3.9.3 The containment penetrations shall be in the following status:

a. The equipment hatch is capable of being closed and held in place by four bolts;

b. One door in each air lock is capable of being closed; and

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:

1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or

2. capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System.

N()l"E:

Penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative control.

APPLICABILITY: During CORE ALTERATIONS, During movement of irradiated fuel assemblies Within containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more containment A.1 Suspend CORE Immediately penetrations not in ALTERATIONS.

required status.

A.2 Suspend movement of Immediately irradiated fuel assemblies within containment.

Farley Units 1 and 2 3.9.3-1 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Programs and Manuals 5.5 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 (OBA) 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 O,

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 OBA 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.

(continued)

Farley Units 1 and 2 5.5-15 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Programs and Manuals 5.5 5.5 Programs and Manuals 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.

Farley Units 1 and 2 5.5-16 Amendment No. (Unit 1)

Amendment No. (Unit 2) to NL-16-0388 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-16-0388 Regulatory Guide 1.183 Conformance Tables REGULATORY GUIDE 1.183 COMFORMANCE TABLES The proposed uses of an AST and the associated proposed facility Conforms- Adequate safety margins modifications and changes to procedures should be evaluated to determine are maintained, as discussed in the No whether the proposed changes are consistent with the principle that sufficient Significant Hazards Consideration.

safety margins are maintained, including a margin to account for analysis Future changes will be evaluated under uncertainties. The safety margins are products of specific values and limits the provisions of 10 CFR 50.59.

contained in the technical specifications (which cannot be changed without NRG 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 NRG 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 mar ins related to subse uent facilit modifications and chan es to rocedures.

1.1.2 The proposed uses of an AST and the associated proposed facility Conforms - There are no facility modifications and changes to procedures should be evaluated to determine modifications being proposed to whether the proposed changes are consistent with the principle that adequate implement AST, and compliance with defense in depth is maintained to compensate for uncertainties in accident the GDCs are maintained. No new progression and analysis data. Consistency with the defense-in-depth reliance is placed on compensatory philosophy is maintained if system redundancy, independence, and diversity are programmatic actions (including preserved commensurate with the expected frequency, consequences of manual operator actions) to maintain challenges to the system, and uncertainties. In all cases, compliance with the adequate defense-in-depth.

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 Not Applicable - There are no safeguards equipment should be evaluated to be sure that the modification modifications being proposed with this does not invalidate assumptions made in facility PRAs and does not adversely License Amendment Request.

impact the facility's severe accident management program.

ES - 1

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

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section 1.1.3 The design basis accident source term is a fundamental assumption upon which Conforms - See RG Section 1.3 a significant portion of the facility design is based. Additionally, many aspects of discussions.

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 updatinQ in the future as additional modifications are performed.

1.1.3 This approach would create two tiers of analyses, those based on the previous Conforms - This is a full scope AST source term and those based on an AST. The radiological acceptance criteria implementation for the radiological would also be different with some analyses based on whole body and thyroid dose consequences of the FNP Design criteria and some based on TEDE criteria. Full implementation of the AST Basis Accidents.

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 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 Conforms- This License Amendment that are consistent with corresponding data used in other design basis safety Request includes re-evaluation of the analyses, radiological and nonradiological, unless these data would result in radiological consequences of the most nonconservative results or otherwise conflict with the guidance in this guide. severe DBAs. It relies on assumptions and inputs that do not create a conflict with, or render non-conservative, other design basis safety analyses.

ES-2

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

-

Table A: Conformance With Regulator\/ Guide 1".183 Seetiqn C

,~G

~section RG Position FN P Analysis .

1.1.4 Although the AST provided in this guide was based on a limited spectrum of Conforms - No changes are proposed severe accidents, the particular characteristics have been tailored specifically in this License Amendment Request to for OBA analysis use. The AST is not representative of the wide spectrum of Emergency Preparedness possible events that make up the planning basis of emergency preparedness. requirements.

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.

1.2.1 Full implementation is a modification of the facility design basis that addresses Conforms - This License Amendment all characteristics of the AST, that is, composition and magnitude of the Request involves recalculation of the radioactive material, its chemical and physical form, and the timing of its dose consequences of the most severe release. Full implementation revises the plant licensing basis to specify the AST OBAs. The characteristics of the AST in place of the previous accident source term and establishes the TEOE dose as methods are addressed in the the new acceptance criteria. This applies not only to the analyses performed in recalculations. The OBA LOCA has the application (which may only include a subset of the plant analyses), but also been re-analyzed per Appendix A.

to all future design basis analyses. At a minimum for full implementations, the OBA 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 TEOE 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 Not Applicable - This License based on one or more of the characteristics of the AST or (2) entails re- Amendment Request is for full scope evaluation of a limited subset of the design basis radiological analyses. The AST implementation for the radiological NRC staff will allow licensees flexibility in technically justified selective dose consequences of the major FNP implementations provided a clear, logical, and consistent design basis is OBA.

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 E5- 3

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables Table A: *.Conformance .With Requlatorv Guide 1.1.83 Section C

RG

~ s"~ctlon ". " RG Position , FNP Analysis 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 analysis guidance is provided in Regulatory Position 1.3 of this guide. NRG 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 OBA LOCA. However, this licensee could extend use of the timing characteristic to adjust the closure time on isolation valves not included in the oriqinal approval.

1.3.1 There are several regulatory requirements for which compliance is Conforms- This full scope AST License demonstrated, in part, by the evaluation of the radiological consequences of Amendment Request is salient to: a) design basis accidents. These requirements include, but are not limited to, the Control Room Habitability (GDC 19 and following. NUREG-0737 Item 111.D.3.4), b) AST

Environmental Qualification of Equipment (1 O CFR 50.49) (1 O CFR 50.67), and c) Facility Siting

Control Room Habitability (GDC 19 of Appendix A to 10 CFR Part 50) (1 O CFR 100.11 ). Control Room

Emergency Response Facility Habitability (Paragraph IV.E.8 of Habitability and compliance with the Appendix E to 10 CFR Part 50) Alternative Source Term requirements

Alternative Source Term (10 CFR 50.67) are the principal subjects of this

Environmental Reports (10 CFR Part 51) submittal and are discussed in

Facility Siting (1 O CFR 100.11 )5 Sections 3 and 4 of this License There may be additional applications of the accident source term identified in Amendment Request.

the technical specification bases and in various licensee commitments. These include, but are not limited to, the following from Reference 2, NUREG-0737. Regarding Emergency Response Facility Habitability, FNP will continue E5 - 4

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

'~t;;i<3.' .~

~;section* *.

Post-Accident Access Shielding (NUREG-0737, 11.B.2) to meet the NUREG-0654 Planning

Post-Accident Sampling Capability (NUREG-0737, 11.B.3) Standard for Emergency Facilities and

Accident Monitoring Instrumentation (NUREG-0737, 11.F.1) Equipment as described in the FNP

Leakage Control (NUREG-0737, 111.D.1.1) Emergency Plan. Design Basis dose

Emergency Response Facilities (NUREG-0737, 111.A.1.2) calculations for non-control room

Control Room Habitability (NUREG-0737, 111.D.3.4) Emergency Response Facilities, such as the Technical Support Center, are not part of the FNP current licensing basis. The Emergency Response Facilities continue to meet NUREG-0696 habitability requirements.

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 Conforms- The License Amendment modification should be supported by evaluations of all significant radiological Request for this full scope application and nonradiological impacts of the proposed actions. This evaluation should of the AST evaluated the impact of the consider the impact of the proposed changes on the facility's compliance with proposed change against the Current the regulations and commitments listed above as well as any other facility- Licensing Basis, mitigating system specific requirements. These impacts may be due to (1) the associated facility design basis requirements, and modifications or (2) the differences in the AST characteristics. The scope and Technical Specifications. No facility extent of the re-evaluation will necessarily be a function of the specific proposed modifications are proposed as part of facility modification 6 and whether a full or selective implementation is being this License Amendment Request and pursued. The NRG staff does not expect a complete recalculation of all facility compliance with regulations and radiological analyses, but does expect licensees to evaluate all impacts of the commitments are maintained.

proposed changes and to update the affected analyses and the design bases appropriately. An analysis is considered to be affected if the proposed modification chan es one or more assum tions or in uts used in that anal sis E5 - 5

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

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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 TEOE 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 NRG staff has performed an evaluation of the impact of the AST on three Conforms- There are no plant representative operating reactors (Ref. 14). This evaluation determined that modifications that are planned to radiological analysis results based on the TI0-14844 source term assumptions implement the AST analyses. The (Ref. 1) and the whole body and thyroid methodology generally bound the radiological and nonradiological results from analyses based on the AST and TEOE methodology. Licensees impacts of full scope implementation of may use the applicable conclusions of this evaluation in addressing the impact the AST have been considered and of the AST on design basis radiological analyses. However, this does not discussed in the License Amendment exempt the licensee from evaluating the remaining radiological and Request, as applicable.

nonradiological impacts 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 OBA LOCA analysis as described in Conforms - The OBA LOCA analysis is Appendix A of this guide should be performed, as a minimum. Other design provided in this License Amendment basis analyses are updated in accordance with the guidance in this section. Request which is consistent with Appendix A.

ES - 6 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

.,:Guide J. 183 Sectiori C 1.3.2 A selective implementation of an AST and any associated facility modification Not Applicable - This License based on the AST should evaluate all the radiological and nonradiological Amendment Request is a full scope impacts of the proposed actions as they apply to the particular implementation. AST implementation that evaluates the Design basis analyses are updated in accordance with the guidance in this dose consequences of the most severe section. There is no minimum requirement that a OBA LOCA analysis be FNP DBAs.

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 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 Not Applicable-The FNP AST analysis existing analyses have sufficient margin and need not be recalculated. As used does not rely on sensitivity or scoping in this guide, a sensitivity analysis is an evaluation that considers how the analyses.

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 post accident 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 anal ses the whole bod and th roid doses determined with E5 -7 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

. Table A: Conformance With ReQulatorv Guide *1.183.Section G RG ,, " ...

Section . RG:Posjtion, * .c .. ' *' , FNP Analvsis 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 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 Not Applicable- The FNP AST design the approved AST and the TEDE criteria for all design basis radiological basis radiological analyses do not rely analyses. The implementation may have been supported in part by sensitivity or on sensitivity or scoping analyses.

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 1.3.4 This guidance is also applicable to selective implementations to the extent that Not Applicable - This is a full scope the affected analyses are within the scope of the approved implementation as License Amendment Request that described in the facility design basis. In these cases, the characteristics of the evaluates the dose consequences of AST and TEDE criteria identified in the facility design basis need to be the most severe FNP DBAs.

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.

E5 - 8

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables Table A: Conformance With Regulatory Guide 1.183 Section C HG '.

,section RG Position. * * * .*. . **' *. .

FNP Analysis * .*.

1.3.5 Current environmental qualification (EQ) analyses may be impacted by a Conforms -The FNP AST License proposed plant modification associated with the AST implementation. The EQ Amendment Request is not proposing analyses that have assumptions or inputs affected by the plant modification to modify the equipment qualification should be updated to address these impacts. The NRG staff is assessing the design basis to adopt AST. The FNP effect of increased cesium releases on EQ doses to determine whether licensee EQ analysis will continue to be based action is warranted. Until such time as this generic issue is resolved, licensees on TID-14844 assumptions.

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.

1.4 The use of an AST changes only the regulatory assumptions regarding the Not Applicable - No facility analytical treatment of the design basis accidents. The AST has no direct effect modifications are proposed or planned on the probability of the accident. Use of an AST alone cannot increase the core as implementation actions of the FHA damage frequency (GDF) or the large early release frequency (LERF). AST analysis.

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 PRAs should be evaluated.

1.4 Consideration should be given to the risk impact of proposed implementations Not Applicable- The FNP AST License that seek to remove or downgrade the performance of previously required Amendment Request is not seeking to engineered safeguards equipment on the basis of the reduced postulated remove or downgrade the performance doses. The NRG staff may request risk information if there is a reason to of previously required engineered question adequate protection of public health and safety. 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 Not Applicable-The FNP AST License the design basis that are not addressed in currently approved NRG staff Amendment Request is not utilizing risk positions. For guidance, refer to Regulatory Guide 1.174, "An Approach for insights as a basis for any proposed Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant- changes.

Specific Changes to the Licensing Basis" (Ref. 15).

E5- 9

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

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1.5 According to 10 CFR 50.90, an application for an amendment must fully Conforms- The License Amendment describe the changes desired and should follow, as far as applicable, the form Request is formatted in accordance prescribed for original applications. Regulatory Guide 1.70, "Standard Format with accepted NRG/industry guidance.

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

1.5 If the licensee has used a current approved version of an NRG-sponsored The LOCA dose calculation was computer code, the NRC staff review can be made more efficient if the licensee performed using RADTRAD 3.10. The identifies the code used and submits the inputs that the licensee used in the FHA dose calculations uses RADTRAD calculations made with that code. In many cases, this will reduce the need for Version 3.03.

NRC staff confirmatory analyses. This recommendation does not constitute a requirement that the, licensee use NRG-sponsored computer codes. The MSLB, Control Rod Ejection, 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 ES -10

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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 Conforms- Approval of this License 10 CFR 50.71, "Maintenance of Records, Making of Reports." The regulations Amendment Request will result in the in 10 CFR 50.71 (e) require that the FSAR be updated to include all changes necessary revisions to the FSAR, with made in the facility or procedures described in the FSAR and all safety revised FSAR pages submitted evaluations performed by the licensee in support of requests for license pursuant to 10 CFR 50.71 (e).

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 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 analY.ses 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 Conforms- This License Amendment design analyses or consideration of possible accidental events, that could result Request applies the AST methods in hazards not exceeded by those from other accidents considered credible. when evaluating the dose The AST must address events that involve a substantial meltdown of the core consequences of the most severe with the subsequent release of appreciable Quantities of fission products. DBAs applicable to FNP.

ES - 11 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

-- Table A: Conformance With Regulatory Guide 1.183 Section C RG Section RG Position FN P Analysis 2.2 The AST must be expressed in terms of times and rates of appearance of Conforms - For the DBAs that release radioactive fission products released into containment, the types and quantities to Containment (LOCA, FHA, and of the radioactive species released, and the chemical forms of iodine released. 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 Conforms- This License Amendment represent a spectrum of credible severe accident events. Risk insights may be Request considers a number of release used, not to select a single risk-significant accident, but rather to establish the scenarios, as applicable, for the DBAs range of events to be considered. Relevant insights from applicable severe being revised for use of AST. The most accident research on the phenomenology of fission product release and limiting of these releases are analyzed transport behavior may be considered. for radiological consequences.

2.4 The AST must have a defensible technical basis supported by sufficient Conforms- The OBA AST dose experimental and empirical data, be verified and validated, and be documented calculations have been developed in a scrutable form that facilitates public review and discourse. based on NUREG-1465 and this Regulatory Guide. The calculations, which utilizes RADTRAD and Bechtel LocaDose were developed in accordance with 10 CFR 50 Appendix B, Criterion Ill.

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

E5 - 12

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

>~;:: ** Tq.bJe A: Conformance With Regu1!3,torv'Guide. :1. t8~.Sectio6.C I*

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3.1 The inventory of fission products in the reactor core and available for release to Conforms -The FNP OBAs that release the containment should be based on the maximum full power operation of the to the Containment are the LOCA, core with, as a minimum, current licensed values for fuel enrichment, fuel FHA, and Control Rod Ejection. Core burnup, and an assumed core power equal to the current licensed rated thermal Inventory has been determined using power times the EGGS evaluation uncertainty.a The period of irradiation should an appropriate isotope generation and be of sufficient duration to allow the activity of dose-significant radionuclides to depletion code, such as ORIGEN2 or reach equilibrium or to reach maximum values.9 The core inventory should be ORIG EN-ARP.

determined using an appropriate isotope generation and depletion computer code such as ORIG EN 2 (Ref. 17) or ORIG EN-ARP (Ref. 18). Core inventory factors (Ci/MWt) provided in TI014844 and used in some analysis computer codes were derived for low burnup, low enrichment fuel and should not be used with hioher burnup and hiqher enrichment fuels.

3.1 For the OBA LOCA, all fuel assemblies in the core are assumed to be affected With the exception of OBAs where and the core average inventory should be used. For OBA events that do not cladding damage is postulated with a involve the entire core, the fission product inventory of each of the damaged fuel gap release, the analyses of events rods is determined by dividing the total core inventory by the number of fuel which involve fuel damage assume that rods in the core. To account for differences in power level across the core, the entire core is affected with a source radial peaking factors from the facility's core operating limits report (COLR) or term based upon full power, core technical specifications should be applied in determining the inventory of the average conditions. The source term damaged rods. for OBAs where cladding damage is postulated with a gap release 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 The analysis of the FHA considers postulated to occur during power operations at less than full rated power or radioactive decay between the time of those postulated to occur at the beginning of core life. For events postulated to core shutdown and the beginning of occur while the facility is shutdown, e.g., a fuel handling accident, radioactive fuel movement.

decay from the time of shutdown may be modeled.

E5 - 13

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

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3.2 The core inventory release fractions, by radionuclide groups, for the gap release Conforms - The LOCA AST calculation and early in-vessel damage phases for OBA LOCAs are listed in Table 1 for models Table 2 in the release fraction BWRs and Table 2 for PWRs. These fractions are applied to the equilibrium and timing file.

core inventory described in Regulatory Position 3.1.

Table 2 PWR Core Inventory Fraction Released Into Containment Gap Early Release In-vessel Group Phase 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 o.oo 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 3.2 For non-LOCA events, the fractions of the core inventory assumed to be in the Conforms - The FHA, Control Rod gap for the various radionuclides are given in Table 3. The release fractions Ejection, and Locked Rotor accidents from Table 3 are used in conjunction with the fission product inventory result in fue.1 damage, so the non-calculated with the maximum core radial peaking factor.

LOCA gap fractions of Table 3 are used. While the SGTR and MSLB Table 3. 11 Non-LOCA Fraction of Fission Product Inventory in Gap accidents conservatively assume a pre-existing 1% leaking fuel source term for Table 3 the RCS, this is not the result of damage caused by the accident, and Group Fraction so the non-LOCA gap fractions of 1-131 0.08 Table 3 are not included for these Kr-85 0.10 events.

ES -14

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

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Other Noble Gases 0.05 Other Halogens 0.05 Alkali Metals 0.12 3.3 Table 4 tabulates the onset and duration of each sequential release phase for Conforms - The LOCA AST calculation OBA LOCAs at PWRs and BWRs. The specified onset is the time following the models Table 4 in the release fraction initiation of the accident (i.e., time= 0). The early in-vessel phase immediately and timing file.

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 3.3 For facilities licensed with leak-before-break methodology, the onset of the gap Conforms - The LOCA AST calculation release phase may be assumed to be 1O minutes. A licensee may propose an models Table 4 in the release fraction alternative time for the onset of the gap release phase, based on facility-specific and timing file.

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 aap release phase onsets in Table 4 should be used.

3.4 Elements listed in Table 5 in each radionuclide group that should be considered Conforms The source term in the in design basis analyses. design basis analysis represents the most dose significant isotopes from the Table 5 elements listed in Table 5 of Radionuclide Groups Regulatory Guide 1.183.

Group Elements Noble Gases Xe, Kr Halogens I, Br ES- 15

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables Tabl~ A: Conformance With Regulatory Guide 1.183 Section C *

.RG

section RG Position FNP Analysis 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 3.5 Of the radioiodine released from the reactor coolant system (RCS) to the Conforms - The chemical composition containment in a postulated accident, 95 percent of the iodine released should of the iodine released from the RCS to be assumed to be cesium iodide (Csl), 4.85 percent elemental iodine, and 0.15 containment in the LOCA event is 95%

percent organic iodide. This includes releases from the gap and the fuel pellets. aerosol, 4.85% elemental, and 0.15%

With the exception of elemental and organic iodine and noble gases, fission organic. All non iodine and non-noble products should be assumed to be in particulate form. The same chemical form gas fission products are assumed to be is assumed in releases from fuel pins in FHAs and from releases from the fuel in particulate form. The chemical pins through the RCS in DBAs other than FHAs or LOCAs. However, the composition of iodine species in the transport of these iodine species following release from the fuel may affect non-LOCA events are based upon the these assumed fractions. The accident-specific appendices to this regulatory guidance in the respective appendices guide provide additional details. of Regulatory Guide 1.183.

3.6 The amount of fuel damage caused by non-LOCA design basis events should Conforms - The amount of fuel damage be analyzed to determine, for the case resulting in the highest radioactivity in the Locked Rotor event is based release, the fraction of the fuel that reaches or exceeds the initiation upon the fraction of the core which temperature of fuel melt and the fraction of fuel elements for which the fuel clad experiences DNB as reported in the is breached. Although the NRG staff has traditionally relied upon the departure Updated Final Safety Analysis Report from nucleate boiling ratio (DNBR) as a fuel damage criterion, licensees may (FSAR). The fraction of the fuel rods propose other methods to the NRG staff, such as those based upon enthalpy assumed to melt in the Control Rod deposition, for estimating fuel damage for the purpose of establishing Ejection event is conservatively based radioactivity releases. 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 Conforms - The AST dose committed effective dose equivalent (CEDE) from inhalation and the deep dose consequences are calculated in TEDE.

equivalent (ODE) from external exposure. The calculation of these two E5 - 16

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables 1;*,,:: ,,, .: '****

, :;."' '. '., ...? , *~;1 :*

-T~.~le A: Cqr:iformgnGe With.H~o!:ll~torv Guide, *1.183 Section c ;* ., i * * * .... * * ,;.;, ****f :. " *.~, ' :

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i i 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 4.1.2 The exposure-to-CEDE factors for inhalation of radioactive material should be Conforms ~ Dose Conversion Factors derived from the data provided in' ICRP Publication 30, "Limits for Intakes of for inhalation in this analysis are taken Radionuclides by Workers" (Ref. 19). Table 2.1 of Federal Guidance Report 11, from Table 2.1 of Federal Guidance "Limiting Values of Radionuclide Intake and Air Concentration and Dose Report 11.

Conversion Factors for Inhalation, Submersion, and Ingestion" (Ref. 20),

provides tables of conversion factors acceptable to the NRG staff. The factors in the column headed "effective" yield doses correspondinq 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 Conforms - Offsite breathing rates used 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 in the analysis are consistent with the accident, the breathing rate should be assumed to be 1.8 x 10-4 cubic meters values specified in Section 4.1.3 of per second. After that and until the end of the accident, the rate should be Regulatory Guide 1.183.

assumed to be 2.3 x 10-4 cubic meters per second.

4.1.4 The DOE should be calculated assuming submergence in semi-infinite cloud Conforms - Dose Conversion Factors assumptions with appropriate credit for attenuation by body tissue. The DOE is tor air submergence are taken from the nominally equivalent to the effective dose equivalent (EDE) from external Table 111.1 of Federal Guidance Report exposure if the whole body is irradiated uniformly. Since this is a reasonable 12.

assumption for submergence exposure situations, EDE may be used in lieu of ODE in determining the contribution of external dose to the TEDE. Table 111.1 of Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil" (Ref. 21 ), provides external EDE conversion factors acceptable to the NRG staff. The factors in the column headed "effective" yield doses corresponding to the EDE.

