ML062490242

Enclosure 1, E. I. Hatch Request to Implement an Alternative Source Term - AST Safety Assessment

ML062490242
Person / Time
Site:	Hatch
Issue date:	08/29/2006
From:	Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML062490242 (110)

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Enclosure 1 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term AST Safety Assessment

Enclosure I Page ii of viii Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment TABLE OF CONTENTS LIST OF TABLES ...................................................................................................................................... iii LIST OF FIGURES ...................................................................................................................................... iv A CRON YM S AN D A BBREVIATIONS ................................................................................................ v I. INTRODUCTION ................................................................................................................................. I 1.1 O verview and Objectives ............................................................................................................... 1 1.2 Sum m ary of Technical Specification Changes .............................................................................. 1 1.3 Changes to Main Control Room Unfiltered Inleakage Assumptions ...................................... 2 1.4 Sum m ary ........................................................................................................................................ 3

2. EVALUATION ..................................................................................................................................... 3 2.1 Changes to Current Licensing Basis ........................................................................................ 3 2.1.1 M SIV Alternate Leakage Treatment ............................................................................ 3 2.1.2 M CR and TSC Inleakage ........................................................................................... 3 2.1.3 Standby Liquid Control System .................................................................................... 3 2.1.4 Turbine Building Ventilation ...................................................................................... 4 2.1.5 Secondary Containment Bypass Leakage ..................................................................... 4 2.1.6 D rywell Sprays ........................................................................................................... 4 2.1.7 Atmospheric D ispersion Factors ................................................................................. 4 2.1.8 Cable Spreading Room Fans ....................................................................................... 4 2.1.9 FHA Decontam ination Factor ...................................................................................... 5 2.1.10 Dose Equivalent 1-131 ................................................................................................. 5 2.1.11 Design Inputs and A ssumptions .................................................................................. 8 2.2 M ethodology .................................................................................................................................. 9 2.2.1 Accident Radiological Consequence A nalysis ............................................................ 9 2.2.2 Suppression Pool pH Control A nalysis .................................................................... 10 2.2.3 NUREG-0737 A nalysis ............................................................................................. 10 2.3 Environmental Q ualification ................................................................................................. 11 2.4 Atmospheric D ispersion Factors ............................................................................................. 11 2.4.1 M eteorological Data ................................................................................................. 11 2.4.2 ARCON96 ...................................................................................................................... 12 2.5 Accident Radiological Consequence A nalyses ....................................................................... 13 2.5.1 Com m on Inputs and A ssumptions ............................................................................ 13 2.5.2 Loss-of-Coolant Accident (LOCA) .......................................................................... 17 2.5.3 Fuel Handling Accident (FHA) ....................................................................................... 32 2.5.4 Control Rod Drop Accident (CRDA) ..................................................................... 36 2.5.5 M ain Steam Line Break (M SLB) ....................................................... .................... 40 2.6 Suppression Pool pH Control ................................................................................................ 45 2.6.1 Inputs and Assumptions ................................................................................................. 45 2.6.2 M ethod of Evaluation ............................................................................................... 46 2.6.3 Results ............................................................................................................................ 47 2.7 Crediting of Non-Safety Related System s ............................................................................ 48 2.7.1 M SIV A lternate Leakage Treatment ........................................................................ 48 2.7.2 Standby Liquid Control System ................................................................................ 53 2.7.3 Turbine Building Ventilation ..................................................................................... 54 2.8 NUREG-0737 Evaluation ................................................ 60 2.8.1 Post-Accident Access Shielding ................................................................................ 61

Enclosure 1 Page iii of viii Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment 2.8.2 Post-Accident Radiation Monitor .............................................................................. 61 2.8.3 Leakage Control ......................................................................................................... 61 2.8.4 Control Room and TSC Radiation Protection .......................................................... 61

3. C ON C LU SIO N S ................................................................................................................................. 62

4. REFE R EN C E S .................................................................................................................................... 62 APPENDIX A: Regulatory Guide 1.183 Conformance Matrix ..................................................... A-1 APPENDIX B: Design Inputs and Assumptions for DBA Analyses ............................................. B-1 LIST OF TABLES Table 1. Proposed Technical Specification Changes ............................................................................. 2 Table 2. Unit I Potential Secondary Containment Bypass Leakage ........................................................ 5 Table 3. Polestar Methods for Activity Removal ................................................................................... II Table 4. Atmospheric Dispersion Factors - Main Control Room X/Q (sec/i 3)..................12 Table 5. Atmospheric Dispersion Factors - Offsite x/Q (sec/m 3) .......................................................... 13 Table 6. Common DBA Radiological Consequence Analyses Inputs and Assumptions ....................... 14 T able 7. Core Inventory ............................................................................................................................... 16 Table 8. Core R elease Rates........................................................................................................................ 27 Table 9. Flow Rates from DW to Torus .................................................................................................. 27 Table 10. DW Activity Removal Rates .................................................................................................. 28 Table 11. DW Pressure Used in Containment Activity Removal Model .............................................. 28 Table 12. DW Temperature Used in Containment Activity Removal Model ........................................ 28 Table 13. DW Steam Mole Fraction Used in Containment Activity Removal Model .......................... 29 Table 14. Main Steam Line Deposition Rates ........................................................................................ 29 Table 15. Condenser Deposition Rates ................................................................................................. 29 Table 16. Containment Leakage Reduction Parameters ......................................................................... 30 Table 17. MSIV Leakage Reduction Parameters .................................................................................. 30 Table 18. Containment and MSIV Leakage Rates ........................................................................... ........... 31 Table 19. Offsite Doses for LOCA ......................................................................................................... 31 Table 20. MCR Doses for LOCA ............................................................................................................ 32 Table 21. TSC Dose for LOCA .............................................................................................................. 32 Table 22. Activity Releases from FHA ................................................................................................. 35 Table 23. Decontamination Factors as a Function of Water Depth ........................................................ 36 Table 24. D oses from FHA ......................................................................................................................... 36 Table 25. Offsite Doses from CRDA .................................................................................................... 39 Table 26. MCR Doses from CRDA ...................................................................................................... 40 Table 27. TSC Doses from CRDA ......................................................................................................... 40 Table 28. Primary Coolant Iodine Activities .......................................................................................... 43 Table 29. Noble Gas Release Rates ....................................................................................................... 44 Table 30. Steam Blowdown Rate ............................................................................................................ 44 Table 31. Offsite Doses from MSLB ................................................................................................... 44 Table 32. MCR Doses from MSLB ....................................