4.1.5 The TEDE should be determined for the most limiting person at the EAB. The Conforms - The TEDE was determined maximum EAB TEDE for any two-hour period following the start of the for the most limiting person at the EAB.

radioactivity release should be determined and used in determining compliance The maximum two-hour TEDE was with the dose criteria in 10 CFR 50.67. The maximum two-hour TEDE should be determined by calculating the determined by calculating the postulated dose for a series of small time postulated dose for a series of small increments and performinq a "slidinq" sum over the increments for successive time increments and performino a ES -17

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

> . , ., Table A: Corifor:marn~~ 'With Reaulatdrv Gµide 1.183. Section* C *

R<3 ,*.

section , HG Position . . , ;;; *!** >>>:*, * , ,* ** ,. *FN_P J\n~a,Ysi$ ~7 ** **, ~ . -~* ,, ,_ "**{ .~

two-hour periods. The maximum TEDE obtained is submitted. The time 'sliding' sum over increments for increments should appropriately reflect the progression of the accident to successive two-hour periods.

capture the peak dose interval between the start of the event and the end of radioactivitv release (see also Table 6).

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

with the dose criteria in 10 CFR 50.67.

4.1.7 No correction should be made for depletion of the effluent plume by deposition Conforms - No correction is made for on the ground. 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 Conforms - The analyses consider the exposure to control room personnel. The applicable sources will vary from applicable sources of contamination to facility to facility, but typically will include: the control room atmosphere for each

Contamination of the control room atmosphere by the intake or event.

infiltration of the radioactive material contained in the radioactive plume released from the facility, With respect to external and

Contamination of the control room atmosphere by the intake or containment shine sources and their infiltration of airborne radioactive material from areas and structures impact on control room doses, the adjacent to the control room envelope, physical design of the control room

Radiation shine from the external radioactive plume released from the envelope and the surrounding auxiliary facility, building provide more than 18" of

Radiation shine from radioactive material in the reactor containment, concrete shielding between the

Radiation shine from radioactive material in systems and components operators and shine sources in all inside or external to the control room envelope, e.g., radioactive material directions around the control room.

buildup in recirculation filters.

The Control Room Emergency Filtration System filters are located outside of and above the control room envelope. The control room ceiling is approximately 24" thick. Accordingly, shielding from the walls and the filter unit casings prevents an appreciable ES - 18

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

.: Table A: Conformance With Requlatorv_ Guide 1.183 S~ction C -

"RG "."- -

" /

'sectton RG Position -- FNP Analvsis dose to the operators during the accident.

The control room is surrounded by the Auxiliary Building (and so 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 envelope wall is only 1 foot thick inside the Auxiliary Building ES - 19

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

Table A: Conformance Witfi*Reglila:tqfy Guide t.183.Sectiori e

HG. ., ' ' *~* ,

Secti_on . RG

. Position ...-*

-~-

(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 wall are included in the Fuel Handling Accident evaluation of control room doses and were found to be not sionificant.

4.2.2 The radioactive material releases and radiation levels used in the control room Conforms - The SNC AST dose dose analysis should be determined using the same source term, transport, and calculations use the same source term, release assumptions used for determining the EAB and the LPZ TEDE values, transport, and release assumptions for unless these assumptions would result in non-conservative results for the Control Room, EAB, and EPZ dose control room. values.

4.2.3 The models used to transport radioactive material into and through the control Conforms - The models used to room, 15 and the shielding models used to determine radiation dose rates from transport radioactive material into and external sources, should be structured to provide suitably conservative through the control room have been estimates of the exposure to control room personnel. structured to provide suitably conservative estimates of the exposure to control room personnel. Shielding models used in the FHA have been structured to provide suitably conservative estimates of the exposure to CR personnel.

4.2.4 Credit for engineered safety features that mitigate airborne radioactive material Conforms - For the AST DBAs covered within the control room may be assumed. Such features may include control under this License Amendment room isolation or pressurization, or intake or recirculation filtration. Refer to Request, credit is taken for control Section 6.5.1, "ESF Atmospheric Cleanup System," of the SRP (Ref. 3) and room isolation and reconfiguring into Regulatory Guide 1.52, "Design, Testing, and Maintenance Criteria for Post the emergency ventilation mode upon accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration accident initiation by a high radiation or and Adsorption Units of Light-Water-Cooled Nuclear Power Plants" (Ref. 25), for Safety Injection signal, where guidance. The control room design is often optimized for the DBA LOCA and appropriate.

the protection afforded for other accident sequences may not be as E5 -20

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

,,. .,Table.,A:"Conforniance With ReQulatorv Guide -1.183. Section C

HG*.,

6'

.section -*- 'HG. Position ",,,, f=NP: Analvsis 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 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 tor 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 Conforms- No credit is taken tor the equipment or prophylactic drugs. Deviations may be considered on a case-by- use of personal protective equipment case basis. or prophylactic druos.

4.2.6 The dose receptor tor these analyses is the hypothetical maximum exposed Conforms - Control room occupancy individual who is present in the control room for 100% of the time during the first and breathing rates are consistent with 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 this regulatory position.

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.

4.2.7 Control room doses should be calculated using dose conversion factors Conforms - Control room doses are identified in Regulatory Position 4.1 above for use in offsite dose analyses. The calculated using dose conversion ODE from photons may be corrected for the difference between finite cloud factors identified in Position 4.1 above.

geometry in the control room and the semi-infinite cloud assumption used in calculating the dose conversion factors. The following expression may be used Equation 1 from Regulatory Guide to correct the semi-infinite cloud dose, ODE., to a finite cloud dose, DDE!inite , 1.183 is used for finite cloud correction where the control room is modeled as a hemisphere that has a volume, V, in when calculating the ODE immersion cubic feet, equivalent to that of the control room (Ref. 22). doses due to airborne activity inside the control room in the Fuel Handling DDE V 0338 Accident.

'(if DDEtinite=

1173 4.3 The guidance provided in Regulatory Positions 4.1 and 4.2 should be used, as Not Applicable - This full scope AST applicable, in re-assessing the radiological analyses identified in Regulatory implementation LAR is for the Position 1.3.1, such as those in NUREG-0737 (Ref. 2). Design envelope source radiological consequences of major terms provided in NUREG-0737 should be updated tor consistency with the FNP DBAs.

AST. In qeneral, radiation exposures to plant personnel identified in Regulatory ES - 21 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

. Table A: Conformance With Reaulatorv Guide 1.183 Section C RG ..

Section RG Position FN P Analysis 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.

4.4 The radiological criteria for the EAB, the outer boundary of the LPZ, and for the The EAB and LPZ acceptance criteria control room are in 10 CFR 50.67. These criteria are stated for evaluating from Table 6 of RG 1.183 are applied.

reactor accidents of exceedingly low probability of occurrence and low risk of The control room acceptance of 5 rem public exposure to radiation, e.g., a large-break LOCA. The control room TEDE is taken from 10 CFR criterion applies to all accidents. For events with a higher probability of 50.67(b)(2)(iii).

occurrence, postulated EAB and LPZ doses should not exceed the criteria tabulated in Table 6.

Table 617 Accident Dose Criteria Accident or Case EAB and LPZ Analysis Release Duration Dose Criteria LOCA 25 rem TEDE 30 days for containment and ECCS leakaoe PWR Steam Affected SG: time to isolate; Generator Tube Unaffected SG(s): until cold Rupture shutdown is established Fuel Damage or 25 rem TEDE Pre-incident Spike Coincident Iodine 2.5 rem TEDE Spike PWR Main Steam Until cold shutdown is Line Break established Fuel Damage or 25 rem TEDE Pre-incident Spike E5-22 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

' . Table ,A: Cqnformance With Reaulatorv Guide 1:183 Section C --

HG *- -- - -.

Section RG Position FNP Analysis Coincident Iodine 2.5 rem-TEDE Spike PWR Locked Rotor 2.5 rem TEDE Until cold shutdown is Accident established PWR Rod Ejection 6.3 rem TEDE 30 days for containment Accident pathway; until cold shutdown is established for secondary pathwav Fuel Handling 6.3 rem TEDE 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Accident 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 Conforms - The EAB and LPZ reference General Design Criteria 19 (GDC 19) from Appendix A to 10 CFR acceptance criteria from Table 6 of RG Part 50 or specify criteria derived from GDC 19. These criteria are generally 1.183 are applied. The control room specified in terms of whole body dose, or its equivalent to any body organ. For occupant acceptance criteria of 5 rem facilities applying for, or having received, approval for the use of an AST, the TEDE is taken from 10 CFR applicable criteria should be updated for consistency with the TEDE criterion in 50.67(b)(2)(iii).

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 Conforms- The FN P AST dose safety analyses and evaluations required by 10 CFR 50.34; they are considered calculations were prepared and to be a significant input to the evaluations required by 10 CFR 50.92 or 10 CFR accepted by SNC under a 10 CFR 50 50.59. These analyses should be prepared, reviewed, and maintained in Appendix B Quality Assurance accordance with quality assurance programs that comply with Appendix B, program.

"Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50.

5.1.1 These design basis analyses were structured to provide a conservative set of Not Applicable- This License assumptions to test the performance of one or more aspects of the facility Amendment Request is not proposing desion. Manv physical processes and phenomena are represented by deviations to conformance with this E5-23

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables Table A: Conformance With f{equlatorv Guide 1.183 Section .G

<RG . *.. . ..

Section *

RG Position

FNP Analvsis conservative, bounding assumptions rather than being modeled directly. The Regulatory Guide.

staff has selected assumptions and models that provide an appropriate and prudent safety margin against unpredicted events in the 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 Conforms - Only safety-related safety-related, are required to be operable by technical specifications, are Engineered Safety Features are powered by emergency power sources, and are either automatically actuated credited in the analysis with an or, in limited cases, have actuation requirements explicitly addressed in assumed single active failure that emergency operating procedures. The single active component failure that results in the greatest impact on the results in the most limiting radiological consequences should be assumed. radiological consequences. A loss of Assumptions regarding the occurrence and timing of a loss of offsite power offsite power is assumed concurrent should be selected with the objective of maximizing the postulated radiological with the start of each event as that consequences. maximizes the dose impact.

5.1.3 The numeric values that are chosen as inputs to the analyses required by 1O Conforms - Numerical values are CFR 50.67 should be selected with the objective of determining a conservative selected and biased for each postulated dose. In some instances, a particular parameter may be conservative application in a conservative direction in one portion of an analysis but be nonconservative in another portion of the with the objective of maximizing the same analysis. For example, assuming minimum containment system spray dose consequences. Numerical values flow is usually conservative for estimating iodine scrubbing, but in many cases for parameters which are controlled by may be nonconservative when determining sump pH. Sensitivity analyses may Technical Specifications are either be needed to determine the appropriate value to use. As a conservative used as direct inputs in the analysis, or alternative, the limiting value applicable to each portion of the analysis may be more conservative values may be used used in the evaluation of that portion. A single value may not be applicable for a to enhance safety margin.

parameter for the duration of the event, particularly for parameters affected by changes in density. For parameters addressed by technical specifications, the E5-24

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

TableA: ConfbrnianceWith Regu!atorv Guid~ 1.1J3~_ Section G

HG *,_.___- *; , '

.se~tioh ** RG Positioh *, FNP Ahalvsis 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 (NOT), 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 Conforms-The FNP OBA analysis change to the design basis of the facility that is voluntarily initiated by the assumptions and methods are licensee. In order to issue a license amendment authorizing the use of an AST compatible with the AST and the TEOE and the TEOE dose criteria, the NRC staff must make a current finding of criteria.

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 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 TEOE criteria.

5.2 The appendices to this regulatory guide provide accident-specific assumptions Conforms- See Tables B, C, 0, E, F, that are acceptable to the staff for performing analyses that are required by 1O and G of this Enclosure.

CFR 50.67. The OBAs addressed in these attachments were selected from accidents that may involve damage to irradiated fuel. This guide does not address OBAs with radiological consequences based on technical specification reactor or secondary coolant-specific activities only. The inclusion or exclusion of a particular OBA in this guide should not be interpreted as indicating that an analysis of that OBA is required or not required. Licensees should analyze the OBAs that are affected by the specific proposed applications of an AST.

E5-25

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

" ' ' ~

Table A: Conformance With RequlatOrv Guide 1.1.83 Section C

.;fiG

'Section

R~ Position FNP Analysis 5.2 The NRC staff has determined that the analysis assumptions in the appendices Conforms - See Tables B, C, D, E, F, to this guide provide an integrated approach to performing the individual and G of this Enclosure.

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.

5.2 The NRC is committed to using probabilistic risk analysis (PRA) insights in its Conforms- PRA was not used as a regulatory activities and will consider licensee proposals for changes in analysis basis for acceptability of this AST assumptions based upon risk insights. The staff will not approve proposals that License Amendment Request.

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.

5.3 Atmospheric dispersion values (X/Q) for the EAB, the LPZ, and the control room Conforms - The X/Q used for the EAB that were approved by the staff during initial facility licensing or in subsequent and the LPZ were previously approved licensing proceedings may be used in performing the radiological analyses by the NRC in License Amendments identified by this guide. Methodologies that have been used for determining X/Q 165/157 and 166/158.

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).

5.3 References 22 and 28 should be used if the FSAR X/Q values are to be revised Not Applicable - The X/Q values used or if values are to be determined for new release points or receptor distances. are those described in the FNP FSAR.

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 E5-26 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

~FNP An~I sis' worst 2-hour exposure period. The NRG computer code PAVAN (Ref. 29) implements Regulatory Guide 1.145 (Ref. 28) and its use is acceptable to the NRG staff. The methodology of the NRG computer code ARCON96 19 (Ref. 26) is generally acceptable to the NRG 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 anal sis methodolo should be reviewed b the NRG staff.

6.0 The assumptions in Appendix I to this guide are acceptable to the NRG staff for Conforms - FNP is retaining the use of performing radiological assessments associated with equipment qualification. the TIO 14844 source term as the basis The assumptions in Appendix I will supersede Regulatory Positions 2.c(1) and for Environmental Qualification.

2.c(2) and Appendix O 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 NRG 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 TI014844 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 T1014844) on EQ doses pending the outcome of the evaluation of the eneric issue.

Footnote For example, a proposed modification to change the timing of a containment Conforms - No modifications are being 6 isolation valve from 2.5 seconds to 5.0 seconds might be acceptable without proposed as part of this AST License any dose calculations. However, a proposed modification that would delay Amendment Request.

containment spray actuation could involve recalculation of OBA LOCA doses, re-assessment of the containment pressure and temperature transient, recalculation of sum H, re-assessment of the emer enc diesel enerator E5-27

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

' . -~, ,_ ... ~ *,z_ Table k Conformance .With ReQulatoryGuide 1.183 Section C . '"

RG 1

~*section RG Position . ' FNP Analysis '

loading sequence, integrated doses to equipment in the containment, and more.

Footnote In performing screenings and evaluations pursuant to 10 CFR 50.59, it may be Not Applicable - This activity is a 7 necessary to compare dose results expressed in terms of whole body and License Amendment Request made thyroid with new results expressed in terms of TEDE. In these cases, the pursuant to 10 CFR Part 90.

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.

Footnote The uncertainty factor used in determining the core inventory should be that Conforms - A 1.02 uncertainty factor is 8 value provided in Appendix K to 10 CFR Part 50, typically 1.02. used for those events resulting in fuel damaqe.

Footnote Note that for some radionuclides, such as Cs-137, equilibrium will not be Conforms - A conservative core factor 9 reached prior to fuel offload. Thus, the maximum inventory at the end of life is applied to the principal radionuclides should be used. to account for cycle-to-cycle variations.

Footnote The release fractions listed here have been determined to be acceptable for use Conforms - Burnup does not exceed 10 with currently approved LWR fuel with a peak burnup up to 62,000 MWD/MTU. 62,000 MWD/MTU at FNP.

The data in this section may not be applicable to cores containing mixed oxide (MOX) fuel.

Footnote The release fractions listed here have been determined to be acceptable for use Conforms - Burnup does not exceed 11 with currently approved LWR fuel with a peak burnup up to 62,000 MWD/MTU 54 GWD/MTU at FNP.

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 NRG-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 qases.

Footnote In lieu of treating the release in a linear ramp manner, the activity for each Conforms - Both RADTRAD and 12 phase can be modeled as being released instantaneously at the start of that LOCADOSE can model the release release phase, i.e., in step increases. either in a linear ramp manner, or E5-28

Enclosure 5 to N L-16-0388 Regulatory Guide 1.183 Conformance Tables

'Table A: Conformance With Reoulatory Guide 1.183 Section C '*,,' ,-

RG . ' ..~*

section ,, RG Position \ FNP Analysis instantaneous release, as required.

Footnote The prior practice of basing inhalation exposure on only radioiodine and not Conforms - Offsite inhalation doses 13 including radioiodine .in external exposure calculations is not consistent with the are calculated consistent with the definition of TEDE and the characteristics of the revised source term. definition of TEDE.

Footnote With regard to the EAB TEDE, the maximum two-hour value is the basis for Not Applicable - This activity is a 14 screening and evaluation under 10 CFR 50.59. Changes to doses outside of the License Amendment Request made two-hour window are only considered in the context of their impact on the pursuant to 10 CFR Part 90.

maximum two-hour EAB TEDE.

Footnote The iodine protection factor (IPF) methodology of Reference 22 may not be Conforms - The iodine protection 15 adequately conservative for all DBAs and control room arrangements since it factor methodology of Reference 22 is models a steady-state control room condition. Since many analysis parameters not used in this application.

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 methodoloQies.

Footnote This occupancy is modeled in the X/Q values determined in Reference 22 and Conforms - The control room 16 should not be credited twice. The ARCON96 Code (Ref. 26) does not occupancy assumptions are incorporate these occupancy assumptions, making it necessary to apply this incorporated in the dose calculations correction in the dose calculations.

Footnote For PWRs with steam generator alternative repair criteria, different dose criteria Conforms - Refer to ARC line items in 17 may apply to steam Qenerator tube rupture and main steam line break analyses. Tables D and E.

Footnote Note that for some parameters, the technical specification value may be Conforms - Filter efficiencies for the 18 adjusted for analysis purposes by factors provided in other regulatory guidance. Penetration Room Filters (PRF), the For example, ESF filter efficiencies are based on the guidance in Regulatory Control Room (CR) Pressurization Guide 1.52 (Ref. 25) and in Generic Letter 99-02 (Ref. 27) rather than the Intake Filters, and the Control Room surveillance test criteria in the technical specifications. Generally, these Recirculation Filters are developed adjustments address potential changes in the parameter between scheduled from Technical Specification surveillance tests. Surveillance requirements, with margin added for filter inefficiency and bypass leakage around the filter (in accordance with the prior CLB analyses of this type: double the inefficiency allowed by TS 5.5.11 (eTS)

ES -29

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

- a..\ ' *< "- -- . ."'

,_ ~ ;,,,, ,,, :_Table, A: eonforrnance'With 'ReQulat6rv Guide 1_:183 Seclion C

RG . -, . ;

Section RG Pbsiti0n*- ',., ,fNP'Analvsis .. : , : , .. .. , * * .. ;,.

and further reduce efficiency by a 0.5%

bypass amount).

Typically, the efficiency (ef) is calculated as follows:

ef = (1 - 2 * (1 - eTS) - 0.005)

Example: PRF Charcoal Efficiency =

(1-2*(1-0.95)-0.005) = .895 = 89.5%

This methodology assures compliance with Technical Specification 5.5.11 requirements and US NRC RG-1.52.

Footnote The ARCON96 computer code contains processing options that may yield X/Q Conforms - The ARCON96 processing 19 values that are not sufficiently conservative for use in accident consequence options and input parameters were assessments or may be incompatible with release point and ventilation intake based on the release point and configurations at particular sites. The applicability of these options and ventilation intake configurations at associated input parameters should be evaluated on a case-by-case basis. The FNP.

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.

E5-30

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

- - _Table 8:-0onformance With Regulatory Guide_ 1-.183 Appendix A (Loss of Coolant Accident)

,HG ..

Section - HG Position ., , , FNP Analysis * *-

A-1 Acceptable assumptions regarding core inventory and the release of Conforms -See discussions in Table radionuclides from the fuel are provided in Regulatory Position 3 of this guide. A.

A-2 If the sump or suppression pool pH is controlled at values of 7 or greater, the Conforms - The pH of the containment chemical form of radioiodine released to the containment should be assumed to sump is maintained equal to or greater be 95% cesium iodide (Csl), 4.85 percent elemental iodine, and 0.15 percent than 7.0 after the onset of the spray organic iodide. Iodine species, including those from iodine re-evolution, for recirculation mode. Therefore the sump or suppression pool pH values less than 7 will be evaluated on a case-by- radioiodine composition of 95 percent case basis. Evaluations of pH should consider the effect of acids and bases cesium iodide, 4.85 percent elemental created during the LOCA event, e.g., radiolysis products. With the exception of iodine, and 0.15 percent organic iodide elemental and organic iodine and noble gases, fission products should be is used. The containment sump pH assumed to be in particulate form. analysis was previously reviewed by the NRC in FNP License Amendment 166/158.

A-3.1 The radioactivity released from the fuel should be assumed to mix Conforms - The radioactivity released instantaneously and homogeneously throughout the free air volume of the from the fuel is modeled as mixing primary containment in PWRs or the drywell in BWRs as it is released. This instantaneously and homogeneously in distribution should be adjusted if there are internal compartments that have the Containment.

limited ventilation exchange. The swppression 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 Conforms - An aerosol natural within the containment may be credited. Acceptable models for removal of deposition rate of 0.1 hr-1 is assumed iodine and aerosols are described in Chapter 6.5.2, "Containment Spray as a based upon values presented Section Fission Product Cleanup System," of the Standard Review Plan (SAP), VI of NUREG/CR-6189.

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.

ES - 31

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

*'* * :"(~ble .B: '.Conformance With ReguJatorv Guide 11t 1a3* Appendix A ('l~oss of C6otantAccident) . .

,"RG~~ ,~'1c_~:, * "-" ~- 'r ,_., ~ * <

' 'i:

Sebtiorr* *, .RG Position -,_ :f.N P An.alvsis ~* '" ~ -.

A-3.3 Reduction in airborne radioactivity in the containment by containment spray Conforms - Containment Spray is systems that have been designed and are maintained in accordance with credited for elemental and particulate Chapter 6.5.2 of the SRP (Ref. A-1) may be credited. Acceptable models for the iodine removal.

removal of iodine and aerosols are described in Chapter 6.5.2 of the SRP and 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 RAOTRAO (Refs. A-1 to A-3).

A-3.3 The evaluation of the containment sprays should address areas Within the Conforms - Containment Spray covers primary containment that are not covered by the spray drops. The mixing rate less than 90% of the Containment attributed to natural convection between sprayed and unsprayed regions of the volume, so the modeling includes both containment building, provided that adequate flow exists between these regions, the sprayed volume and unsprayed is assumed to be two turnovers of the unsprayed regions per hour, unless other volume. A flow rate of 12,045 cfm is rates are justified. The containment building atmosphere may be considered a used between the sprayed and single, well-mixed volume if the spray covers at least 90% of the volume and if unsprayed volume which correlates to adequate mixing of unsprayed compartments can be shown. two turnovers of the unsprayed region per hour.

A-3.3 The SRP sets forth a maximum decontamination factor (OF) for elemental Conforms - Elemental and aerosol iodine based on the maximum iodine activity in the primary containment removal coefficients are calculated for atmosphere when the sprays actuate, divided by the activity of iodine remaining the sprayed regions of the containment at some time after decontamination. The SRP also states that the particulate using the guidelines of Chapter 6.5.2 of iodine removal rate should be reduced by a factor of 1O when a OF of 50 is the Standard Review Plan. The reached. The reduction in the removal rate is not required if the removal rate is elemental iodine removal coefficients based on the calculated time-dependent airborne aerosol mass. There is no are limited to a maximum value of specified maximum OF for aerosol removal by sprays. The maximum activity to 13.7/hr, and are set to zero when the be used in determining the OF is defined as the iodine activity in the columns elemental iodine decontamination labeled "Total" in Tables 1 and 2 of this guide multiplied by 0.05 for elemental factor (OF) reaches a value of 200. The iodine and by 0.95 for particulate iodine (i.e., aerosol treated as particulate in aerosol removal coefficients are SRP methodology). reduced by a factor of 1O when the aerosol OF reaches 50.

E5-32

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables I* **

.. Table B: Qontormance WithHequlatorv Guide 1.183 Appendi.l;C: A (Loss of Coolant Accident)

'RG '. ' . ..

,,Sectioo RG Position .. FNP..Analvsis A-3.4 Reduction in airborne radioactivity in the containment by in-containment Conforms - No credit is taken for in-recirculation filter systems may be credited if these systems meet the guidance containment recirculation filter systems.

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.

A-3.5 Reduction in airborne radioactivity in the containment by suppression pool Not Applicable - FNP is a PWR.

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 suooression pool liquid pH is not maintained qreater than 7.

A-3.6 Reduction in airborne radioactivity in the containment by retention in ice Conforms - No credit is taken for ice condensers, or other engineering safety features not addressed above, should condensers or other engineering safety be evaluated on an individual case basis. See Section 6.5.4 of the SRP (Ref. A- features to reduce airborne 1). radioactivity in containment.

A-3.7 The primary containment (i.e., drywell for Mark I and II containment designs) Conforms - The containment leak rate 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 is the maximum 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 value allowed by the FNP Technical hours to 50% of the technical specification leak rate. For BWRs, leakage may Specifications. It is reduced to 50% of 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 that value after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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.

A-3.7 For BWRs with Mark Ill containments, the leakage from the drywell into the Not Applicable. FNP is a PWR.

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 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 throuqhout the drvwell and the primary containment.

E5-33

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables Table*B: Conformance With Regulatory Guide 1.183 Appendix A '(Loss~ orCoolant Accident)

'RG-'*

section RG Position *

A-3.8 If the primary containment is routinely purged during power operations, releases Conforms - Based upon the isolation of via the purge system prior to containment isolation should be analyzed and the the mini-purge flow within 30 seconds, resulting doses summed with the postulated doses from other release paths. the mini-purge system will be isolated The purge release evaluation should assume that 100% of the radionuclide before the onset of the gap release as inventory in the reactor coolant system liquid is released to the containment at defined in Table 4 of this Regulatory the initiation of the LOCA. This inventory should be based on the technical Guide. Therefore, only those nuclides specification reactor coolant system equilibrium activity. Iodine spikes need not in the RCS source term are available be considered. If the purge system is not isolated before the onset of the gap for release.

release phase, the release fractions associated with the gap release and early in-vessel phases should be considered as applicable.

A-4 For facilities with dual containment systems, the acceptable assumptions Not Applicable. FNP does not have a related to the transport, reduction, and release of radioactive material in and dual containment.

from the secondary containment or enclosure buildings are as follows.

A-5.1 With the exception of noble gases, all the fission products released from the fuel Conforms - With the exception of noble to the containment (as defined in Tables 1 and 2 of this guide) should be gases, all the fission products released assumed to instantaneously and homogeneously mix in the primary from the fuel to the containment containment sump water (in PWRs) or suppression pool (in BWRs) at the time instantaneously and homogeneously of release from the core. In lieu of this deterministic approach, suitably mix in the primary sump water.

conservative mechanistic models for the transport of airborne 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 The FNP Technical Specifications do from all components in the ESF recirculation systems above which the technical not provide a specific limit for specifications, or licensee commitments to item 111.D.1.1 of NUREG-0737 (Ref. operational leakage from ECCS A-8), would require declaring such systems inoperable. The leakage should be systems. However, administrative assumed to start at the earliest time the recirculation flow occurs in these limits ensure that operational leakage systems and end at the latest time the releases from these systems are is adequately controlled. In the terminated. Consideration should also be given to design leakage through analysis, an assumed leakage from valves isolating ESF recirculation systems from tanks vented to atmosphere, ECCS systems is taken as 20,000 e.Q., emerqency core coolinq system (ECCS) pump miniflow return to the cc/hr for leakage of sump water outside E5-34

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

.., ' * *.* '" ,,,. Table 6: .Cohforrnaric~ With .H~oulatory Guide t.183 Aopendbc A (tos~ ot Coolant Aqcident). *

-Rta"*:... ' . * .. ._ * *-* . . .... .. ** .. , '~

}3~dtfon " "~d #~sition : . '.; .* :. . . , - . . .- , . . .:. .* . . ** ** *

FN P Analysis . .. ,.

refueling water storage tank. of containment into the Auxiliary Building, which is multiplied by two, consistent with this Regulatory Position. In addition, two times the assumed leak rate of 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 Conforms - With the exception of should be assumed to be retained in the liquid phase. 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 Conforms - It is assumed for the case the liquid that becomes airborne should be assumed equal to the fraction of the when the temperature of the ECCS leakage that flashes to vapor. This flash fraction, FF, should be determined leakage exceeds 212° F that the using a constant enthalpy, h, process, based on the maximum time-dependent fraction of total iodine in the liquid that temperature of the sump water circulating outside the containment: becomes airborne is equal to the fraction of the leakage that flashes to FF= t, h -h r2 vapor. This flash fraction, FF, is determined assuming a constant hfg enthalpy, h, process, and is based on Where: h11 is the enthalpy of liquid at system design temperature and pressure; the maximum time-dependent sump h12 is the enthalpy of liquid at saturation conditions (14.7 psia, 212°F); and h19 is water temperature.

the heat of vaporization at 212°F.

E5-35 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

- 'Table B: Conforma,nce)fl/.ith R~qulatorv Guide 1. f83 Appendix A (Loss. of Cgqlant Accident)

RG Section RG. Position * ** .. , . .. . FNP 'Analysis

A-5.5 If the temperature of the leakage is less than 212°F or the calculated flash Conforms - Since the calculated fraction is less than 10%, the amount of iodine that becomes airborne should be flashing fraction is less than 10%, and assumed to be 10% of the total iodine activity in the leaked fluid, unless a without a basis for justifying a smaller smaller amount can be justified based on the actual sump pH history and area value, 10% of the iodine in the ECCS ventilation rates. leakage is assumed to be released.

A-5.6 The radioiodine that is postulated to be available for release to the environment Conforms - The radioiodine that is is assumed to be 97% elemental and 3% organic. Reduction in release activity postulated to be available for release to by dilution or holdup within buildings, or by ESF ventilation filtration systems, the environment is modeled as 97%

may be credited where applicable. Filter systems used in these applications elemental and 3% organic.

should be evaluated against the guidance of Regulatory Guide 1.52 (Ref. A-5) and Generic Letter 99-02 (Ref. A-6).

A-6 For BWRs, the main steam isolation valves (MSIVs) have design leakage that Not Applicable. FNP is a PWR.

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 assumptions are acceptable for evaluatinq the consequences of MSIV leakaqe.

A-7 The radiological consequences from post-LOCA primary containment purging Conforms - FNP uses hydrogen as a combustible gas or pressure control measure should be analyzed. If the recombiners for post-accident installed containment purging capabilities are maintained for purposes of severe hydrogen control. As such, the accident management and are not credited in any design basis analysis, containment mini-purge system is radiological consequences need not be evaluated. If the primary containment assumed to not be available for purging is required within 30 days of the LOCA, the results of this analysis combustible gas management and this should be combined with consequences postulated for other fission product pathway is assumed to remain closed release paths to determine the total calculated radiological consequences from following a containment isolation the LOCA. Reduction in the amount of radioactive material released via ESF signal.

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).

E5-36

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

~ *. ~ " t

HG ..

,*se~fl~ri , RG Position * * ***

__ .. f=:NP:Anal sis;_

Footnote This document describes statistical formulations with differing levels of Conforms - The removal rate constants A-1 uncertainty. The removal rate constants selected for use in design basis selected for use in the LOCA calculations should be those that will maximize the dose consequences. For calculation are those that will maximize BWRs, the simplified model should be used only if the release from the core is the dose consequences.

not directed through the suppression pool. Iodine removal in the suppression ool affects the iodine s ecies assumed b the model to be resent initial! .

ES-37

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables Table~C: Conformance With Reaulatory Guide t.183 Appendix B (Pµel Handling Accident)

.'RG*

' Section.* :. RG Position . .. . .. * * **

B-1 Acceptable assumptions regarding core inventory and the release of Conforms - See discussions in Table radionuclides from the fuel are provided in Regulatory Position 3 of this guide. A.

B-1.1 The number of fuel rods damaged during the accident should be based on a Conforms - The FHA is a single fuel conservative analysis that considers the most limiting case. This analysis assembly dropped from within either should consider parameters such as the weight of the dropped heavy load or the Containment, the Fuel Handling the weight of a dropped fuel assembly (plus any attached handling grapples), Building, or the Auxiliary Building the height of the drop, and the compression, torsion, and shear stresses on without interaction with any other fuel the irradiated fuel rods. Damage to adjacent fuel assemblies, if applicable assemblies. The number of fuel rods (e.g., events over the reactor vessel), should be considered. damaged is equal to one fuel assembly.

B-1.2 The fission product release from the breached fuel is based on Regulatory Conforms - The fission product release Position 3.2 of this guide and the estimate of the number of fuel rods is equal to the gap release, with breached. All the gap activity in the damaged rods is assumed to be isotopic fractions as given in Table 3 of instantaneously released. Radionuclides that should be considered include RG 1.183 (8% for 1-131, 10% for Kr-85, xenons, kryptons, halogens, cesiums, and rubidiums. 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 Conforms - The chemical forms of should be assumed to be 95% cesium iodide (Csl), 4.85 percent elemental radioiodine released from the fuel to iodine, and 0.15 percent organic iodide. The Csl released from the fuel is the spent fuel pool is assumed to be assumed to completely dissociate in the pool water. Because of the low pH of 95% cesium iodide (Csl), 4.85 percent the pool water, the iodine re-evolves as elemental iodine. This is assumed to elemental iodine, and 0.15 percent occur instantaneously. The NRC staff will consider, on a case-by-case basis, organic iodide. The Csl released from justifiable mechanistic treatment of the iodine release from the pool. the fuel completely dissociates in the pool water and re-evolves as elemental iodine. The dissociation and re-evolution occurs instantaneously.

ES - 38

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

-; . :.. :.. :.Table e.: Conformance Witb Begulatorv'Guiae .J .183. A'pPeridb(Ef(fl:J~I Handlina Accident) '* *. i* * *.

,,* .;;.. :,. *~-; .,

<RG .: ** '* ** *' .: .. . .. . ' .... ,*

~s06tio~. AG.Position * * * *. * * .* . ***' *.. *. * . . .. : . . .. :FNPAnalysis * * *** * . . * .. ~;., .....,

B-2 If the depth of water above the damaged fuel is 23 feet or greater, the _ Conforms -Water level is greater than decontamination factors for the elemental and organic species are 500 and 1, 23 feet for each case. Therefore, the respectively, giving an overall effective decontamination factor of 200 (i.e., pool water is assumed to have a 99.5% of the total iodine released from the damaged rods is retained by the decontamination factor of 500 for water). This difference in decontamination factors for elemental (99.85%) and iodine isotopes in an *organic form. This organic iodine (0.15%} species results in the iodine above the water being assumption leads to an overall composed of 57% elemental and 43% organic species. If the depth of water is effective decontamination factor of 200 not 23 feet, the decontamination factor will have to be determined on a case- for the iodine isotopes released from by-case method (Ref. B-1). the gap.

B-3 The retention of noble gases in the water in the fuel pool or reactor cavity is Conforms - Noble gases are not negligible (i.e., decontamination factor of 1). Particulate radionuclides are scrubbed by the pool water assumed to be retained by the water in the fuel pool or reactor cavity (i.e., {decontamination factor of 1).

infinite decontamination factor). 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 Conforms - For releases in assumed to be released to the environment over a 2-hour time period. 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 Conforms - A rectuction in the amount by engineered safety feature (ESF) filter systems may be taken into account of radioactive material released from provided these systems meet the guidance of Regulatory Guide 1.52 and the spent fuel pool area of the Auxiliary Generic Letter 99-02 (Refs. B-2, B-3). Delays in radiation detection, actuation Building is credited by use of the of the ESF filtration system, or diversion of ventilation flow to the ESF filtration Penetration Room Filter (PRF) system.

system 1 should be determined and accounted for in the radioactivity release This system meets the requirements of analyses. Regulatory 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.

E5-39 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables Table C: Conformance With Reoulatorv Guide* 1.183 Appendix B (Fliel Hand lino Accident) .

RG. .,,. ';'-J _

Section BG Position . *

FNP Analvsis ..

B-4.3 The radioactivity release from the fuel pool should be assumed to be drawn Conforms - There is no credit taken for into the ESF filtration system without mixing or dilution in the fuel building. If mixing or dilution in the spent fuel pool mixing can be demonstrated, credit for mixing and dilution may be considered area of the Auxiliary Building.

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.

B-5.1 If the containment is isolated;:: during fuel handling operations, no radiological Not Applicable - Containment is not consequences need to be analyzed. assumed to be isolated during fuel handling operations.

B-5.2 If the containment is open during fuel handling operations, but designed to Not Applicable -The containment automatically isolate in the event of a fuel handling accident, the release equipment hatch and personnel airlock duration should be based on delays in radiation detection and completion of are modeled as being open during an containment isolation. If it can be shown that containment isolation occurs FHA and no credit it taken in the before radioactivity is released to the environment, 1 no radiological analysis for closing them.

consequences need to be analyzed.

B-5.3 If the containment is open during fuel handling operations (e.g., personnel air Conforms - The FHA radiological lock or equipment hatch is open), 3 the radioactive material that escapes from release is over a two-hour period.

the reactor cavity pool to the containment is released to the environment over a 2-hour time period.

B-5.4 A reduction in the amount of radioactive material released from the Not Applicable - No credit is taken for containment by ESF filter systems may be taken into account provided that ESF filter systems to mitigate these systems meet the guidance of Regulatory Guide 1.52 and Generic radioactive material release from the Letter 99-02 (Refs. B-2 and B-3). Delays in radiation detection, actuation of Containment.

the ESF filtration system, or diversion of ventilation flow to the ESF filtration system should be determined and accounted for in the radioactivity release analvses. 1 B-5.5 Credit for dilution or mixing of the activity released from the reactor cavity by The free volume of the FNP natural or forced convection inside the containment may be considered on a containment is 2.0E6 cubic feet. The case-by-case basis. Such credit is oenerallv limited to 50% of the containment free volume used in the FHA dose E5-40

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables I*:: ;** ' --

.c *,fable C:.QooformanCe With *Regulato,.Y J3uid_e 1.183*ApperidixB:{E"uel.Han'dlihg Ac.cident} * .

-;fitl. -: . .- " L'

~~

~

,r, - . *- _. '

Sectian* * - HG Position**: '""

,,- ' -~ -: -- -:'~ '11" _.

*** --1--:r *-r ' ; . -:. ,, ', _,

FNP Analysis _, " ,>>O free volume. This evaluation should consider the magnitude of the calculation was 1.0E6 cubic feet.

containment volume and exhaust rate, the potential for bypass 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.

Footnote These analyses should consider the time for the radioactivity concentration to Conforms - The FHA calculation B-1 reach levels corresponding to the monitor setpoint, instrument line sampling demonstrates that a sufficient time, detector response time, diversion damper alignment time, and filter concentration of radioactivity occurs at system actuation, as applicable. the Control Room Ventilation System sensor to result in control room isolation within the assumed 60 second delay time.

Footnote Containment isolation does not imply containment integrity as defined by Not Applicable - Containment is not B-2 technical specifications for non-shutdown modes. The term isolation is used assumed to be isolated during fuel here collectively to encompass both containment integrity and containment handling operations.

closure, typically in place during shutdown periods. To be credited in the analysis, the appropriate form of isolation should be addressed in technical specifications.

Footnote The staff will generally require that technical specifications allowing such Conforms - FNP TS 3.9.3 establishes B-3 operations include administrative controls to close the airlock, hatch, or open the requirements for containment penetrations within 30 minutes. Such administrative controls will generally penetrations during refueling require that a dedicated individual be present, with necessary equipment operations. The FHA dose calculation available, to restore containment closure should a fuel handling accident takes no credit for manual isolation of occur. Radiolooical analyses should oenerallv not credit this manual isolation. containment after the event.

ES - 41

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

.Tabl~ D: .ConformanGe Witl:t Reoulatorv Guide. 1*.183 Appendix E (Main. Steam Line Break Accident) .

. RG ~, \ "

Section *RG Position . * . * , , . .* . . ... * * ** ~FNP* Analysi~. '-~ ~

E-1 Assumptions acceptable to the NRG staff regarding core inventory and the - Conforms - See discussions in Table release of radionuclides from the fuel are provided in Regulatory Position 3 of A.

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.

E-2 If no or minimal fuel damage is postulated for the limiting event, the activity Consistent with the FNP current released should be the maximum coolant activity allowed by the technical licensing basis a leaking fuel term is specifications. Two cases of iodine spiking should be assumed. conservatively included with the two cases of iodine spikinQ.

E-2.1 A reactor transient has occurred prior to the postulated main steam line break Conforms - The Main Steam Line (MSLB) and has raised the primary coolant iodine concentration to the Break Accident dose calculation maximum value (typically 60 µCi/gm DE 1-131) permitted by the technical includes a case for a preaccident specifications (i.e., a preaccident iodine spike case). 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 Conforms - The Main Steam Line spike in the primary system. The increase in primary coolant iodine Break Accident dose calculation concentration is estimated using a spiking model that assumes that the iodine includes a case for a concurrent iodine release rate from the fuel rods to the primary coolant (expressed in curies per spike causing the iodine release rate unit time) increases to a value 500 times greater than the release rate from the fuel rods to the RCS to corresponding to the iodine concentration at the equilibrium value (typically increase to a value 500 times greater 1.0 µCi/gm DE 1-131) specified in technical specifications (i.e., concurrent than the release rate that yields the iodine spike case). A concurrent iodine spike need not be considered if fuel equilibrium iodine concentration 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 . specified in the technical specifications.

Shorter spike durations may be considered on a case-by-case basis if it can The iodine spike duration is 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

be shown that the activity released by the 8-hour spike exceeds that available For conservatism, the concurrent for release from the fuel gap of all fuel pins. iodine spike is assumed even with initial RCS activity from 1% leaking fuel.

E5-42

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables I**"* A

- Table D: Conformance With Requlatorv Guide 1. ~ 83 Appendix E (Main Steam . Line Break

. .. Accident)

RG " .

.*Section RGPosition :* . '* ' '

FNP Analv~is E-3 The activity released from the fuel should be assumed to be released Conforms - The initial activity from the instantaneously and homogeneously through the primary coolant. leaking 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 Conforms - The iodine releases from be 95% cesium iodide (Csl), 4.85 percent elemental iodine, and 0.15 percent the steam generators to the organic iodide. Iodine releases from the steam generators to the environment environment are 97% elemental and should be assumed to be 97% elemental and 3% organic. These fractions 3% organic for the pre-accident case apply to iodine released as a result of fuel damage and to iodine released and the concurrent iodine spike case, durina normal operations, includina iodine soikina. includinq failed (leakinq) fuel.

E-5.1 For facilities that have not implemented alternative repair criteria (see Ref. E- Conforms - FNP is licensed to ARC.

1, DG-1074), the primary-to-secondary leak rate in the steam generators The assumed primary-to-secondary should be assumed to be the leak rate limiting condition for operation leak rate in the two intact steam specified in the technical specifications. For facilities with traditional generator generators are 0.65 gpm (936 gallons specifications (both per generator and total of all generators), the leakage per day). This is conservative relative should be apportioned between affected and unaffected steam generators in to FNP TS 3.4.13 which allows 150 such a manner that the calculated dose is maximized. gallons per day per Steam Generator.

E-5.2 The density used in converting volumetric leak rates (e.g., gpm) to mass leak Conforms - The assumed density is rates (e.g., lbm/hr) should be consistent with the basis of the parameter being 62.4 lbm/ft3 .

converted. The ARC leak rate correlations are generally based on the collection of cooled 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/ft3).

E-5.3 The primary-to-secondary leakage should be assumed to continue until the Conforms - For the faulted steam primary system pressure is less than the secondary system pressure, or until generator, primary-to-secondary the temperature of the leakage is less than 100°C (212°F). The release of leakage continues for the duration of radioactivity from unaffected steam generators should be assumed to the event. The release from the continue until shutdown cooling is in operation and releases from the steam unaffected steam generators continues generators have been terminated. until the Reactor Coolant System is reduced to cold shutdown conditions in 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

ES -43 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

. *Tabl'e*D: Conformance_ With Re!:lulatorv Guide *l.*ts3*Appendix E (Main Steam Line Break Accident)

HG. °'

Section . HG-Po'sition, . ;. FN.P Analysis E-5.4 All noble gas radionuclides released from the primary system are assumed to Conforms - All noble gases are be released to the environment without reduction or mitigation. 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 Conforms - See below.

particulate releases from the steam generators. This model is shown in Figure E-1 and summarized below:

E-5.5.1 A portion of the primary-to-secondary leakage will flash to vapor, based on the Conforms - The leakage of the faulted thermodynamic conditions in the reactor and secondary coolant. steam generator is modeled as a direct

During periods of steam generator dryout, all of the primary-to- vapor flow from the RCS to the secondary leakage is assumed to flash to vapor and be released to environment without partitioning. For the environment with no mitigation. the intact steam generators, primary-to-

With regard to the unaffected steam generators used for plant secondary leakage mixes with the cooldown, the primary-to-secondary leakage can be assumed to mix secondary water without flashing for with the secondary water without flashing during periods of total tube the duration of the event.

submergence.

E-5.5.2 The leakage that immediately flashes to vapor will rise through the bulk water Conforms - For conservatism, no credit of the steam generator and enter the steam space. Credit may be taken for is taken for scrubbing.

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.

E-5.5.3 The leakage that does not immediately flash is assumed to mix with the bulk Conforms - The leakage that does not water. 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 Conforms - For flows out of the intact is the function of the steaming rate and the partition coefficient. A partition SGs, radioactivity to the environment is coefficient for iodine of 100 may be assumed. The retention of particulate a function of the steaming rate, and the radionuclides in the steam generators is limited by the moisture carryover iodine partition factor is assumed to be from the steam generators. 100. Moisture carryover is modeled at 0.1%.

E-5.6 Operating experience and analyses have shown that for some steam Conforms - The steam generator with qenerator desiqns, tube uncoverv may occur for a short period followinq any the faulted main steamline in the MSLB E5-44 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

' . *,: ., Tab.le~ D: Conformance. With* Re endix IE Main Steam' Line Break Accident .-.

AG FNP Aria,I sis '. ' .*

reactor trip (Ref. E-3). The potential impact of tube uncovery on the transport accident is assumed to blow model parameters (e.g., flash fraction, scrubbing credit) needs to be completely dry, causing a direct considered. The impact of emergency operating procedure restoration release of radioactivity from that source strate ies on steam enerator water levels should be evaluated. to the environment.

Footnote Facilities licensed with, or applying for, alternative repair criteria (ARC) should Conforms- FNP is licensed to ARC.

E-1 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 anal ses.

Footnote The activity assumed in the analysis should be based on the activity Consistent with the FNP current E-2 associated with the projected fuel damage or the maximum technical licensing basis a failed (leaking) fuel specification values, whichever maximizes the radiological consequences. In term is conservatively included with the determining dose equivalent 1-131 (DE 1-131), only the radioiodine associated two cases of iodine spiking.

with normal operations or iodine spikes should be included. Activity from ro"ected fuel dama e should not be included.

ES-45

Enclosure 5 to N L-16-0388 Regulatory Guide 1.183 Conformance Tables

.' Table E: Confor:manc*e Witb Regulatory Gl!ide 1.183 Appendix F (Steam Generator TubeHuptureAccident)

BG. . .

,},,' '

.~Section RG Position FN P Analysis F-1 Assumptions acceptable to the NRC staff regarding core inventory and the Conforms - See discussions in Table release of radionuclides from the fuel are in Regulatory Position 3 of this A.

guide. The release from the breached fuel is based on Regulatory Position 3.2 of this c:::iuide and the estimate of the number of fuel rods breached.

F-2 If no or minimal;- fuel damage is postulated for the limiting event, the activity Consistent with the FNP current released should be the maximum coolant activity allowed by technical licensing basis a failed (leaking) fuel specification. Two cases of iodine spiking should be assumed. term is conservatively included with the two cases of iodine spikinc:::i.

F-2.1 A reactor transient has occurred prior to the postulated steam generator tube Conforms - Case 1 is a pre-accident rupture (SGTR) and has raised the primary coolant iodine concentration to the spike using the maximum Dose maximum value (typically 60 µCi/gm DE 1-131) permitted by the technical Equivalent Iodine permitted by the FNP specifications (i.e., a preaccident iodine spike case). Technical Specifications.

F-2.2 The primary system transient associated with the SGTR causes an iodine Conforms - The concurrent iodine spike in the primary system. The increase in primary coolant iodine spike case assumes the RCS transient concentration is estimated using a spiking model that assumes that the iodine associated with the accident creates an release rate from the fuel rods to the primary coolant (expressed in curies per iodine spike, causing the iodine release unit time) increases to a value 335 times greater than the release rate rate from the fuel rods to the RCS to corresponding to the iodine concentration at the equilibrium value (typically increase to a value 335 times greater 1.0 µCi/gm DE 1-131) specified in technical specifications (i.e., concurrent than the release rate that yields the iodine spike case). A concurrent iodine spike need not be considered if fuel equilibrium iodine concentration 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 . specified in the technical specifications.

Shorter spike durations may be considered on a case-by-case basis if it can Initial RCS activity conservatively be shown that the activity released by the 8-hour spike exceeds that available includes a 1% leaking fuel source term, for release from the fuel gap of all fuel pins. consistent with the FNP current licensing basis for this event. An 8-hour release duration is modeled.

F-3 The activity released from the fuel, if any, should be assumed to be released Conforms - Mixing in the primary instantaneously and homogeneously through the primary coolant. coolant is assumed to be instantaneously and homogeneouslv.

F-4 Iodine releases from the steam generators to the environment should be Conforms - The iodine released to the assumed to be 97% elemental and 3% organic. environment is assumed to be 97%

elemental and 3% organic.

ES - 46

Enclosure 5 to NL- J 6-0388 Regulatory Guide 1.183 Conformance Tables

Table E: Conformance With Reaulatorv. Guide

. 1.183-- Aooelldix

- F.. (Steam . Generator:. Tube . Rupture

- :Accident)

RG ' '*'

'"t:r ""*

section RG Position ' .,, , fNP Analysis .

F-5.1 The primary-to-secondary leak rate in the steam generators should be Conforms - The assumed primary-to-assumed to be the leak rate limiting condition for operation specified in the secondary leak rate in the two intact technical specifications. The leakage should be apportioned between affected steam generators are 0.65 gpm (936 and unaffected steam generators in such a manner that the calculated dose is gallons per day). This is conservative maximized. 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 Conforms - The assumed density is rates (e.g., lbm/hr) should be consistent with the basis of surveillance tests 62.4 lbm/ft3 .

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 Qm/cc (62.4 lbm/ft3 ).

F-5.3 The primary-to-secondary leakage should be assumed to continue until the Conforms - It is assumed that cold primary system pressure is less than the secondary system pressure, or until shutdown is established at 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the temperature of the leakage is less than 100°c (212°F). The release of terminating the accident.

radioactivity from the unaffected steam generators should be assumed to continue until shutdown cooling is in operation and releases from the steam Qenerators have been terminated.

F-5.4 The release of fission products from the secondary system should be Conforms - The SGTR assumes a evaluated with the assumption of a coincident loss of offsite power. 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 Conforms - Noble gases are modeled be released to the environment without reduction or mitigation. as going directly to the environment without reduction or mitiaation.

F-5.6 The transport model described in Regulatory Positions 5.5 and 5.6 of Conforms - The transport model Appendix E should be utilized for iodine and particulates. described in Position 5.5 and 5.6 of Appendix E is applied to releases from the steam Qenerators.

E5-47

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables Table E: Conformance - With. Requlatorv Guide. 1.183 Appendix. .F (Steam GeneratorTube

. . . Rupture

- Accident)

RG <<\"*

Section RG Position FNP Analysis ~

Footnote Facilities licensed with, or applying for, alternative repair criteria (ARC) should Conforms - FNP is licensed to ARC.

F-1 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 performinQ radiological analyses.

Footnote The activity assumed in the analysis should be based on the activity Consistent with the FNP current F-2 associated with the projected fuel damage or the maximum technical licensing basis, the initial RCS activity specification values, whichever maximizes the radiological consequences. In conservatively considers 1% leaking determining dose equivalent 1-131 (DE 1-131), only the radioiodine associated fuel with the two cases of iodine with normal operations or iodine spikes should be included. Activity from spiking.

projected fuel damaQe should not be included.

ES - 48

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

-~~

- * 'rablffF: Conformance With_*Regulatory Guide 1.*rn3 Appendix G (Locked Rotpr Accident) *' '

J~G.: . . '

. ,' ~-

i

~

Se,~tJ6n '*

RG'Ppsi_tion. .. FN P Analysis G-1 Assumptions acceptable to the NRC staff regarding core inventory and the Conforms - See discussions in Table release of radionuclides from the fuel are in Regulatory Position 3 of this A.

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.

G-2 If no fuel damage is postulated for the limiting event, a radiological analysis is Conforms - The transient causes fuel not required as the consequences of this event are bounded by the damage and so a radiological analysis consequences projected for the main steam line break outside containment. is provided.

G-3 The activity released from the fuel should be assumed to be released Conforms - The gap activity in the instantaneously and homogeneously through the primary coolant. 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 Conforms - The iodine releases from be 95% cesium iodide (Csl), 4.85 percent elemental iodine, and 0.15 percent the steam generators to the organic iodide. Iodine releases from the steam generators to the environment environment are 97% elemental and should be assumed to be 97% elemental and 3% organic. These fractions 3% organic for the pre-accident case apply to iodine released as a result of fuel damage and to iodine released and the concurrent iodine spike case, durinq normal operations, includinq iodine spikinq. includinq damaqed fuel.

G-5.1 The primary-to-secondary leak rate in the steam generators should be Conforms - Leakage is 1 gpm, which is assumed to be the leak-rate-limiting condition for operation specified in the bounding over the Technical technical specifications. The leakage should be apportioned between the Specification limit of 150 gallons per steam qenerators in such a manner that the calculated dose is maximized. day per steam generator.

G-5.2 The density used in converting volumetric leak rates (e.g., gpm) to mass leak Conforms - The assumed density is rates (e.g., lbm/hr) should be consistent with the basis of surveillance tests 62.4 lbm/ft3 .

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/ft3 ).

G-5.3 The primary-to-secondary leakage should be assumed to continue until the Conforms - The accident terminates primary system pressure is less than the secondary system pressure, or until after 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and cold shutdown -

the temperature of the leakage is less than 100°C (212°F). The release of conditions have been achieved.

E5-49

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables I':: ,, *; ..

.. '* . *_*Table F: Conform~nce.With:Begulatorv Guide* 1.l83 Abp~ndlx.G (Lockeq F{oto.r Accident) *." . ;

.. .. ~;,

.. .. << '~ ' ,,

HG~ .. *'*

' ~' ,

l~ ' ' "

.Section*: ._. HG Positioh : *. * ,' ' ', .. . *' . ' ' .

.. . . . ' FNP AnaJvsis I \

radioactivity should be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated.

G-5.4 The release of fission products from the secondary system should be Conforms - The Locked Rotor Accident evaluated with the assumption of a coincident loss of offsite power. 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 Conforms - Noble gases are assumed be released to the environment without reduction or mitigation. 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 Conforms - The transport model should be utilized for iodine and particulates. described in Position 5.5 and 5.6 of Appendix E is applied to releases from the steam generators.

Footnote Facilities licensed with, or applying for, alternative repair criteria (ARC) should Conforms - FNP is licensed to ARC.

G-1 use this section in conjunction with the guidance that is being developed in Draft Regulatory Guide DG-1074, "Steam Generator Tube Integrity (USN RC, December 1998), for acceptable assumptions and methodologies for performing radiological analyses.

E5-50 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables Tablef G: Conformance With Regulatory Guide 1.183 Appendix H{Rod Ejection Accident)

RG Position , FNP Analysis , ,, ,

H-1 Assumptions acceptable to the NRC staff regarding core inventory are in Conforms - See discussions in Table Regulatory Position 3 of this guide. For the rod ejection accident, the release A. The fission product release is based from the breached fuel is based on the estimate of the number of fuel rods upon Appendix H the amount of breached and the assumption that 10% of the core inventory of the noble damaged fuel and the assumption that gases and iodines is in the fuel gap. The release attributed to fuel melting is 10% of the core inventory of noble based on the fraction of the fuel that reaches or exceeds the initiation gases and iodine isotopes are in the temperature for fuel melting and the assumption that 100% of the noble gases fuel rod gap.

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 For releases from containment involve gases and 50% of the iodines in that fraction are released to the reactor fuel melting, 100% of the noble gases coolant. and 50 % of the iodine isotopes contained in the portion of the fuel which melts is available for release from containment 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 Applicable - Failed fuel is not required as the consequences of this event are bounded by the postulated for this event.

consequences projected for the loss-of-coolant accident (LOCA), main steam line break, and steam qenerator tube rupture.

H-3 Two release cases are to be considered. In the first, 100% of the activity Conforms - Two release pathways are released from the fuel should be assumed to be released instantaneously and considered. In the release from homogeneously through the containment atmosphere. In the second, 100% of containment, 100% of the activity from the activity released from the fuel should be assumed to be completely fuel melting and fuel cladding damage dissolved in the primary coolant and available for release to the secondary instantaneously reaches the system. 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 ES - 51

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

rable G: CoQformance With ReQulatorv Guide 1.183 APP.er:idix H (Rod Ejection Accident) ' ,, ~

'RG "'

section RG Position~.. ~ '

FNP Analysis ._

available for release to the secondary system and eventually to the environment.

H-4 The chemical form of radioiodine released to the containment atmosphere Conforms - The chemical form of should be assumed to be 95% cesium iodide (Csl), 4.85% elemental iodine, radioiodine released to the and 0.15% organic iodide. If containment sprays do not actuate or are containment atmosphere is assumed to terminated prior to accumulating sump water, or if the containment sump pH be 95% cesium iodide, 4.85%

is not controlled at values of 7 or greater, the iodine species should be elemental iodine, and 0.15% organic evaluated on an individual case basis. Evaluations of pH should consider the iodide. Since containment sprays will effect of acids created during the rod ejection accident event, e.g., pyrolysis not necessarily be activated in this and radiolysis products. With the exception of elemental and organic iodine event, no credit is taken for pH being and noble gases, fission products should be assumed to be in particulate controlled at values of 7 or greater.

form.

H-5 Iodine releases from the steam generators to the environment should be The containment distribution was used assumed to be 97% elemental and 3% organic. for the secondary system pathway in the CREA model. This distribution, although different from RG 1.183, Appendix H, Section 5, is acceptable because the removal mechanism for all chemical forms of iodine is the same for this pathway.

H-6.1 A reduction in the amount of radioactive material available for leakage from Conforms - Radioactive material the containment that is due to natural deposition, containment sprays, removal from the containment recirculating filter systems, dual containments, or other engineered safety atmosphere by sprays and other features may be taken into account. Refer to Appendix A to this guide for engineered safety features is not guidance on acceptable methods and assumptions for evaluating these credited. Natural deposition of mechanisms. elemental iodine is credited.

H-6.2 The containment should be assumed to leak at the leak rate incorporated in Conforms - The containment is 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 , assumed to leak to the environment at and at 50% of this leak rate for the remaining duration of the accident. Peak the technical specification limit of accident pressure is the maximum pressure defined in the technical 0.15%/day for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the specifications for containment leak testinQ. LeakaQe from subatmospheric accident and half this rate thereafter.

ES- 52

Enclosure 5 to NL-16-0388 Regulatory Guide 1.183 Conformance Tables

,_ *... *, **Table

G:,'Gonformance

. With ReQulatorv*Guide

. - 1.183

.. Appendix*H (Rod

. Ejection

' . Accident)

RG Section

. RG .Position, p

FNP Analysis 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 Conforms - The total leakage from the for operation specified in the technical specifications should be assumed to primary system to the secondary exist until shutdown cooling is in operation and releases from the steam system is assumed to be 1 gpm, generators have been terminated. 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 Conforms - The water density of both rates (e.g., lbm/hr) should be consistent with the basis of surveillance tests the primary and secondary coolants is used to show compliance with leak rate technical specifications. These tests assumed to be 62.4 lbm/ft3 .

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 qm/cc (62.4 lbm/ft3 ).

H-7.3 All noble gas radionuclides released to the secondary system are assumed to Conforms - It is assumed that noble be released to the environment without reduction or mitigation. 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 Conforms - The transport model should be utilized for iodine and particulates. described in Position 5.5 and 5.6 of Appendix E is applied to releases from the steam generators.

Footnote Facilities licensed with, or applying for, alternative repair criteria (ARC) should Conforms - FNP is licensed to ARC.

H-1 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 radioloqical analyses.

E5-53 to NL-16-0388 Loss of Coolant Accident Analysis Joseph M. Farley Nuclear Plant - Units 1 and 2 Alternative Source Term License Amendment Request Enclosure 6 Loss-of-Coolant Accident Analysis E6 - 1 to NL-16-0388 Loss of Coolant Accident Analysis LOSS-OF-COOLANT ACCIDENT DOSE CONSE QUENCES USING AST METHODS Licensee Document Number: SM-1080538201-003, Version 3 Method/Computer Program Used: RADTRAD Version 3.10 Regulatory Guidance: RG-1.183, including Appendix A Model Discussion The calculation was performed in four parts, evaluating the contributions from four separate release paths: Containment mini-purge , Containment Leakage, ECCS Leakage Outside of Containment, and potential leakage fro m the Refueling Water Storage Tank (RWST). The dose contributions from each of these pathways were summed to obtain the doses to the Main Control Room (MGR), the Exclusion Area Boundary (EAB), and the Low Population Zone (LPZ).

The accident duration is 30 days, per FNP Current Licensing Basis (CLB).

Results and Acce12tance Limits Release EAB LPZ Control Room

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

Containment Purge 0.001 0.0004 0.002 Containment Leakage 12.9 5.6 3.6 ECCS Leakage 0.25 0.23 0.81 RWST Back-leakage 0.13 0.13 0.26 Total 13.2 6.0 4.7 Acceptance Limit 25 25 5 (Note that rounding is applied to all values)

Key Assum12tions and lnQuts Source Term Parameters Parameter Value Reactor Power Level: 2775 MWt (+2% uncertainty= 2831 MWt)

E6-2 to NL-16-0388 Loss of Coolant Accident Analysis Table 1 - Core Cycle-to-Cycle Augments Isotope Factor Kr-85 1.15 Xe-133 1.05 Cs-134 1.35 Cs-136 1.25 Cs-137 1.20 Halogens, Other Noble Gases and 1.03 Particulates Table 2 - Core Source Term Activity Core Adjusted Activity Core Adjusted Nuclide Nuclide (Ci) Factor Activity (Ci) Factor Activity (Ci) (Ci)

Co-58 O.OOE+OO 1.03 O.OOE+OO Te-134 1.30E+08 1.03 1.34E+08 Co-60 O.OOE+OO 1.03 O.OOE+OO 1-130 2.50E+06 1.03 2.58E+06 Br-82 3.80E+05 1.03 3.91E+05 1-131 7.50E+07 1.03 7.73E+07 Br-83 9.70E+06 1.03 9.99E+06 1-132 1.10E+08 1.03 1.13E+08 Br-84 1.70E+07 1.03 1.75E+07 1-133 1.60E+08 1.03 1.65E+08 Kr-83m 9.70E+06 1.03 9.99E+06 1-134 1.70E+08 1.03 1.75E+08 Kr-85 7.20E+05 1.15 8.28E+05 1-135 1.50E+08 1.03 1.55E+08 Kr-85m 2.10E+07 1.03 2.16E+07 Xe-131m 8.40E+05 1.03 8.65E+05 Kr-87 4.00E+07 1.03 4.12E+07 Xe-133 1.50E+08 1.05 1.58E+08 Kr-88 5.70E+07 1.03 5.87E+07 Xe-133m 4.80E+06 1.03 4.94E+06 Rb-86 1.40E+05 1.03 1.44E+05 Xe-135 3.50E+07 1.03 3.61E+07 Rb-89 7.40E+07 1.03 7.62E+07 Xe-135m 3.00E+07 1.03 3.09E+07 Sr-89 7.70E+07 1.03 7.93E+07 Xe-138 1.30E+08 1.03 1.34E+08 Sr-90 5.70E+06 1.03 5.87E+06 Cs-134 1.10E+07 1.35 1.49E+07 Sr-91 9.50E+07 1.03 9.79E+07 Cs-134m 3.60E+06 1.03 3.71E+06 Sr-92 1.00E+08 1.03 1.03E+08 Cs-136 3.30E+06 1.25 4.13E+06 Y-90 5.90E+06 1.03 6.08E+06 Cs-137 7.60E+06 1.20 9.12E+06 Y-91 9.90E+07 1.03 1.02E+08 Cs-138 1.40E+08 1.03 1.44E+08 Y-91m 5.50E+07 1.03 5.67E+07 Ba-139 1.40E+08 1.03 1.44E+08 Y-92 1.00E+08 1.03 1.03E+08 Ba-140 1.30E+08 1.03 1.34E+08 Y-93 1.20E+08 1.03 1.24E+08 Ba-141 1.30E+08 1.03 1.34E+08 Y-95 1.30E+08 1.03 1.34E+08 La-140 1.40E+08 1.03 1.44E+08 E6-3 to NL-16-0388 Loss of Coolant Accident Analysis Activity Core Adjus~ed Activity Core Adjusted Nuclide Nuclide (Ci) Factor Activity (Ci) Factor Activity (Ci) (Ci)

Zr-95 1.30E+08 1.03 1.34E+08 La-141 1.30E+08 1.03 1.34E+08 Zr-97 1.30E+08 1.03 1.34E+08 La-143 1.20E+08 1.03 1.24E+08 Nb-95 1.30E+08 1.03 1.34E+08 La-142 1.20E+08 1.03 1.24E+08 Nb-95m 9.40E+05 1.03 9.68E+05 Ce-141 1.30E+08 1.03 1.34E+08 Nb-97 1.30E+08 1.03 1.34E+08 Ce-143 1.20E+08 1.03 1.24E+08 Mo-99 1.40E+08 1.03 1.44E+08 Ce-144 9.40E+07 1.03 9.68E+07 Tc-99m 1.20E+08 1.03 1.24E+08 Pr-143 1.20E+08 1.03 1.24E+08 Ru~103 1.10E+08 1.03 1.13E+08 Nd-147 5.10E+07 1.03 5.25E+07 Ru-105 7.60E+07 1.03 7.83E+07 Pm-147 9.70E+06 1.03 9.99E+06 Ru-106 3.40E+07 1.03 3.50E+07 Pm-148 2.10E+07 1.03 2.16E+07 Rh-103m 1.00E+08 1.03 1.03E+08 Pm-148m 2.30E+06 1.03 2.37E+06 Rh-105 6.90E+07 1.03 7.11 E+07 Pm-149 4.60E+07 1.03 4.74E+07 Pd-109 2.20E+07 1.03 2.27E+07 Pm-151 1.50E+07 1.03 1.55E+07 Sb-124 8.80E+04 1.03 9.06E+04 Sm-153 3.30E+07 1.03 3.40E+07 Sb-125 9.40E+05 1.03 9.68E+05 Eu-154 7.10E+05 1.03 7.31E+05 Sb-126 ?.90E+04 1.03 8.14E+04 Eu-155 4.60E+05 1.03 4.74E+05 Sb-127 7.90E+06 1.03 8.14E+06 Eu-156 1.20E+07 1.03 1.24E+07 Sb-129 2.40E+07 1.03 2.47E+07 Np-238 2.30E+07 1.03 2.37E+07 Te-125m 2.00E+05 1.03 2.06E+05 Np-239 1.40E+09 1.03 1.44E+09 Te-127 7.80E+06 1.03 8.03E+06 Pu-238 1.60E+05 1.03 1.65E+05 Te-127m 1.00E+06 1.03 1.03E+06 Pu-239 2.20E+04 1.03 2.27E+04 Te-129 2.40E+07 1.03 2.47E+07 Pu-240 3.10E+04 1.03 3.19E+04 Te-129m 3.50E+06 1.03 3.61E+06 Pu-241 8.30E+06 1.03 8.55E+06 Te-131 6.70E+07 1.03 6.90E+07 Pu-243 1.80E+07 1.03 1.85E+07 Te-131m 1.10E+07 1.03 1.13E+07 Am-241 8.20E+03 1.03 8.45E+03 Te-132 1.10E+08 1.03 1.13E+08 Am-242 4.50E+06 1.03 4.64E+06 Te-133 9.10E+07 1.03 9.37E+07 Cm-242 2.20E+06 1.03 2.27E+06 Te-133m 5.80E+07 1.03 5.97E+07 Cm-244 1.30E+05 1.03 1.34E+05 E6 - 4 to NL-16-0388 Loss of Coolant Accident Analysis Parameter Value Initial RCS Source Term Accounts for 1% Failed (Leaking) Fuel (limitation from previous operating experience, not accident related).

Initial RCS Source Term The Iodine concentration is set at the 0.5

µCi/gm RCS Mass 440,900 lbm Release Fractions Per RG-1.183 Release Timing Per RG-1.183 Table 3 - RCS Source Term Nuclide Activity (µCi/gm)

Kr-85 7.70E+OO Kr-85m 1.80E+OO Kr-87 1.20E+OO Kr-88 3.50E+OO 1-131 3.528E-01 1-132 5.796E-01 1-133 6.804E-01 1-134 1.588E-01 1-135 4.788E-01 Xe-133 2.40E+02 Xe-135 7.90E+OO Kr-83m 4.50E-01 Br-83 8.80E-02 Br-84 5.00E-02 1-130 2.20E-02 Xe-131m 2.90E+OO Xe-133m 4.60E+OO Xe-135m 4.50E-01 Xe-138 7.20E-01 Containment Leakage Parameters Parameter Value Containment Volume 2.03E6 cubic Feet Sprayed Volume 1,668,660 cubic feet Unsprayed Volume 361,240 cubic feet Containment Leakage 0.15% of volume per day for first 24 Hours 0.075% of volume per day for remainder Containment Leakage Filtration None Containment Long Term Sump pH pH ~ 7.0 (no re-evolution of Iodine)

Containment spray removal A, Elemental 13.7hr"1 .

Containment spray removal A, Aerosol 5.45 hr" 1 during injection mode E6 - 5 to NL-16-0388 Loss of Coolant Accident Analysis 5.03 hr" 1 during recirculation mode Containment Spray Organic removal None Natural Deposition, Aerosol only 0.1 hr"1 after sprays are terminated Containment Spray Start 90 seconds Containment Spray Stop 8 Hours Containment Spray Flow 2,480 gal/min in injection phase 2,290 gal/min in recirculation phase Iodine Chemical Form 95% Cesium Iodide, 4.85% elemental, 0.15% organic Containment Purge Leakage Parameter Value Iodine Chemical Form 95% Cesium Iodide, 4.85% elemental, 0.15% organic Containment Purge Filtration None Removal by wall deposition 0%

Removal by Sprays 0%

Containment Purge Isolation :S30 seconds Containment Purge Flowrate 2850 CFM ECCS Leakage Parameter Value Sump Volume 49,200 cubic feet Sump temperature Varies, max is 265 °F ECCS Leakage Initiation Time 20 minutes ECCS Leakage Iodine Flashing Factor 10%

Iodine Species ECCS Leakage Released to the Atmosphere Elemental 97%

Organic 3%

ECCS Leakage Rate 40,000 cc/hr RWST Leakage Parameters Parameter Value ECCS Recirculation Start Time 20 minutes Iodine Species ECCS Leakage Released to the Atmosphere from the RWST Elemental 100%

Organic 0%

ECCS Leakage Rate to the RWST 2 gal/min RWST Leakage Iodine Flashing Factors Varies with temperature and pH HWST Capacity 505,562 gallons RWST Volume at Transfer to Recirculation 29,002 gallons E6- 6

Enclosure 6 to NL-16-0388 Loss of Coolant Accident Analysis CR Parameters:

Parameter Value CR Volume 114,000 ft3 CR Pressurization Mode Initiation Automatic at 60 Seconds 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 iodine species Recirculation Filters 94.5% elemental and organic 98.5% particulate CR Unfiltered In-leakage 315 cfm CR Ingress/Egress Unfiltered In-leakage 10 cfm CR Breathing Rate 3.5E-4 m3/sec Occupancy Factors 0-24 hours 1.0 1 - 4 days 0.6 4 -30 days 0.4 Atmospheric Dispersion Factors (sec/m 3 ):

Containment Releases Time (hr) EAB LPZ CR 0-2 7.6E-4 2.80E-4 1.66E-03 2-8 1.1 OE-4 1.36E-03 8-24 1.00E-5 6.81 E-04 24-96 5.40E-6 5.60E-04 96-720 2.90E-6 4.21 E-04 Plant Vent Releases Time (hr) EAB LPZ CR 0- 0.0167 7.6E-4 2.80E-4 2.79E-03 0.0167- 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 Releases Time (hr) EAB LPZ CR 0-2 7.6E-4 2.80E-4 4.97E-04 2-8 1.1 OE-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 E6 - 7 to NL-16-0388 Fuel Handling Accident Analysis Joseph M. Farley Nuclear Plant - Units 1 and 2 Alternative Source Term License Amendment Request Enclosure 7 Fuel Handling Accident Analysis E7- 1 to NL-16-0388 Fuel Handling Accident Analysis FUEL HANDLING ACCIDENT DOSE CONSEQUENCES USING AST METHODS Licensee Document Number: SM-1080538201-001, Version 2 Method/Computer Program Used: RADTRAD Version 3.03 Regulatory Guidance: RG-1.183, including Appendix B Model Discussion The calculation was performed to address a fuel handling accident (FHA) in the containment and in the SFP area of the Auxiliary Building. For the containment accident, the containment equipment hatch and the personnel airlock are presumed to be open and no credit is taken to close them. The open containment airlock could allow areas around the Control Room (CR) to become contaminated, so the calculation accounts for dose impacts of ingress/egress of the CR through the CR doors. Also, a small amount of CR envelope wall is only 1 foot thick, so the shine from the contaminated area through the wall is added to the CR operator dose. Doses in the CR are accumulated over a period of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Releases from the damaged fuel are completed in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

For the accident in the SFP area of the Auxiliary Building, the accident releases also are completed in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The activity is released to the environment through the plant vent stack, and credit is taken for filtration of the iodine isotopes through the Penetration Room Filtration System. Doses from this accident are bounded by the doses from an accident in containment.

Results and Acceptance Limits EAB LPZ Control Room Release (rem TEDE) (rem TEDE) (rem TEDE)

Containment 2.4 0.9 1.0 Spent Fuel Pool 0.5 0.2 0.2 Acceptance Limit 6.3 6.3 5 (Note that rounding is applied to all values)

Key Assumptions and Inputs Source Term Parameters Parameter Value Reactor Power Level 2775 MWt (+2% uncertainty = 2831 MWt)

Reactor Peaking Factor 1.7 Fuel Movement Time 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months post shutdown.

Number of Fuel Assemblies 157 Number of Damaged Assemblies 1 Number of Damaged Fuel Rods 264 E7-2

Enclosure 7 to NL-16-0388 Fuel Handling Accident Analysis Table 1 - Core Cycle-to-Cycle Augments Isotope Factor Kr-85 1.15 Xe-133 1.05 Other Noble Gases 1.03 Other Iodine isotopes 1.03 Table 2 - Core Source Term Isotope Core Activity at 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months post Shutdown (curies)

Kr-85 7.2E+05 Xe-131m 8.1E+05 Xe-133 1.0E+08 Xe-133m 2.0E+06 Xe-135 2.0E+05 1-131 5.4E+07 1-132 4.6E+07 1-133 5.7E+06 1-135 4.1E+03 Fraction of Fission Product Inventory in Gap 1-131 0.08 Kr -85 0.10 Other Noble Gases 0.05 Other Halogens 0.05 Overlaying Pool Depth 23 feet Pool Decontamination Factor Elemental: 500 Organic: 1 Iodine Chemical Form 0% Aerosol, 99.85% Elemental 0.15% Organic Net Decontamination Factor 200 The released activity is obtained by making the product of the100-hour post-shutdown core activity for the isotope, the design margin, the gap fraction, and the radial peaking factor. The product is then divided by the DF and by 157 (the number of assemblies) to achieve the released activity shown in the last column of the table, below.

E?- 3 to NL-16-0388 Fuel Handling Accident Analysis Table 3 - Net Scrubbed Release Activities 100-hr Radial Released Core Design Gap Peaking Group Isotope OF Activity Inventory Margin Fraction Factor (Curies)

(Ci)

Kr-85 7.20E+05 1.15 0.1 1.70E+OO 1 8.97E+02 2

0 1.70E+OO 1

~

Xe-131m 8.10E+05 1.03 0.05 4.52E+02 tD Ci) Xe-133 1.00E+08 1.05 0.05 1.70E+OO 1 5.68E+04 DJ Ill Xe-133m 2.00E+06 1.03 0.05 1.70E+OO 1 1.12E+03 tD Ill Xe-135 2.00E+05 1.03 0.05 1.70E+OO 1 1.12E+02 1-131 5.40E+07 1.03 0.08 1.70E+OO 200 2.41E+02

c DJ 0

1-132 4.60E+07 1.03 0.05 1.70E+OO 200 1.28E+02 OQ tD 1-133 5.70E+06 1.03 0.05 1.70E+OO 200 1.59E+01

J Ill 1-135 4.10E+03 1.03 0.05 1.70E+OO 200 1.14E-02 Containment Release Parameter Value Containment Volume 2.03E6 Cubic Feet Mixing Volume in Containment 1.0E6 Cubic Feet Release Duration 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Containment Hatch Flow Rate 55,000 cfm Containment Release Filtration 0%

Personnel Airlock Flow Rate 1515 cfm Auxiliary Building Mixing Volume 37 ,875 cubic feet Personnel Airlock Release Filtration 0%

Auxiliary Building Ventilation 1505 cfm to plant vent, 1O cfm to CR for ingress/egress Spent Fuel Pool Area Release Parameter Value Fuel Handling Volume 72, 150 cubic feet Overlaying Pool Depth 23 feet Fuel Handling Area Release Rate 5,000 cfm PRF Filtration 89.5% for iodine isotopes CR Parameters Parameter Value CR Volume 114,000 ft 3 CR Isolation Mode Initiation Automatic at 60 Seconds CR Pressurization Mode Initiation Manually at 21 minutes (20 minutes after isolation)

CR Ventilation System Normal Flow Rate 2340 cf m < 60 seconds CR Ventilation Isolation Mode Flow Rate 600 cfm (1 minute to 21 minutes)

CR Ventilation Pressurization Makeup Rate 375 cfm > 21 minutes E7-4 to NL-16-0388 Fuel Handling Accident Analysis CR Ventilation System Recirculation Flow Rate 2700 cfm > 21 minutes CR Ventilation System Charcoal Filter Efficiencies Pressurization Filters 98.5% all iodine species Recirculation Filters 94.5% all iodine species CR Pressurization Mode Unfiltered In-leakage 325 cfm*

CR Ingress/Egress Unfiltered In-leakage 1O cfm throughout (location changes)

CR Breathing Rate 3.5E-4 m3/sec Occupancy Factors 0-8 hours 1.0 CR Ventilation Summary Table 4 - Control Room Ventilation Summary Time Filtered Flow Unfiltered Flow (CFM) (CFM)

Oto 1 minute 0 2340 1 minute to 21 minutes 0 600 21 minutes to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 375 325

For the FHA in Containment, the 1O CFM for ingress and egress to the CR goes from the Auxiliary Building to the CR through the CR door. For the FHA in the ea, the 1O cfm for ingress and egress is conservatively added to CR through the ventilation system and is unfiltered. This unfiltered inleakage starts at time O and continues through the entire accident (8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ).

Atmospheric Dispersion Factors (sec/m 3 )

Containment Releases:

Time (hr) EAB LPZ CR 0-2 7.6E-4 2.80E-4 8.79E-04 2-8 1.10E-4 6.77E-04 Plant Vent Releases:

Time (hr) EAB LPZ CR 0-2 7.6E-4 2.80E-4 1.62E-03 2-8 1.10E-4 1.37E-03 E7-5 to NL-16-0388 Main Steam Line Break Accident Analysis Joseph M. Farley Nuclear Plant - Units 1 and 2 Alternative Source Term License Amendment Request Enclosure 8 Main Steam Line Break Accident Analysis E8 - 1 to NL-16-0388 Main Steam Line Break Accident Analysis MAIN STEAM LINE BREAK ACCIDENT DOSE CONSEQUENCES USING AST METHODS Licensee Document Number: SM-1080538201-002, Version 1 Method/Computer Program Used: LocaDose Version 7.1 Regulatory Guidance:. RG-1.183, including Appendix E Model Discussion:

The calculation was performed to address a Main Steam Line Break (MSLB). Per RG-1.183, two cases are considered for the dose-equivalent 1-131 (DEi) concentrations in the Reactor Coolant Sysfem (RCS):

1. Pre-Accident Iodine Spike - a reactor transient occurs prior to the accident and raises the RCS iodine concentration to the maximum value permitted by the technical specifications.

2. Concurrent Iodine Spike - 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 500 times greater than the release rate that yields the equilibrium iodine concentration specified in the technical specifications.

Primary to Secondary leakage is assumed to be 0.35 gallons per minute (gpm) to the faulted" steam generator (SG), and 0.65 gpm (total) going to the two intact SGs. It is postulated that the MSLB causes the associated faulted" SG to blow dry, releasing activity directly to the environment through the broken main steam line. Activity from two intact SGs released to the environment via steaming until the primary system (RCS) is reduced to cold shutdown conditions (assumed at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

Doses for the Pre-Accident Iodine Spike Case and the Concurrent Iodine Spike Case were calculated, with results shown below.

Results and Acceptance Limits Case Location Dose (Rem TEDE)

Calculated Limit Pre-Accident Iodine EAB 0.94 25 Spike LPZ 0.37 25 Control Room 0.23 5 Concurrent Iodine EAB 0.95 2.5 Spike LPZ 0.45 2.5 Control Room 0.45 5 The maximum 2-hour EAB dose occurs between O and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

(Note that rounding is applied to all values)

EB -2 to NL-16-0388 Main Steam Line Break Accident Analysis Key Assumptions and Inputs Physical Parameters Parameter Value RCS Mass 440,900 lbm (2.00E8 grams)

RCS Volume 1.02E4 cubic feet Intact SG Mass 1.68E5 lbm (each, assumed full)

Intact SG Volume 2690 cubic feet (each)

Faulted SG Volume 2690 cubic feet Coolant Densities Primary and Secondary water at 62.4 lbm/ft3 Table 1 - MSLB Flow Rates Flow Path Time (hour) Release Flow Note From to (lbm)

RCS to Env 0 24 - 4.68E-02 cfm 1 RCS to Intact SGs 0 24 - 8.69E-02 cfm Feedwater to Intact 0 2 4.81 E+05 1.09E+08 g/hr 2 SGs 2 8 7.83E+05 5.92E+07 g/hr 8 24 1.04E+06 2.96E+07 Q/hr Intact SGs to Env 0 2 3.48E+05 4.65E+01 cfm 3 2 8 7.74E+05 3.45E+01 cfm 8 24 1.04E+06 1.74E+01 cfm 0 24 2.17E+06 -

Faulted SG to Env 0 24 4.83E+05 5.00E+02 cfm 4 Flow Rate Notes:

1. RCS Leakage of 1 gpm - Volumetric leakage (gpm) from RCS is divided by 7.48 gal/ft3.

2. Feedwater - Mass release from feedwater to intact SGs is multiplied by 1.1. Flow is the release (lbm) multiplied by 453.6 g/lbm and divided by time duration (hr).

3. Intact SGs - Mass release from the intact SGs is multiplied by 1.1. Flow is the release (lbm) divided by 62.4 lbm/ft3 and by the time duration (min).

4. Faulted SG - Mass release from the faulted SG is multiplied by 1.1. Flow is conservatively high.

E8-3 to NL-16-0388 Main Steam Line Break Accident Analysis Radioactivity Considerations:

No fuel failure occurs as a result of the MSLB.

Iodine Release Species: 97% elemental, 3% organic.

RCS activity includes an assumption of normal operations 1% failed (leaking) fuel in accordance with past CLB (Affects alkali metals)

Initial RCS Activity:

Pre-Accident Spike Case 30 µCi/gm DEi Concurrent Accident Spike 0.5 µCi/gm DEi Normal Iodine RCS Appearance Rate Letdown Flow Rate 145 gal/min Identified Leakage 10 gal/min Unidentified Leakage 1 gal/min Concurrent Accident Spike Appearance 500 times the Normal Rate for the first 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Concentrations of Iodine in Secondary Technical Specification limit of 0.1

µCi/gm DEi Concentrations of Alkali Metals in Secondary - Given 0.1 µCi/gm Dei in the secondary and 0.5

µCi/gm in the RCS, the concentrations of alkali metals are assumed to be 20% of those in the RCS corresponding to 1% failed (leaking) fuel.

Table 2 - Normal RCS Iodine Concentrations 1% FF RCS lnh TEDE Cone en Concen (µCi/g)

Isotope Concen (µCi/g) DCF (rem/Ci) x DCF 30 µCi/g DEi 0.5 µCi/g DEi 0.1 µCi/g DEi 1-131 2.5 3.29E+04 8.23E+04 2.27E+01 3.78E-01 7.56E-02 1-132 0.9 3.81E+02 3.43E+02 8.17E+OO 1.36E-01 2.72E-02 1-133 4.0 5.85E+03 2.34E+04 3.63E+01 6.0SE-01 1. 21E-01 1-134 0.6 1.31E+02 7.86E+01 5.44E+OO 9.07E-02 1. 81E- 02 1-135 2.2 1.23E+03 2.71E+03 2.00E+01 3.33E-01 6.65E-02 Total 1.09E+OS 9.26E+01 1.54E+OO 3.09E-01 Note 2 3 4 5 6 7 Normal RCS Concentration Notes:

1. The note numbers correspond to column numbers.

2. RCS at 1% failed (leaking) fuel - These are the RCS concentrations corresponding to 1% failed (leaking) fuel (FF).

3. DCF - Inhalation TEDE DCFs are from Federal Guidance Report 11, multiplied by 3.7E12 to convert from Sv/Bq to rem/Ci.

4. DCF Weighted Concentration - The concentrations in Column 2 are multiplied by the DCFs in Column 3. The total for this column represents the relative dose corresponding to 1% FF.

EB - 4 to NL-16-0388 Main Steam Line Break Accident Analysis

5. Concentrations at30 µCi/g DEi -The relative dose corresponding to 30 µCi/g of 1-131 is (30 µCi/g)(3.29E4 rem/Ci)= 9.87E5. In Column 5, the concentrations in Column 2 are multiplied by 9.87E5/1.09E5 to obtain the distribution corresponding to 30 µCi/g DEi.

6. Concentrations at 0.5 µCi/g DEi - The concentrations corresponding to 30 µCi/g in Column 5 are multiplied by 0.5/30.

7. Concentrations at 0.1 µCi/g DEi - The concentrations corresponding to 30 µCi/gin Column 5 are multiplied by 0.1/30.

Initial Iodine Activities in RCS and Secondary Coolant The initial iodine activities in the RCS and the secondary coolant corresponding to 30, 0.5, and 0.1 µCi/g DEi are shown in the following table. These are entered in LocaDose as initial activities within nodes.

Table 3 - Initial Iodine Activities in the RCS and Secondary Coolant Initial Activity (Ci)

RCS Pre Spike RCS Con Spike Intact SGs Faulted SG Isotope (30 µCi/g DEi (0.5 µCi/g DEi) (0.1 µCi/g DEi) (0.1 uCi/g DEi) 1-131 4.5E+03 7.6E+01 7.4E+01 1.7E+01 1-132 1.6E+03 2.7E+01 2.7E+01 6.0E+OO 1-133 7.3E+03 1.2E+02 1.2E+02 2.7E+01 1-134 1.1E+03 1.8E+01 1.8E+01 4.0E+OO 1-135 4.0E+03 6.7E+01 6.5E+01 1.5E+01 Total 1.9E+04 3.1E+02 3.0E+02 6.8E+01 Note 2 3 4 5 Initial Iodine Activities Notes:

1. The note numbers correspond to column numbers.

2. RCS Pre-Accident Spike - The concentrations (µCi/g) in the table above, Column 5 are multiplied by the RCS mass of 2.00E8 g and by 1.0E-6 Ci/µCi.

3. RCS Concurrent Spike - The concentrations (µCi/g) in the table above, Column 6 are multiplied by the RCS mass of 2.00E8 g and by 1.0E-6 Ci/µCi.

4. Intact SGs -The mass of fluid released from the intact SGs is 2.17E6 lbm [Table 1],

which is multiplied by 453.6 g/lbm to yield 9.82E8 g. The concentrations (µCi/g) in Table 2, Column 7 are multiplied by this mass and by 1.0E-6 Ci/µCi.

5. Faulted SG - The mass of fluid released from the faulted SG is 4.83E5 lbm [Table 1],

which is multiplied by 453.6 g/lbm to yield 2.19E8 g. The concentrations (µCi/g) in Table 2, Column 7 are multiplied by this mass and by 1.0E-6 Ci/µCi.

E8-5 to NL-16-0388 Main Steam Line Break Accident Analysis Table 4 - Noble Gases and Alkali Metals in RCS (1 % failed (leaking) fuel)

Cone en Activity Isotope (µCi/g) (Ci)

Kr-85m 2.2E+OO 4.4E+02 Kr-85 S.SE+OO 1.1E+03 Kr-87 1. 3E+OO 2.6E+02 Kr-88 3.8E+OO 7.6E+02 Xe-133m 3.2E+OO 6.4E+02 Xe-133 2.9E+02 S.8E+04 Xe-135m 2.0E-01 4.0E+Ol Xe-135 6.lE+OO l.2E+03 Xe-138 7.0E-01 1.4E+02 Rb-88 3.SE+OO 7.6E+02 Rb-89 1. OE-01 2.0E+Ol Cs-134 2.6E-Ol S.2E+Ol Cs-136 1. SE-01 3.0E+Ol Cs-137 1.3E+OO 2.6E+02 Cs-138 9.6E-Ol 1. 9E+02 Total 3.2E+02 6.4E+04 Note 2 3 Noble Gas and Alkali Metals in RCS Notes:

1. The note numbers correspond to the column numbers.

2. RCS Concentrations - These correspond to a 1% FF Assumption.

3. RCS Activities - The concentrations (µCi/g) in Column 2 are multiplied by the RCS mass of 2.00E8 g and by 1.0E-6 Ci/µCi.

EB- 6 to NL-16-0388 Main Steam Line Break Accident Analysis Alkali Metals in the Secondary System The initial concentrations are assumed to be 20% of the RCS initial activities.

Table 5 - Initial Alkali Metal Concentrations in the Secondary System Activity 1 Curies)

Isotope Concentration (µCi/g) Intact SGs Faulted SG Rb-88 7.6E-01 7.5E+02 1.7E+02 Rb-89 2.0E-02 2.0E-01 4.4E+OO Cs-134 5.2E-02 5.1 E+01 1.1E+01 Cs-136 3.0E-02 2.9E+01 6.6E+OO Cs-137 2.6E-01 2.6E+02 5.7E+01 Cs-138 1.9E-01 1.9E+02 4.2E+01 Total 1.3E+OO 1.3E+03 2.9E+02 Note 2 3 4 Initial Alkali Metals in the Secondary System Notes:

1. The note numbers correspond to the columns of the table.

2. Concentrations - the initial RCS concentrations shown in Table 4 are multiplied by 0.2 to obtain secondary coolant concentrations.

3. Intact SGs Activities - the mass of the fluid released from the intact SGs is 2.17E6 lbm, which is multiplied by 453.6 gm/lbm to yield 9.82E8 grams. The concentrations (in

µCi/gm) in Column 2 of this table are multiplied by this mass and then by 1.0E-6 Ci/µCi.

4. Faulted SG Activities - the mass of the fluid released from the faulted SG is 4.83E5 lbm

[Table 1], which is multiplied 453.6 gm/lbm to yield 2.19E8 grams. The concentrations (in

µCi/gm) in Column 2 of this table are multiplied by this mass and then by 1.0E-6 Ci/µCi.

EB- 7 to NL-16-0388 Main Steam Line Break Accident Analysis Radioiodine Appearance Rates Table 6 - RCS Iodine Appearance Rates Decay Total Removal Appearance Rate (Ci/hr)

Isotope Rate (sec- 1 ) Rate (h( 1) Normal Con Spike 1-131 9.98E-07 1.27E-01 9.6E+OO 4.8E+03 1-132 8.43E-OS 4.27E-01 1.2E+01 S.8E+03 1-133 9.21E-06 1.56E-01 1.9E+01 9.SE+03

~-- ----~---*- -

1-134 2.20E-04 9.1SE-01 1.7E+01 8.3E+03 1-135 2.91E-05 2.28E-01 1.SE+01 7.6E+03 Note 2 3 4 5 RCS Iodine Appearance Rates Notes:

1. The note numbers correspond to the column of the table.

2. Decay rates are from the LocaDose manual

3. Total Removal Rate - the decay rate is multiplied by 3600 sec/hr and added to the clean-up rate of 1.14E-1 h(1 and the leakage rate of 9.42E-3 h( 1 *

4. Normal Appearance Rate - the initial iodine activity (Ci) from Table 3 above (column 3) is multiplied by the total removal rate (h( 1) in column 3 of this table.

5. Concurrent Spike Appearance Rate - the normal appearance rate in column 4 of this table is multiplied by 500.

Iodine Appearance Rates in Intact SGs from Feedwater The flow of feedwater into the intact SGs is modeled as an activity production term in LocaDose.

The concentrations corresponding to 0.1 µCi/g from Table 2, Column 7 are multiplied by 1.0E-6 Ci/µCi and by time-dependent flow rates (g/hr) from Table 1, yielding the following appearance rates. These are entered in LocaDose as production terms.

Table 7 - Iodine Appearance Rates in Intact SGs Isotope Feedwater Appearance Rate Ci/hr) 0-2 Hour 2 - 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8-24 hour 1-131 8.3E+OO 4.5E+OO 2.2E+OO 1-132 3.0E+OO 1.6E+OO 8.0E-01 1-133 1.3E+01 7.2E+OO 3.6E+OO 1-134 2.0E+OO 1.1 E+OO 5.4E-01 1-135 7.3E+OO 3.9E+OO 2.0E+OO EB - 8 to NL-16-0388 Main Steam Line Break Accident Analysis Control Room Ventilation Parameters:

Parameter Value Pressurization Mode starts Initiated at start of accident.

CR Make-up Flow Rate 375 cfm (throughout accident)

Pressurization Unfiltered In-leakage 300 cfm (throughout accident)

CR Ingress/egress 1O cfm (throughout accident through CR Vent)

CR Volume 114,000 cubic feet CR Pressurization Filters:

HEPA 98.5% for all particulates Charcoal 98.5% for all iodine species CR Recirculation Flow 2700 cfm (throughout accident)

Iodine Filter Efficiency 94.5% for organic and elemental 98.5% for particulates CR Breathing Rates 3.5E-04 m 3/sec for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months EAB & LPZ Breathing Rates 3.5E-04 m 3/sec for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Atmospheric Dispersion Factors:

Table 8 - Atmospheric Dispersion Factors X/Q (sec/m 3 )

EAB LPZ Control Time (hr) I Room 0-2 7.60E-4 2.BOE-4 1.66E-3 2-8 I - 1.10E-4 1.38E-3 8-24 - 1.0E-05 7.20E-04 Note: The calculation of record for the MSLB has a typographical error in the assumption section stating that the X/Q for the 8 - 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period at the LPZ is 1.1 E-05 sec/m 3

1.0E-05 3

sec/m was used in the LocaDose input files.

EB - 9 to N L-16-0388 Steam Generator Tube Rupture Accident Analysis Joseph M. Farley Nuclear Plant - Units 1 and 2 Alternative Source Term License Amendment Request Enclosure 9 Steam Generator Tube Rupture Accident Analysis E9 - 1 to NL-16-0388 Steam Generator Tube Rupture Accident Analysis STEAM GENERATOR TUBE RUPTURE ACCIDENT DOSE CONSEQUENCES USING AST METHODS Licensee Document Number: SM-1080538201-005, Version 2 Method/Computer Program Used: LocaDose Version 7.11 Regulatory Guidance: RG-1.183, including Appendix F Model Discussion:

The calculation was performed to address a steam generator tube rupture (SGTR). Mass transfers from the primary to secondary are calculated using the conservative hand-calculation method, in accordance with the FNP current licensing basis (CLB). One steam generator (SG) tube is assumed to fail, rupturing cleanly in two. Mass transfer from the primary to the secondary continues until the break flow is terminated. Activity is released to the environment from the faulted generator until operator action is taken to isolate it. Break low and release Isolation is assumed to occur in 30 minutes for the faulted SG.

Primary to secondary leakage through pin-hole leaks in the SG tubes is assumed at a rate of 1 gpm (0.35 gpm to the faulted SG, 0.65 gpm to the intact SGs) until the SGs are isolated or no longer used for cooling. Activity is released from the other two, intact, generators through steaming via the atmospheric relief valves (ARVs) until the primary system (RCS) is reduced to cold shutdown conditions (assumed at 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ).

Doses for the Pre-Accident Iodine Spike Case and the Concurrent Iodine Spike Case were calculated, with results shown below.

Results and Acceptance Limits:

Dose (rem TEDE)

Case Location Calculated Limit Pre-Accident EAB 2.4 25 Iodine Spike LPZ 0.92 25 Control Room 0.48 5 Concurrent EAB 0.82 2.5 Iodine Spike LPZ 0.34 2.5 Control Room 0.17 5 (Note that rounding is applied to all values)

E9 - 2 to NL-16-0388 Steam Generator Tube Rupture Accident Analysis Key Assumptions and Inputs:

Transient Timing Tube Rupture: Time Zero (O)

Reactor Trip: 324 seconds Faulted Generator Isolated 30 min Break Flow Terminated 30 min ARV Release from Faulted SG Ended 30 min RCS cooled to Cold Shutdown 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Accident Ends 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Physical Parameters:

Parameter Value RCS Mass 440,900 lbm (2.00E8 grams)

RCS Volume 1.02E4 cubic feet Intact SG Mass 1.05E5 lbm (each, assumed full)

Intact SG Volume 1685 cubic feet (each)

Faulted SG Volume 1685 cubic feet Coolant Densities Primary and Secondary water at 62.4 lbm/ft3 Table 1 - SGTR Flow Rates Time (hr) Release Iodine Flow Path From I To (lbm) Flow PF Note RCS to Env 0 8 - 4.68E-02 cfm 1.00E+OO 1

RCS to Intact SGs 0 I 8 - 8. 69E-02 cfm 1.00E+OO RCS to Env 0 0.090 3.12E+04 9. 27E+01 cfm 4.76E+OO 2

(Rupture Flow) 0.090 0.5 1.28E+05 s. 32E+01 cfm 6.67E+OO RCS to Faulted SG 0 0.090 3.12E+04 9. 27E+01 cfm 1.27E+OO 3

(Rupture Flow) 0.090 0.5 1.28E+05 8. 32E+01 cfm 1.18E+OO Faulted SG 0 0.090 3.67E+05 1. 09E+03 cfm 1.00E+02 to Env 0.090 0.5 7.90E+04 5 .15E+01 cfm 1.00E+02 4 0 0.5 4.46E+05 - -

FW to 0.090 2 3.27E+05 7.77E+07 g/hr 1.00E+OO 5

Intact SGs 2 8 9.81E+05 7. 42E+07 g/hr 1.00E+OO Intact SGs 0 0.090 7.34E+05 2. 18E+03 cfm 1.00E+02 to Env 0.090 2 4.22E+05 5. 9 OE+01 cfm 1.00E+02 6

2 8 9.34E+05 4 .16E+01 cfm 1.00E+02 0 8 2.09E+06 - -

SGTR Flow Rates Notes:

1. RCS Leakage of 1 gpm - Volumetric leakage (gpm) from RCS is divided by 7.48 gal/ft3.

2. RCS Flow Through Rupture to Environment - Mass release from RCS to the faulted SG is adjusted for reactor trip time:

E9 - 3 to NL-16-0388 Steam Generator Tube Rupture Accident Analysis 0 to 0.090 hr: (21600 lbm)(324 sec)/(224 sec)= 3.12E4 lbm 0.090 to 0.5 hr: (136400 lbm).(1800-324 sec)/(1800-224 sec)= 1.28E5 lbm.

Flow is the release (lbm) divided by 62.4 lbm/ft3 and by the time duration (min). The PFs correspond to flashing fractions.

3. RCS Flow Through Rupture to Faulted SG - Flows are the same as to the environment but the PFs correspond to the non-flashing fractions.

4. Faulted SG to Env - Time-dependent releases are converted into flows. The release from Oto 0.090 hr is calculated as follows: (1133 lbm/sec)(324 sec) = 3.67E5 lbm. Flow is the release (lbm} divided by 62.4 lbm/ft3 and by the time duration (min). PF is 100 for iodine and 1000 for alkali metals.

5. Feedwater to Intact SGs - Mass release (lbm) is multiplied by 453.6 g/lbm and divided by the time duration (hr).

6. Intact SGs to Env - Time-dependent releases are converted into flows. The release from 0 to 0.090 hr is as follows: (2)(1133 lbm/sec}(324 sec)= 7.34E5 lbm. Flow is the release (lbm) divided by 62.4 lbm/ft3 and by the time duration (min). PF is 100 for iodine and 1000 for alkali metals.

Radioactivity Considerations:

No fuel failure occurs as a result of the SGTR.

Iodine Release Species: 97% elemental, 3% organic.

Initial RCS activity includes an assumption of 1% failed (leaking) fuel in accordance with past CLB (Affects alkali metals)

Initial RCS Activity:

Pre-Accident Spike Case 30 µCi/gm Concurrent Accident Spike 0.5 µCi/gm Normal Iodine RCS Appearance Rate Letdown Flow Rate 145 gal/min Identified Leakage 10 gal/min Unidentified Leakage 1 gal/min Concurrent Accident Spike Appearance 335 times the Normal Rate for the first 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Initial Iodine Concentrations in RCS and Secondary Coolant DEi concentrations of 30, 0.5, and 0.1 µCi/g are converted to their dose-equivalent values of 1-131 to 1-135 as shown in the following table. The concentrations in the last three columns are used to calculate initial activities within nodes in Table 3.

E9 - 4 to NL-16-0388 Steam Generator Tube Rupture Accident Analysis Table 2 - Iodine Concentrations in RCS and Secondary Coolant 1% FF RCS lnh TEDE Concen Concen (µCi/g)

Isotope Concen (µCi/g) DCF (rem/Ci) x DCF 30 µCi/g DEi 0.5 µCi/g DEi 0.1 µCi/g DEi 1-131 2.5 3.29E+04 8.23E+04 2.27E+01 3.78E-01 7.56E-02 1-132 0.9 3.81E+02 3.43E+02 8.17E+OO 1.36E-01 2.72E-02 1-133 4.0 5.85E+03 2.34E+04 3.63E+01 6.05E-01 1.21E-01 1-134 0.6 1.31E+02 7.86E+01 5.44E+OO 9.07E-02 1.81E-02 1-135 2.2 1.23E+03 2.71E+03 2.00E+01 3.33E-01 6.65E-02 Total 1. 09E+05 9.26E+01 1.54E+OO 3.09E-01 Note 2 3 4 5 6 7 Iodine Concentrations in RCS and Secondary Coolant Notes:

1. The note numbers correspond to column numbers.

2. Initial RCS activity considers 1% Failed (Leaking) Fuel.

3. DCF - Inhalation TEDE DCFs are from Federal Guidance Report 11, multiplied by 3.7E12 to convert from Sv/Bq to rem/Ci.

4. DCF Weighted Concentration - The concentrations in Column 2 are multiplied by the DCFs in Column 3. The total for this column represents the relative dose corresponding to 1% FF.

5. Concentrations at 30 µCi/g DEi -The relative dose corresponding to 30 µCi/g of 1-131 is (30 µCi/g)(3.29E4 rem/Ci) = 9.87E5. In Column 5, the concentrations in Column 2 are multiplied by 9.87E5/1.09E5 to obtain the distribution corresponding to 30 µCi/g DEi.

6. Concentrations at 0.5 µCi/g DEi - The concentrations corresponding to 30 µCi/g in Column 5 are multiplied by 0.5/30.

7. Concentrations at 0.1 µCi/g DEi - The concentrations corresponding to 30 µCi/gin Column 5 are multiplied by 0.1/30.

Initial Iodine Activities in RCS and Secondary Coolant The initial iodine activities in the RCS and the secondary coolant corresponding to 30, 0.5, and 0.1 µCi/g DEi are shown in the following table. These are entered in LocaDose as initial activities within nodes.

Table 3 - Initial Iodine Activities Initial Activity (Ci)

RCS Pre Spike RCS Con Spike Intact SGs Faulted SG Isotope (30 µCi/g DEi) (0.5 µCi/g DEi) {0.1 µCi/g DEi) (0.1 µCi/g DEi) 1-131 4.5E+03 7.6E+01 7.2E+01 1. 5E+01 1-132 1.6E+03 2.7E+01 2.6E+01 5.5E+OO 1-133 7.3E+03 1.2E+02 1.1E+02 2.4E+01 1-134 1.1E+03 1. 8E+01 1.7E+01 3.7E+OO 1-135 4.0E+03 6.7E+01 6.3E+01 1.3E+01 Total 1. 9E+04 3.1E+02 2.9E+02 6.2E+01 Note 2 3 4 5 E9 - 5

- - to NL-16-0388 Steam Generator Tube Rupture Accident Analysis Initial Iodine Activities Notes:

1. The note numbers correspond to column numbers.

2. RCS Pre-Accident Spike - The concentrations (µCi/g) in Table 2, Column 5 are multiplied by the RCS mass of 2.00E8 g and by 1.0E-6 Ci/µCi.

3. RCS Concurrent Sp,ike - The concentrations (µCi/g) in Table 2, Column 6 are multiplied by the RCS mass of 2.00E8 g and by 1.0E-6 Ci/µCi.

4. Intact SGs - The mass of fluid released from the intact SGs is 2.09E6 lbm [Table 1],

which is multiplied by 453.6 g/lbm to yield 9.48E8 g. The concentrations (µCi/g) in Table 2, Column 7 are multiplied by this mass and by 1.0E-6 CVµCi.

5. Faulted SG - The mass of fluid released from the faulted SG is 4.46E5 lbm [Table 1],

which is multiplied by 453.6 g/lbm to yield 2.02E8 g. The concentrations (µCi/g) in Table 2, Error! Reference source not found.Column 7 are multiplied by this mass and by 1.0E-6 Ci/µCi.

Noble Gases and Alkali Metals in RCS The initial RCS activities of noble gases and alkali metals corresponding to 1% failed (leaking) fuel are shown in the following table. The values in the last column are entered in LocaDose as initial activities.

Table 4 - Noble Gases and Alkali Metals in RCS Cone en Activity Isotope (µCi/g) (Ci)

Kr-85m 2.2E+OO 4.4E+02 Kr-85 S.SE+OO 1.1E+03 Kr-87 1.3E+OO 2.6E+02 Kr-88 3.SE+OO 7.6E+02 Xe-133m 3.2E+OO 6.4E+02 Xe-133 2.9E+02 S.8E+04 Xe-135m 2.0E-01 4.0E+Ol Xe-135 6.lE+OO 1. 2E+03 Xe-138 7.0E-01 l.4E+02 Rb-88 3.SE+OO 7.6E+02 Rb-89 l.OE-01 2.0E+Ol Cs-134 2.6E-01 S.2E+Ol Cs-136 1.SE-01 3.0E+Ol Cs-137 1.3E+OO 2.6E+02 Cs-138 9. 6E-*01 1. 9E+02 Total 3.2E+02 6 .. 4E+04 Note 2 3 Noble Gases and Alkali Metals in RCS Notes:

1. The note numbers correspond to the column numbers.

2. Initial RCS Concentrations - These correspond to a 1% Failed (Leaking) Fuel Assumption.

E9- 6

Enclosure 9 to NL-16-0388 Steam Generator Tube Rupture Accident Analysis

3. RCS Activities - The concentrations (µCi/g) in Column 2 are multiplied by the RCS mass of 2.00E8 g and by 1.0E-6 Ci/µCi.

Alkali Metals in Secondary Coolant The initial activities of alkali metals in the secondary coolant are shown in the following table, corresponding to 20% of the RCS values at 1% failed (leaking) fuel. These are entered in LocaDose as initial activities within nodes.

Table 5 - Alkali Metals in the Secondary System:

Concen Activity (Ci)

Isotope (µCi/g) Intact SGs Faulted SG Rb-88 7.6E-01 7.2E+02 1. SE+02 Rb-89 2.0E-02 1. 9E+01 4.0E+OO Cs-134 S.2E-02 4.9E+01 1.1E+01 Cs-136 3.0E-02 2.BE+01 6.1E+OO Cs-137 2.6E-01 2.SE+02 S.3E+01 Cs-138 1.9E-01 1. BE+02 3.9E+01 Total 1.3E+OO 1. 2E+03 2.7E+02 Note 2 3 4 Initial Alkali Metals in the Secondary System Notes:

1. The note numbers correspond to the columns of the table.

2. Concentrations -the initial RCS concentrations shown in Table 4 are multiplied by 0.2.

3. Intact SGs Activities - the mass of the fluid released from the intact SGs is 2.09E6 lbm

[Table 1], which is multiplied by 453.6 gm/lbm to yield 9.48E8 grams. The concentrations (in µCi/gm) in Column 2 of this table are multiplied by this mass and then by 1.0E-6 Ci/µCi.

4. Faulted SG Activities - the mass of the fluid released from the faulted SG is 4.46E5 lbm

[Table 1], which is multiplied 453.6 gm/lbm to yield 2.02E8 grams. The concentrations (in

µCi/gm) in Column 2 of this table are multiplied by this mass and then by 1.0E-6 Ci/µCi.

Iodine Appearance Rates in RCS In the following table, the total removal rate for each isotope is determined based on the three removal terms. For equilibrium conditions, the production rate is equal to the removal rate. The iodine concentrations in the RCS are multiplied by the removal rates, thereby yielding the production rates. The values in the last column are entered in LocaDose as production terms for the concurrent spike case.

E9 - 7 I

I to NL-16-0388 Steam Generator Tube Rupture Accident Analysis Table 6 - RCS Appearance Rates Decay Total Removal Appearance Rate (Ci/hr)

Isotope Rate (sec- 1 ) Rate (h( 1 ) Normal Con Spike 1-131 9.98E-07 1.27E-01 9.6E+OO 3.2E+03 1-132 8.43E-05 4.27E-01 1.2E+01 3.9E+03 1-133 9.21E-06 1.56E-01 1.9E+01 6.3E+03 1-134 2.20E-04 9 .. 15E-01 1.7E+Ol S.6E+03 1-135 2.91E-05 2.28E-01 1.SE+Ol S.1E+03 Note 2 3 4 5 RCS Iodine Appearance Rates Notes:

1. The note numbers correspond to the column of the table.

2. Decay rates are from the LocaDose manual

3. Total Removal Rate - the decay rate is multiplies by 3600 sec/hr and added to the clean-up rate of 1.14 hr-1 and the leakage rate of 9.42E-3 hr-1.

4. Normal Appearance Rate -the initial iodine activity (ci) from Table 3 (column 3) is multiplied by the total removal rate (hr-1) in column 3 of this table.

5. Concurrent Spike Appearance Rate -the normal appearance rate in column 4 of this table is multiplied by 335.

Iodine Appearance Rates in the Intact SGs From Feedwater The Feedwater system flows are modelled as a source of radioiodine for this analysis. The Technical Specification limit is 0.1 µCi/g. The secondary system iodine concentrations are shown in Table 2, Column 7. These concentrations are multiplied by 1.0E-6 Ci/µCi and by time-dependent flow rates (g/hr) shown in Table 1 to obtain the following:

Table 7 - Iodine Appearance Rates in the Intact SGs Feedwater Appearance Rate (Ci/hr)

Isotope 0.09-2 hr I 2-8 hr 1-131 S.9E+OO S.6E+OO 1-132 2.1E+OO 2.0E+OO 1-133 9.4E+OO 9.0E+OO 1-134 1.4E+OO 1. 3E+OO 1-135 S.2E+OO 4.9E+OO Control Room Ventilation Parameters:

Parameter Value Pressurization Mode starts Initiated at start of accident.

CR Make-up Flow Rate 375 cfm (throughout accident)

Pressurization Unfiltered In-leakage 300 cfm (throughout accident)

CR Ingress/egress 1O cfm (throughout accident through CR Vent)

E9- 8 to NL-16-0388 Steam Generator Tube Rupture Accident Analysis CR Volume 114,000 cubic feet CR Pressurization Filters HEPA 98.5% for all particulates Charcoal 98.5% for all iodine species CR Recirculation Flow 2700 cfm (throughout accident)

Iodine Filter Efficiency 94.5% for organic and elemental 98.5% for particulates CR Breathing Rates 3.5E-04 m3/sec for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months EAB & LPZ Breathing Rates 3.5E-04 m3/sec for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Atmospheric Dispersion Factors:

Table 8 - Atmospheric Dispersion Factors x!Q (sec/m 3 )

EAB LPZ Control Time (hr) Room 0-2 7.60E-4 2.80E-4 1.66E-3 2-8 - 1.10E-4 1.38E-3 E9 - 9 Oto NL-16-0388 Control Rod Ejection Accident Analysis Joseph M. Farley Nuclear Plant - Units 1 and 2 Alternative Source Term License Amendment Request Enclosure 1O Control Rod Ejection Accident Analysis E10 - 1 0 to NL-16-0388 Control Rod Ejection Accident Analysis CONTROL ROD EJECTION ACCIDENT DOSE CONSEQUENCES USING AST METHODS Licensee Document Number: SM-1080538201-004, Version 1 Method/Computer Program Used: LocaDose Version 7.1 Regulatory Guidance: RG-1.183, including Appendix H Model Discussion The calculation was performed to address a Control Hod Ejection Accident (CREA). The scenario for the CREA is that the reactivity excursion due to a control rod ejection leads to localized fuel damage. The local fuel damage results in increased radioactivity in the Reactor Coolant System (RCS). Activity in the steam generators (SG) due to primary-to-secondary leakage is released to the environment via steaming until cold shutdown conditions are established in the RCS.

To release pathways are considered, in accordance with RG-1.183:

Containment Leakage - 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. "

Secondary System Release - 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 eventually to the environment.

Results and Acceptance Limits Location Dose (Rem TEDE)

Calculated Limit EAB 3.8 6.3 LPZ 2.7 6.3 Control Room 3.7 5 The maximum 2-hour EAB dose occurs between O and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

(Note that rounding is applied to all values)

E10 - 2

Enclosure 10 to NL-16-0388 Control Rod Ejection Accident Analysis Key Assumptions and Inputs Physical Parameters Parameter Value Reactor Power Level 2775 MWt (+2% uncertainty= 2831 MWt)

Containment Volume 2.03E6 ft3 Containment Leakage 0.15% per day for first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.075% per day after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Particulate Removal 2.74E-2 per hour, credit is taken for natural deposition in Containment per NUREG/CR-6189 (Table 36)

RCS Mass 440,900 lbm (2.00E8 grams)

RCS Volume 1.02E4 cubic feet SG Mass 1.68E5 lbm (per SG, which is assumed to be full)

SG Volume 2693 cubic feet (each)

Coolant Densities Primary and Secondary water at 62.4 lbm/ft3 Partition Factors Iodine PF= 100, Alkali Metals PF = 1000 (moisture carryover

= 0.1%)

Noble Gases PF = 1 Primary to Secondary Leakage 1 gpm total, for the first 2500 seconds of the accident.

Secondary System Mass releases 468,600 lbm (426,000+10% margin) in first 98 seconds.

Table 1 - Flow Rates Flow Path Time (hour) Flow Note From t0 RCS to Env 0 0. 694 1.34E-01 cfm 1 Containment to the 0 24 2.11 E+OO cfm 2 Environment 24 720 1.06E+OO cfm SGs to Env 0 0.027 4.60E+03 cfm 3 Flow Rate Notes:

1. RCS Leakage of 1 gpm - Volumetric leakage (gpm) from RCS is divided by 7.48 gal/ft3.

2. Containment- Volume of 2.03E6 ft3 is multiplied by 0.0015/day and divided by 1440 min/day for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the flow is halved.

3. SGs - Mass release from the intact SGs of 4.26E5 lbm is multiplied by 1.1. Flow is the release (lbm) divided by 62.4 lbm/ft3 and by the time duration (min).

E10 - 3 0 to NL-16-0388 Control Rod Ejection Accident Analysis Radioactivity Considerations

0.25% of Fuel Rods experience Melting.

100% of the noble gases and 50% of the Iodine isotopes within the melting rods are available for release form the containment and RCS for the containment and secondary system pathways.

10% of the fuel rods experience cladding failure. A radial power peaking factor of 1.7 is applied to the damaged rods.

The fractions of fission product inventory contained within the fuel rod gaps are:

o Iodine isotopes and Noble gases 0.1 O o Other Halogens 0.05 o Alkali Metals 0.12

Core Fission product inventories are taken from an equilibrium cycle based upon a power level of 2831 MWt. To account for potential cycle-to-cycle variations, the following margin factors are applied to the core inventory:

o Kr-85 1.15 o Xe-133 1.05 o Cs-134 1.35 o Cs-136 1.25 o Cs-137 1.20 o Iodine isotopes and other Noble Gases 1.02 o Other Isotopes 1.03

100% of the activity released to from the core due to fuel melting and cladding failure is instantaneously released to and uniformly mixed in the containment at the onset of the accident.

100% of the activity released from the core due to fuel melting and cladding failure is instantaneously mixed within the RCS at the onset of the accident. Compared to the gap release, any RCS iodine activity due to spiking is negligible.

Chemical form of iodine released to containment is 95% particulate, 4.85 elemental, and 0.15% organic. The containment distribution is used for the secondary system release pathway, because the removal mechanism for this pathway is the same for all chemical forms of iodine.

Radial peaking factor for rods with cladding damage is assumed to be 1.7.

RCS activity includes an assumption of normal operations 1% failed (leaking) fuel in accordance with current licensing basis (affects alkali metals).

The radioiodine concentration in the secondary system is assumed to be at the Technical Specification limit of 0.1 µCi/gm DEi

The concentrations of Alkali Metals in Secondary are based upon a ration of the concentration in the RCS: Given 0.1 µCi/gm DEi in the secondary and 0.5 µCi/gm in the RCS, the concentrations of alkali metals in the secondary are assumed to be 20% of those in the RCS.

E10-4 0 to NL-16-0388 Control Rod Ejection Accident Analysis Containment and RCS Activities Table 2 Reports the Containment and RCS Activities.

The Activity in the RCS or Containment is the sum of the activity from Fuel melting plus the activity released from the gap of the damaged fuel.

Table 2 - Containment and RCS Activities Core Activities (curies)

Activity Margin Core Gap Fuel Gap Isotope (curies) Factor release fraction Melt release Total 1-131 7.50E+07 1.02 0.50 0.10 9.6E+04 1.3E+06 1.4E+06 1-132 1.10E+08 1.02 0.50 0.10 1.4E+05 1.9E+06 2.0E+06 1-133 1.60E+08 1.02 0.50 0.10 2.0E+05 2.8E+06 3.0E+06 1-134 1.70E+08 1.02 0.50 0.10 2.2E+05 2.9E+06 3.2E+06 1-135 1.50E+08 1.02 0.50 0.10 1.9E+05 2.6E+06 2.8E+06 Kr-83m 9.70E+06 1.02 1.00 0.10 2.5E+04 1.7E+05 1.9E+05 Kr-85m 2.10E+07 1.02 1.00 0.10 5.4E+04 3.6E+05 4.2E+05 Kr-85 7.20E+05 1.15 1.00 0.10 2.1 E+03 1.4E+04 1.6E+04 Kr-87 4.00E+07 1.02 1.00 0.10 1.0E+05 6.9E+05 8.0E+05 Kr-88 5.70E+07 1.02 1.00 0.10 1.5E+05 9.9E+05 1.1 E+06 Kr-89 6.90E+07 1.02 1.00 0.10 1.8E+05 1.2E+06 1.4E+06 Xe-131m 8.40E+05 1.02 1.00 0.10 2.1E+03 1.5E+04 1.7E+04 Xe-133m 4.80E+06 1.02 1.00 0.10 1.2E+04 8.3E+04 9.5E+04 Xe-133 1.50E+08 1.05 1.00 0.10 3.9E+05 2.7E+06 3.1E+06 Xe-135m 3.00E+07 1.02 1.00 0.10 7.7E+04 5.2E+05 6.QE+05 Xe-135 3.50E+07 1.02 1.00 0.10 8.9E+04 6.1 E+05 7.0E+05 Xe-137 1.40E+08 1.02 1.00 0.10 3.6E+05 2.4E+06 2.8E+06 Xe-138 1.30E+08 1.02 1.00 0.10 3.3E+05 2.3E+06 2.6E+06 Br-82 3.80E+05 1.03 1.00 0.05 9.8E+02 3.3E+03 4.3E+03 E10 - 5 0 to NL-16-0388 Control Rod Ejection Accident Analysis Core Activities (curies)

Activity Margin Core Gap Fuel Gap Isotope (curies) Factor release fraction Melt release Total Br-83 9.70E+06 1.03 1.00 0.05 2.5E+04 8.5E+04 1.1 E+05 Br-84 1.70E+07 1.03 1.00 0.05 4.4E+04 1.5E+05 1.9E+05 Br-85 2.10E+07 1.03 1.00 0.05 5.4E+04 1.8E+05 2.4E+05 Br-86 1.50E+07 1.03 1.00 0.05 3.9E+04 1.3E+05 1.7E+05 Br-87 3.40E+07 1.03 1.00 0.05 8.8E+04 3.0E+05 3.9E+05 Br-88 3.60E+07 1.03 1.00 0.05 9.3E+04 3.2E+05 4.1E+05 Rb-86 1.40E+05 1.03 1.00 0.05 3.6E+02 1.2E+03 1.6E+03 Rb-88 5.70E+07 1.03 1.00 0.05 1.5E+05 5.0E+05 6.5E+05 Rb-89 7.40E+07 1.03 1.00 0.05 1.9E+05 6.5E+05 8.4E+05 Rb-90 7.20E+07 1.03 1.00 0.05 1.9E+05 6.3E+05 8.2E+05 Rb-91 8.90E+07 1.03 1.00 0.05 2.3E+05 7.8E+05 1.0E+06 Cs-134m 3.60E+06 1.03 1.00 0.05 9.3E+03 3.2E+04 4.1E+04 Cs-134 1.10E+07 1.35 1.00 0.05 3.7E+04 1.3E+05 1.6E+05 Cs-136 3.30E+06 1.25 1.00 0.05 1.0E+04 3.5E+04 4.5E+04 Cs-137 7.60E+06 1.20 1.00 0.05 2.3E+04 7.8E+04 1.0E+05 Cs-138 1.40E+08 1.03 1.00 0.05 3.6E+05 1.2E+06 1.6E+06 Cs-139 1.40E+08 1.03 1.00 0.05 3.6E+05 1.2E+06 1.6E+06 Cs-140 1.20E+08 1.03 1.00 0.05 3.1E+05 1.1 E+06 1.4E+06 Cs-141 9.10E+07 1.03 1.00 0.05 2.3E+05 8.0E+05 1.0E+06 Note 2 3 4 5 6 7 8 Containment and RCS Activities notes:

1. The note numbers correspond to column numbers.

2. Core Activity-At shutdown.

3. Margin Factor - Accounts for cycle variations.

E10 - 6 0 to NL-16-0388 Control Rod Ejection Accident Analysis

4. Core Release - Applies to melted fuel rods.

5. Gap Fraction - Applies to fuel rods with cladding failure.

6. Fuel Melt Activity - Product of Activity (Ci) in Column 2, margin in Column 3, core release in Column 4 and fuel melt fraction *of 0.0025.

7. Gap Activity - Product of Activity (Ci) in Column 2, margin in Column 3, gap fraction in Column 5, fuel cladding failure fraction of 0, and RPF of 1.7.

8. Total Activity- This is the sum of Columns 6 and 7.

Initial Iodine Activities in the Secondary Coolant The initial iodine activities in the secondary coolant corresponding to 0.1 µCi/g DEi are shown in the following table. These are entered in LocaDose as initial activities within nodes.

Table 3 - Initial Iodine Activities in the RCS and Secondary Coolant Isotope RCS Inhalation Concentration x 0.1 µCi/g DEi Activity in Concentration TEDE DCF DCF Concentration Secondary (1% Leaking (Rem/Ci)

(µCi/g) (Ci)

Fuel) (µCi/g) 1-131 2.5 3.29E+04 8.23E+04 7.56E-02 1.7E+01 1-132 0.9 3.81E+02 3.43E+02 2.72E-02 6.2E+OO 1-133 4.0 5.85E+03 2.34E+04 1.21 E-01 2.8E+01 1-134 0.6 1.31 E+02 7.86E+01 1.81 E-02 4.1E+OO 1-135 2.2 1.23E+03 2.71E+03 6.65E-02 1.5E+01 Total 1.09E+05 3.09E-01 7.1 E+01 Note 2 3 4 5 6 Initial Iodine Activities Notes:

1. The note numbers correspond to column numbers.

2. RCS at 1% Leaking Fuel - These are the RCS concentrations (µCi/g) corresponding to 1% failed (leaking) fuel.

3. DCF - Inhalation TEDE DCFs are from FGR 11, multiplied by 3.17E12 to convert from Sv/Bq to Rem/Ci

4. OCR Weighted Concentration - the concentrations in Column 2 are multiplied by the DCFs in Column 3. The total for this column represents the relative dose corresponding to 1% failed (leaking) fuel.

5. Concentrations at 0.1 µCi/g DEi - the relative dose corresponding to 0.1 µCi/g DEi is (0.1 µCi/g)(3.29E4 rem/ci) = 3.29E3. In Column 5, the concentrations in Column 2 are multiplied by 3.29E3/1.09E5 to obtain the distribution corresponding to 0.1 µCi/g DEi.

6. Activity in Secondary - the concentrations (µCi/g) in Column 5 are multiplied by the mass of 2.29E8 grams and by 1.0E-06 Ci/µCi.

E10 - 7 0 to NL-16-0388 Control Rod Ejection Accident Analysis Alkali Metals in the Secondary System The initial concentrations are assumed to be 20% of the RCS initial activities.

Table 4 - Alkali Metals in RCS (1 % failed (leaking) fuel) and Secondary Isotope Concentration (µCi/q) Activity in RCS 1% Leak Fuel Seconda---"---f------"---'----'-------1 Secondary (Curies)

Rb-88 3.8E+OO 7.6E - - f - - - - - -1.7E+02

1 Rb-89 1.0E-01 2.0E - - - - f - - - - -4.6E+OO

1 Cs-134 2.6E-01 5.2E - - f - - - - - -1.2E+01

1 Cs-136 1.5E-01 3.0E - - f - - - - - -6.9E+OO

1 Cs-137 1.3E+OO 2.6E - - - - - 5.9E+01 Cs-138 9.6E-01 1.9E-01~~~..,.__~~~~~~--+

4.4E+01 Total 6.6E+OO 1.3E+OO ~~~..,._~~~~~~---i 3.0E+02 Note 2 3 4 Initial Alkali Metals in the Secondary System Notes:

1. The note numbers correspond to the columns of the table.

2. RCS Concentrations - the initial RCS concentrations are those corresponding to 1%

failed (leaking) fuel under normal operations.

3. Secondary concentrations - the concentrations in the secondary are the RCS concentrations (corresponding to 0.5 µCi/g DEi) multiplied by 0.2 to achieve concentrations corresponding to 0.1 µCi/g DEi.

4. Secondary Activities - the concentrations in column 3 are multiplied by the mass of 2.29E8 grams and by 1.0E-06 Ci/µCi Radioiodine Appearance Rates Iodine Appearance Rates in Intact SGs from Feedwater A mass flow rate from the Feedwater to the steam generators is generated to develop an appearance rate of iodine into the SG (for steaming to the environment). The mass released (426E5 lbm) is adjusted to add a 10% margin, converted to grams, and divided by the release time to create a mass flow rate:

[(4.26E5 lbm)(1.1 )(453.6 g/lbm)]/(98 sec)(3600 sec/hr)]= 7.81 E9 grams/hour The flow of feedwater into the intact SGs is modeled as an activity production term in LocaDose. The concentrations corresponding to 0.1 µCi/g from Table 4, Column 3 are multiplied by 1.0E-6 Ci/µCi and by time-dependent flow rates (g/hr) from Table 2, yielding the following appearance rates. These are entered in LocaDose as production terms.

E10 - 8 O to NL-16-0388 Control Rod Ejection Accident Analysis Table 5 - Iodine Appearance Rates in Intact SGs Isotope Feedwater Appearance Rate (Ci/hr) 1-131 5.9E+02 1-132 2.1E+02 1-133 9.4E+02 1-134 1.4E+02 1-135 5.2E+02 Control Room Ventilation Parameters Parameter Value Pressurization Mode starts Initiated at start of accident.

CR Make-up Flow Rate 375 cfm (throughout accident)

Pressurization Unfiltered In-leakage 300 cfm (throughout accident)

CR Ingress/egress 10 cfm (throughout accident through CR Vent)

CR Volume 114,000 cubic feet CR Pressurization Filters 98.5% for all radionuclide groups except noble gases CR Recirculation Flow 2700 cfm (throughout accident)

Iodine Filter Efficiency 98.5% for particulates 94.5% for all other radionuclide groups except noble gases CR Breathing Rates 3.5E-04 m 3/sec for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months EAB & LPZ Breathing Rates 3.5E-04 m3/sec for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Atmospheric Dispersion Factors:

Table 6 - Atmospheric Dispersion Factors X/Q (sec/m 3 )

EAB LPZ Control Time (hr) Room 0-2 7.60E-4 2.80E-4 1.66E-3 2-8 - 1.1 OE-4 1.38E-3 8-24 - 1.0E-05 7.20E-04 24-96 - 5.4E-06 5.6E-04 96-720 - 2.9E-06 4.21 E-04 E10 - 9 1 to NL-16-0388 Locked Rotor Accident Analysis Joseph M. Farley Nuclear Plant - Units 1 and 2 Alternative Source Term License Amendment Request Enclosure 11 Locked Rotor Accident Analysis E11 - 1 1 to NL-16-0388 Locked Rotor Accident Analysis LOCKED ROTOR ACCIDENT DOSE CONSEQUENCES USING AST METHODS Licensee Document Number: SM-1080538201-006, Version 1 Method/Computer Program Used: LocaDose Version 7.1 Regulatory Guidance: RG-1.183, including Appendix G

Model Discussion The calculation was performed to address a Locked Rotor Accident (LAA). The scenario for the LAA is that a reactor coolant pump rotor is postulated to seize, leading to reduced coolant flow and reactor trip. The transient causes fuel damage, resulting in increased radioactivity in the Reactor Coolant System (RCS). Activity in the steam generators (SG) due to primary-to-secondary leakage is released to the environment via steaming until cold shutdown conditions are established in the RCS.

Results and Acceptance Limits:

Location Dose (Rem TEDE)

Calculated Limit EAB 1.2 2.5 LPZ 0.83 2.5 Control Room <5 5 Note that the control room dose for this accident was not reported in the FSAR per the current licensing basis. Control room doses for this accident using the AST methods are analyzed as being less than the 5 Rem TEDE limit. However, a reassessment is being performed assuming a delayed manual CREFS initiation. The results of this reassessment are expected to remain less than 5 rem TEDE and be non-limiting.

The maximum 2-hour EAB dose occurs between 6 and 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

(Note that rounding is applied to all values)

Key Assumptions and Inputs Physical Parameters Parameter Value Reactor Power Level: 2775 MWt {+2% uncertainty= 2831 MWt)

RCS Mass: 440,900 lbm {2.00E8 grams)

RCS Volume: 1.02E4 cubic feet SG Mass: 1.68E5 lbm (per SG, which is assumed to

be full)

SG Volume: 2690 cubic feet (each)

Secondary System Margin 10% increase is added to mass flows Coolant Densities: Primary and Secondary water at 62.4 lbm/ft3 E11 - 2 1 to NL-16-0388 Locked Rotor Accident Analysis Partition Factors: Iodine PF = 100 Alkali Metals PF = 1000 (moisture carryover

= 0.1%)

Noble Gases PF = 1 Primary to Secondary Leakage: 1 gpm total.

Table 1: Flow Rates Before 10% Margin Adjustment Pathway Time Release (lbm) Flow Note From To RCS to SG 0 8 - 1.34E-01 CFM 1 Feedwater to 0 2 7.63E+05 1. 73E+08 g/hr 2 SG 2 8 9.29E+05 7.03E+07 g/hr SG to 0 2 5.64E+05 7.53E+01 cfm 3 Environment 2 8 9.17E+05 4.08E+01 cfm Flow Rate Notes:

1. RCS - Volumetric leakage (gallons/minute) from the RCS is divided by 7.48 gal/ft3.

2. Feedwater - The Feedwater flow to the SGs is 693,629 lbm in the first two hours and 844,963 lbm from 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Mass release from the feedwater to the SG is then increased by 10% for margin. Flow is the release (lbm) multiplied by 453.6 grams/lbm and divided by the time duration (hour).

3. SG - Mass release from the SG is 512,325 lbm in the first two hours and 833221 lbm from 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The mass release from the SG is increased by 10% for margin. The flow is then the release (lbm) divided by 62.4 lbm/ft3 and divided by the time duration (min).

Radioactivity Considerations:

20% of the fuel rods experience cladding failure. A radial power peaking factor of 1.7 is applied to the damaged rods.

The fractions of fission product inventory contained within the fuel rod gaps are:

0 1-131 0.08 o Kr-85 0.10 o Other Halogens and Noble Gases 0.05 o Alkali Metals 0.12

Core Fission product inventories are taken from an equilibrium cycle based upon a power level of 2831 MWt. To account for potential cycle-to-cycle variations, the following margin factors are applied to the core inventory:

o Kr-85 1.15 o Xe-133 1.05 o Cs-134 1.35 o Cs-136 1.25 o Cs-137 1.20 E11 - 3 1 to NL-16-0388 Locked Rotor Accident Analysis o Iodine isotopes and other noble gases 1 .02 o Other Isotopes 1.03

100% of the activity released from the core due to cladding failure is instantaneously mixed within the RCS at the onset of the accident. Compared to the gap release, any RCS iodine activity due to spiking is negligible.

Chemical form of iodine released to the environment is 97% elemental, and 3% organic.

The removal mechanism for this pathway is the same for all chemical forms of iodine.

Radial peaking factor for rods with cladding damage is assumed to be 1.7.

RCS activity includes an assumption of normal operations 1% failed (leaking) fuel in accordance with past CLB (Affects alkali metals)

The initial radioiodine concentration in the secondary system is assumed to be at the Technical Specification limit of 0.1 µCi/gm DEi

The initial concentrations of Alkali Metals in Secondary are based upon a ration of the concentration in the RCS: Given 0.1 µCi/gm DEi in the secondary and 0.5 µCi/gm in the RCS, the concentrations of alkali metals in the secondary are assumed to be 20% of those in the RCS.

Containment and RCS Activities Table 2 - Containment and RCS Activities Core Core Activity Margin Gap Gap Activity Margin Gap Gap Isotope (curies) Factor fraction release Isotope (curies) Factor fraction release 1-131 7.50E+07 1.02 0.08 2.1E+06 Br-83 9.70E+06 1.03 0.12 4.1E+05 1-132 1.10E+08 1.02 0.05 1.9E+06 Br-84 1.70E+07 1.03 0.12 7.1E+05 1-133 1.60E+08 1.02 0.05 2.8E+06 Br-85 2.10E+07 1.03 0.12 8.8E+05 1-134 1.70E+08 1.02 0.05 2.9E+06 Br-86 1.50E+07 1.03 0.12 6.3E+05 1-135 1.50E+08 1.02 0.05 2.6E+06 Br-87 3.40E+07 1.03 0.12 1.4E+06 Kr-83m 9.70E+06 1.02 0.05 1.7E+05 Br-88 3.60E+07 1.03 0.12 1.5E+06 Kr-85m 2.10E+07 1.02 0.05 3.6E+05 Rb-86 1.40E+05 1.03 0.12 5.9E+03 Kr-85 7.20E+05 1.15 0.10 2.8E+04 Rb-88 5.70E+07 1.03 0.12 2.4E+06 Kr-87 4.00E+07 1.02 0.05 6.9E+05 Rb-89 7.40E+07 1.03 0.12 3.1E+06 Kr-88 5.70E+07 1.02 0.05 9.9E+05 Rb-90 7.20E+07 1.03 0.12 3.0E+06 Kr-89 6.90E+07 1.02 0.05 1.2E+06 Rb-91 8.90E+07 1.03 0.12 3.7E+06 E11 - 4 1 to NL-16-0388 Locked Rotor Accident Analysis Core Core Activity Margin Gap Gap Activity Margin Gap Gap Isotope (curies) Factor fraction release Isotope (curies) Factor fraction release Xe- Cs-131m 8.40E+05 1.02 0.05 1.5E+04 134m 3.60E+06 1.03 0.12 1.5E+05 Xe-133m 4.80E+06 1.02 0.05 8.3E+04 Cs-134 1.10E+07 1.35 0.12 6.1E+05 Xe-133 1.50E+08 1.05 0.05 2.7E+06 Cs-136 3.30E+06 1.25 0.12 1.7E+05 Xe-135m 3.00E+07 1.02 0.05 5.2E+05 Cs-137 7.60E+06 1.20 0.12 3.7E+05 Xe-135 3.50E+07 1.02 0.05 6.1E+05 Cs-138 1.40E+08 1.03 0.12 5.9E+06 Xe-137 1.40E+08 1.02 0.05 2.4E+06 Cs-139 1.40E+08 1.03 0.12 5.9E+06 Xe-138 1.30E+08 1.02 0.05 2.3E+06 Cs-140 1.20E+08 1.03 0.12 5.0E+06 Br-82 3.80E+05 1.03 0.12 1.6E+04 Cs-141 9.10E+07 1.03 0.12 3.8E+06 Note: 2 3 4 5 Note: 2 3 4 5 Containment and RCS Activities Notes:

1. The note numbers the column numbers and the columns are repeated.

2. Core Activity - Activity in the core for the isotope at shutdown.

3. Margin Factor - per isotope accounting for cycle variations

4. Gap Release Fraction per RG-1.183

5. The released activity is the product of the core activity multiplied by the margin factor multiplied by the release fraction multiplied by the amount of damaged fuel (20%)

multiplied by the radial power peaking factor (1.7).

Initial Iodine Activities in the Secondary Coolant The initial iodine activities in the secondary coolant corresponding to 0.1 µCi/g DEi are shown in the following table. These are entered in LocaDose as initial activities within nodes.

Table 3 - Initial Iodine Activities in the RCS and Secondary Coolant Isotope RCS Concentration Inhalation Concentration 0.1 µCi/g DEi Activity in (1 % Leaking Fuel) TEDE DCF x DCF Concentration Secondary

(µCi/g) (Rem/Ci) (µCi/g) (Ci) 1-131 2.5 3.29E+04 8.23E+04 7.56E-02 1.7E+01 1-132 0.9 3.81E+02 3.43E+02 2.72E-02 6.2E+OO 1-133 4.0 5.85E+03 2.34E+04 1.21 E-01 2.8E+01 E11 - 5 1 to NL-16-0388 Locked Rotor Accident Analysis Isotope RCS Concentration Inhalation Concentration 0.1 µCi/g DEi Activity in (1 % Leaking Fuel) TEDE DCF x DCF Concentration Secondary

(µCi/g) (Rem/Ci) (µCi/g) (Ci) 1-134 0.6 1.31 E+02 7.86E+01 1.81 E-02 4.1 E+OO 1-135 2.2 1.23E+03 2.71E+03 6.65E-02 1.5E+01 Total 1.09E+05 3.09E-01 7.1 E+01 Note: 2 3 4 5 6 Initial Iodine Activities Notes:

1. The note numbers correspond to column numbers.

2. RCS at 1% Leaking Fuel - These are the RCS concentrations (µCi/g) corresponding to 1% failed (leaking) fuel during normal operations.

3. DCF - Inhalation TEDE DCFs are from FGR 11, multiplied by 3.17E12 to convert from Sv/Bq to Rem/Ci

4. OCR Weighted Concentration - the concentrations in Column 2 are multiplied by the DCFs in Column 3. The total for this column represents the relative dose corresponding to 1% failed (leaking) fuel.

5. Concentrations at 0.1 µCi/g DEi - the relative dose corresponding to 0.1 µCi/g DEi is (0.1 µCi/g)(3.29E4 rem/ci) = 3.29E3. In Column 5, the concentrations in Column 2 are multiplied by 3.29E3/1.09E5 to obtain the distribution corresponding to 0.1 µCi/g DEi.

6. Activity in Secondary- the concentrations (µCi/g) in Column 5 are multiplied by the mass of 2.29E8 grams and by 1.0E-06 Ci/µCi.

Alkali Metals in the Secondary System The initial concentrations are assumed to be 20% of the RCS initial activities.

Table 4 - Alkali Metals in RCS (1 % failed (leaking) fuel) and Secondary Isotope Concentration (µ.Ci/q) Activity in Secondary RCS 1% Leak Fuel Secondary (Curies)

Rb-88 3.8E+OO 7.6E-01 1.7E+02 Rb-89 1.0E-01 2.0E-02 4.6E+OO Cs-134 2.6E-01 5.2E-02 1.2E+01 Cs-136 1.5E-01 3.0E-02 6.9E+OO Cs-137 1.3E+OO 2.6E-01 5.9E+01 Cs-138 9.6E-01 1.9E-01 4.4E+01 Total 6.6E+OO 1.3E+OO 3.0E+02 Note: 2 3 4 Initial Alkali Metals in the Secondary System Notes:

1. The note numbers correspond to the columns of the table.

2. RCS Concentrations - the initial RCS concentrations are those corresponding to 1%

failed (leaking) fuel under normal operations.

E11 - 6 1 to NL-16-0388 Locked Rotor Accident Analysis

3. Secondary concentrations - the concentrations in the secondary are the RCS concentrations (corresponding to 0.5 µCi/g DEi) multiplied by 0.2 to achieve concentrations corresponding to 0.1 µCi/g DEi.

4. Secondary Activities - the concentrations in column 3 are multiplied by the mass of 2.29E8 grams and by 1.0E-06 Ci/µCi Radioiodine Appearance Rates Iodine Appearance Rates in Intact SGs from Feedwater The flow of feedwater into the intact SGs is modeled as an activity production term in LocaDose.

The concentrations corresponding to 0.1 µCi/g from Table 3, Column 5 are multiplied by 1.0E-6 Ci/µCi and by time-dependent flow rates (g/hr.) from Table 1, yielding the following appearance rates. These are entered in LocaDose as production terms.

Table 5 - Iodine Appearance Rates in Intact SGs Feedwater Appearance Rate (Ci/hr)

Isotope 0-2 Hour 0-8 hour 1-131 1.3E+01 5.3E+OO 1-132 4.7E+OO 1.9E+OO 1-133 2.1 E+01 8.5E+OO 1-134 3.1E+OO 1.3E+OO 1-135 1.2E+01 4.7E+OO Control Room Ventilation Parameters Parameter Value Isolation Mode Not modeled Pressurization Mode starts Initiated at start of accident (under reassessment for delayed manual start)

Normal CR Make-up Flow Rate 2340 CFM (+10 CFM ingress/egress) unfiltered until pressurization mode is started '

CR Pressurization Flow Rate 375 cfm Pressurization Unfiltered In-leakage 300 cfm CR Ingress/egress 1o cfm (continuous unfiltered through CR Vent)

CR Volume 114,000 cubic feet CR Pressurization Filters 98.5% for all radioactive groups except noble gases CR Recirculation Flow 2700 cfm (manual start at 20 minutes until end of accident)

Iodine Filter Efficiency 98.5% for particulates 94.5% for all other radionuclide groups CR Breathing Rates 3.5E-04 m3/sec for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months EAB & LPZ Breathing Rates 3.5E-04 m3/sec for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months E11 - 7 1 to NL-16-0388 Locked Rotor Accident Analysis Atmospheric Dispersion Factors Table 6 - Atmospheric Dispersion Factors XIQ (sec/m 3)

EAB LPZ Control Time (hr) I Room 0-2 7.60E-4 2.80E-4 1.66E-3 2-8 - 1.10E-4 1.38E-3 E11 - 8

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-16-0388 FNPAST Accident Analysis Input Values Comparison Tables 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 (AST) at Farley Nuclear Plant (FNP), 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) Accident Table 5- Steam Generator Tube Rupture (SGTR) Accident Table 6- Control Rod Ejection Accident Table 7- Locked Rotor Accident Additionally, Table 1, "Control Room Parameters," is provided to show the parameters of interest for Control Room habitability. In this table, the LOCA parameters are provided as they resulted in the most limiting dose to the Control Room occupants.

Table 1: Control Room Parameters Input/Assumption CLB Value New AST Value Reason for Change 3 3 Control Room Volume 114,000 ft 114,000 ft No change Normal Operation Filtered Make-up Flow 0 cfm Ocfm No change Rate Filtered Recirculation Ocfm Ocfm No change Flow Rate Unfiltered Make-up Flow Ocfm 2340 cfm 60 seconds of normal Control Room HVAC Rate operation is assumed after accident initiation.

Unfiltered In-leakage Ocfm Ocfm No change Emergency Operation Recirculation Mode Filtered Make-up Flow 375 cfm 375 cfm No change Rate Filtered Recirculation 2700 cfm 2700 cfm No change Flow Rate Unfiltered Make-up Flow Ocfm Ocfm No change Rate Unfiltered In-leakage 53 cfm 325 cfm The revised value is intended to provide operational margin to the CR measured CR in-leakage. Includes 10 cfm for CR ingress/egress.

Filter Efficiencies Pressurization Filters All iodine 98.5% All iodine 98.5% No change Recirculation Filters Elemental - 94.5% Elemental - 94.5% No change Organic - 94.5% Organic- 94.5%

Particulate - 98.5% Particulate - 98.5%

Particulate 98.5% 98.5% No change Occupancy 0-24 hours 100% 100% No change 1-4 days 60% 60%

4-30 days 40% 40%

3 3 Breathing Rate 3.47E-4 m /sec 3.5E-4 m /sec Rounded up for conservatism.

E12 - 2 2 to NL-16-0388 FNP AST Accident Analysis Input Values Comparison Tables Table 2: LOCA Inputs and Assumptions Input/Assumption CLB Value For ,Offsit!! and N!!W AST Value For Offsite Reason for Change Control Room and Control Room Containment Purge Iodine Chemical Form 2.5% particulate, 95.5% 95% cesium iodide, 4.85% Adoption of RG 1.183 methodology.

elemental, 2.0% organic elemental, 0.15% organic 3 3 Containment Volume 2,030,000 ft 2,030,000 ft No change Containment Purge 0% 0% No change Filtration Removal by Wall None None No change Deposition Removal by Sprays None None No change Containment Leakage Iodine Chemical Form 2.5% particulate, 95.5% 95% cesium iodide, 4.85% Adoption of RG 1.183 methodology.

elemental, 2.0% organic elemental, 0.15% organic Containment Sump pH >7.0 >7.0 No change 3 3 Containment Sprayed 1, 668, 660 ft 1,668,660 ft No change Volume 3 3 Containment unsprayed 361,340 ft 361,340 ft No change Volume Containment Spray O seconds 90 seconds Provides additional conservatism to Start Time Containment Leakage Pathway.

Containment Spray Stop 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months No change Time Containment Spray 2480 gpm injection mode 2480 gpm injection mode No change Flow Rate 2290 gpm recirculation mode Elemental Iodine Spray 2.7 hr" 1

13.7 hr*1 Revision is consistent with RG 1.183 Appendix A Removal Coefficient RP 3.3 1 1 Aerosol Spray Removal 5.45 hr- injection mode 5.45 hr" injection mode No change 1 1 Coefficient 5.03 hr" recirculation mode 5.03 hr- recirculation mode Organic Iodine Spray None None No change Removal Natural Deposition Elemental, Organic, Aerosol - Elemental, Organic iodine - Aerosol natural deposition is permitted per None None Appendix A of RG 1.183.

1 Aerosols - 0.1 hr" in unsprayed regions only Containment Leakage No change Rate 0.15%/day 0.15%/day 0 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.075%/day 0.075%/day 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 30 days Containment Leakage 0% 0% No change Filtration ECCS Leakage to the Auxiliary Building Iodine Che'mical Form 0% aerosol, 98% elemental, 0% aerosol, 97% elemental, The revised percentages are as specified in RG 2% organic 3% organic 1.183.

3 3 Containment Sump 49,200 ft 49,200 ft No change Volume ECCS Recirculation Start 20 minutes 20 minutes No change Time 20 minutes ECCS Leakage Flow Rate 4,000 cc/hr 20, 000 cc/hr ECCS flow rate increased to provide additional operating margin in the analysis.

ECCS Flashing Fraction 15% 10% ECCS flashing fraction recalculated based on Section 5.4 of RG 1.183.

ECCS Leakage to the RWST (Not Explicitly Modeled in the CLB)

E12 - 3 2 to NL-16-0388 FNP AST Accident Analysis Input Values Comparison Tables Table 3: FHAlnputs and Assum ltions Input/Assumption C:LB Value For Offsite and. .New AST Value For Offsite Reason for Change Control Room and Control Room Iodine Chemical Form 0% aerosol, 99.75% elemental, 0% aerosol, 99.85% Chemical composition is as described in RG 1.183 0.25% organic elemental, 0.15% organic Appendix B Section 2.

Number of Fuel 1 1 No change Assemblies Damaged Percentage of Fuel 100% 100% No change Rods Damaged No. of rods exceeding 0 0 No change 6.3 kw/ft above 54 GWD/MTU Water Level Above 23 ft 23 ft No change Damaged Fuel Pool Decontamination Elementary- 400 Elementary- 500 Decontamination Factors are as described in RG Factors Organic - 1 Organic - 1 1.183 Appendix B Section 2.

Delay Before Fuel 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months No change Movement Containment Release 0% 0% No change Filtration Table 4: MSLB Accident Inputs and Assumptions Input/Assumption CLB Value For Offsite and New AST Value For Offsite Reason for Change Control Room and Control Room Maximum Pre-Accident 30 µCi/gm Dose Equivalent 1-131 30 µCi/gm Dose Equivalent No change Iodine Spike 1-131 Concentration Concurrent Iodine 500 X Equilibrium 500 X Equilibrium No change Spike Appearance Rate Initial Steam Generator 0.1 µCi/gf Dose Equivalent 0.1 µCi/gf Dose Equivalent No change Iodine Source Term Iodine Iodine Iodine Chemical Form Not provided 0% aerosol, 97% elemental, The AST chemical form is as provided in RG 1.183 3% organic Appendix E, Section 4.

Percentage of Fuel 0% 1% No change. Note- the MSLB does not result in Rods Failed failed fuel. This is a leaking fuel pre-condition included for conservatism.

RCS Mass 440,000 lbm 440,900 lbm AST calculation more precisely accounts for CVCS mass.

Steam Generator 168,000 lbm/SG 168,000 lbm/SG No change Secondary Liquid Mass Intact Steam Generator 0- 2 hrs: 316, 715 lbm 0 - 2 hrs: 316, 715 lbm No change Steam Release 2 - 8 hrs: 703,689 lbm 2- 8 hrs: 703,689 lbm 8- 24 hrs: 948,000 lbm 8 - 24 hrs: 948,000 lbm Primary-Secondary 0.65 gpm to two intact SGs 0.65 gpm to two intact SGs No change Leak Rate 0.35 gpm to faulted SG 0.35 gpm to faulted SG 3 3 Density Used for 62.4 lbm/ft 62.4 lbm/ft No change Leakage Volume-to-Mass Conversion Duration of Intact SG Not modeled Not modeled Tube uncover does not occur with intact SGs Tube Uncovery After Reactor Trip Time to Cool RCS to 24 hrs 24 hrs No change 200°F Intact Steam Generator 10 100 RG 1.183 Appendix E Section 5.5.4 allows an Iodine partition factor iodine partition factor of 100 for the intact SG.

Intact Steam Generator Not modeled 0.1% (Alkali Metal Partition New AST value conservatively included.

Moisture Carryover Factor =1000)

Fraction E12 - 4 2 to NL-16-0388 FNP AST Accident Analysis Input Values Comparison Tables Table 5: SGTR.Accident Inputs and Assu.mptions "

Input/Assu~ption CLBVa!ue For Offsite 'and New AST Value For Offsite *~ '. Reasori for Change Control Room and Control Room < < *'

Maximum Pre-Accident 30 µCi/gm Dose Equivalent 1-131 30 µCi/gm Dose Equivalent No change Iodine Spike 1-131 Concentration Concurrent Iodine 500 X Equilibrium 335 X Equilibrium RG 1.183 Appendix F Section 2.2 allows the 335 Spike Appearance Rate factor.

Initial Steam Generator 0.1 µCi/gm Dose Equivalent I- 0.1 µCi/gm Dose No change Iodine Source Term 131 Equivalent 1-131 Iodine Chemical Form Not provided 0% aerosol, 97% Iodine chemical form is per RG 1.183 Appendix F elemental, 3% organic Section 4.

Percentage of Fuel 0% 1% No change. Note- the SGTR does not result in Rods Failed failed fuel. This is a leaking fuel pre-condition included for conservatism.

RCS Mass 441,000 lbm 440,900 lbm AST calculation more precisely accounts for eves mass.

Steam Generator 105,000 lbm/SG 105,000 lbm/SG No change Secondary Liquid Mass Intact Steam Generator 0-0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months -1133 O- 0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months -1133 No change Steam Release lbm/second lbm/second 0.09 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months - 422,000 lbm 0.09 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months - 422,000 2- 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months - 934,000 lbm lbm 2- 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months - 934,000 lbm Ruptured Steam 0-0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months -1133 0- 0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months -1133 No change Generator Steam lbm/second lbm/second Release 0.09 -0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months - 79,000 lbm 0.09 - 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months - 79,000 lbm Feedwater Flow to 0.09 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months - 327,000 lbm 0.09 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months - 327,000 No change Intact Steam 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months - 981,000 lbm lbm Generators 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months - 981,000 lbm 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 No change Primary-Secondary lgpm lgpm No change Leak Rate 3 0 Density Used for 62.4 lbm/ft 62.4 lbm/ft No change Leakage Volume-to-Mass Conversion Ruptured Tube Break 0 - 0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months - 21,600 lbm 0 - 0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months - 21,600 No change Flow 0.09 - 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months - 79,000 lbm lbm 0.09- 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months - 79,000 lbm Duration of Ruptured 30 minutes 30 minutes No change Tube Break Flow Break Flow Flashing 0-0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months - 21% 0 - 0.09 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months - 21% No change Fraction 0.09 -0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months -15% 0.09 - 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months -15%

Duration of Intact SG 0 minutes O minutes No change Tube Uncovery After Reactor Trip Time to Cool RCS to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months No change 200°F Intact Steam Generator 100 100 No change Iodine Partition Coefficient Intact Steam Generator Not provided 0.1% Carryover is provided for per RG 1.183 Appendix Moisture Carryover F Section 5.6.

Fraction E12 - 5 2 to NL-16-0388 FNP AST Accident Analysis Input Values Comparison Tables

), i'

, Table s::cont~ol R"Qd Ejecticm* Acc~ctent]nput$ a)'fdAS:sumptioris* ,** i.,

,)nput/Aefsurijpti~n . New.1STVahJ,~ FC>_r,Offslti!

CLB ValueForOffSite and . * **,*Reason' for Change ..

""'" <- c'*'-""

'. *' CQn.trcii Room ,, +.* ..** * * *and,(:ontrol Room *

'.~.;: ,: < .,,,., ,>> / .... ,ii Fuel Rod Gap Fractions lodine-12% Iodine isotopes and noble AST gap fractions are per RG 1.183 Table 3 Krss-10% gases-0.10 Other halogens - 0.05 Alkali metals -0.12 Fuel Rod Peaking Factor Not provided 1.7 Radial peaking factor is applied per RG 1.183 Section 3.1.

Percentage of Fuel Rods 10% 10% No change Damaged Percentage of Fuel That 0.25% 0.25% No change Experiences Melting Number of rods 0 0 No change exceeding 6.3 kw/ft above 54 GWD/MTU Initial Steam Generator 0.1 µCi/gm 0.1 µCi/gm No change Iodine Source Term Iodine Chemical Form - Not Provided 95% aerosol, 4.85% Although different from RG 1.183, Appendix H Secondary Release elemental, 0.15% organic Section 5, this is acceptable because the removal mechanism for all chemical forms of iodine is the same for this pathway Iodine Chemical Form - Not provided 95% aerosol, 4.85% Iodine chemical form is in accordance with RG Containment Release elemental, 0.15% organic 1.183 Appendix H Section 4 3 3 Containment Volume 2.03E6 ft 2.03E6 ft No change Containment Leakage No change Rate 0.15% 0.15%

0 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.075% 0.075%

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 30 days Containment Leakage 0% 0% No change Filtration Natural Deposition in 50% plateout of RCS release Elemental iodine - None Natural deposition is credited per RG 1.183 Containment Aerosols - 2.74E-2 hr" Appendix H Section 6.1.

1 Iodine/Particulate Not provided 5.0 hr" Particulate removal by Containment Spray is per Removal by RG 1.183 Appendix H Section 6.1.

Containment Sprays RCS Mass 435,000 lbm 440,900 lbm The AST calculation reflects the liquid mass resulting from SG replacement.

Steam Generator 168,000 lbm 168,000 lbm No change Secondary Liquid Mass Primary-Secondary Leak 150 gallons per day per SG 1 gpm total AST value increased for additional conservatism.

Rate 3 3 Density Used for 62.4 lbm/ft 62.4 lbm/ft No change Leakage Volume-to-Mass Conversion Secondary Steam 426,000 lbm 426,000 lbm No change Release Time until Primary and 2500 seconds 2500 seconds No change Secondary Pressures Equalize Duration of SG Tube O minutes O minutes No change Uncovery Following Reactor Trip Steam Generator Iodine 10 100 RG 1.183 Appendix H Section 7.4 allows an Partition Coefficient iodine partition factor of 100 for SG releases.

Steam Generator Not provided 0.1% Carryover is provided for per RG 1.183 Appendix Moisture Carryover H Section 7.4.

Fraction E12 - 6 2 to NL-16-0388 FNP AST Accident Analysis Input Values Comparison Tables

> Input/Assumption -~LB.Vallie Fo~ ~ff:site and ', Nev/.i:'STValue For Offsitli" ::,, . Reason for Change . A

'*-:::>-" "-~ . *,;,""'~ , .,-,, *,controlRqom ,_.. , ;,a:nd~tohtr({fRO~()~:>> v,-~*,,.,;'.* ~,~'-, :* *,,.,~~-- ;;~--,* "._,_..:-*~~:

Fuel Rod Gap Fractions Iodine -12% 1-131- 0.08 AST gap fractions are per RG 1.183 Table 3 Kr 0.10 Other Halogens and Noble Gases - 0.05 Alkali Metals - 0.12 Fuel Rod Peaking Factor Not provided Radial peaking factor is applied per RG 1.183 Section 3.1.

Number of rods 0 0 0 exceeding 6.3 kw/ft above 54 GWD/MTU Initial Steam Generator 0.1 µCi/gm 0.1 µCi/gm No change Iodine Source Term Iodine Chemical Form Not provided 95% particulate Iodine chemical form is in accordance with RG 4.85% elemental 1.183 Appendix G Section 5.6.

0.15% organic RCS Mass 441,000 lbm 440,900 lbm AST calculation more precisely accounts for eves mass.

Steam Generator 168,000 lbm 168,000 lbm No change Secondary Liquid Mass Primary-Secondary Leak 150 gallons per day per SG 1 gpm total AST value increased for additional conservatism.

Rate 3 3 Density Used for 62.4 lbm/ft 62.4 lbm/ft No change Leakage Volume-to-Mass Conversion Secondary Steam 426,000 lbm 426,000 lbm No change Release Time until Primary and 2500 seconds 2500 seconds No change Secondary Pressures Equalize Duration of SG Tube 0 minutes 0 minutes No change Uncovery Following Reactor Trip Steam Generator Iodine 10 100 RG 1.183 Appendix G Section 5.6 allows an Partition Coefficient iodine partition factor of 100 for SG releases.

Intact Steam Generator Not provided 0.1% Carryover is provided for per RG 1.183 Appendix Moisture Carryover G Section 5.6.

Fraction E12 - 7 3 to NL-16-0388 FNP AST LAR Supporting Information Joseph M. Farley Nuclear Plant - Units 1 and 2 Alternative Source Term License Amendment Request Enclosure 13 FNP AST LAR Supporting Information

- LOCA RADTRAD Input Files in CD Format

- LOCA RADTRAD Output Files in CD Format

- FHA RADTRAD Input Files in CD Format

- FHA RADTRAD Output Files in CD Format

- ARCON96 Files For RWST X/Qs E13 - 1 4 to NL-16-0388 Summary of Regulatory Commitments Joseph M. Farley Nuclear Plant - Units 1 and 2 Alternative Source Term License Amendment Request Enclosure 14 Summary of Regulatory Commitments E14 - 1 4 to NL-16-0388 Summary of Regulatory Commitments Enclosure 14 Summary of Regulatory Commitments The following table identifies the regulatory commitments in this document. Any other statements in this submittal represent intended or planned actions. They are provided for information purposes and are not considered to be regulatory commitments.

REGULATORY COMMITMENTS DUE DATE/EVENT

1. Administrative controls will be established to ensure Prior to appropriate personnel are aware of the open status of the implementation of the penetration flow path(s) during core alterations or movement LAR of irradiated fuel assemblies within the containment.

2. Existing administrative controls for open containment airlock Prior to doors will be expanded to ensure specified individuals are implementation of the designated and readily available to isolate any open LAR penetration flow path(s) in the event of an FHA inside containment.

3. With the Personnel Airlock open during fuel handling Prior to operations or core alterations, the Containment Purge implementation of the System will be in operation. LAR

4. In the event of an FHA, the containment will be evacuated Prior to and the Personnel Airlock will be closed within 30 minutes of implementation of the detection of the accident. LAR

5. In the event of an FHA, Control Room occupants will use the Prior to secondary door to the Control Room for ingress and egress. implementation of the LAR E14- 2

v t e Letter Sequence Request
CAC:MF8861, (Approved, Closed)
Initiation Request, Request, Request Acceptance Supplement, Supplement Administration Questions 24 March 2017 21 July 2017 Answers 23 May 2017 8 June 2017 Results Approval Other:
MONTHYEAR2017-03-24 [Table View]

NL-16-0388, Alternative Source Term License Amendment Request

Text

Enclosures:

3.0 TECHNICAL EVALUATION

Attachment:

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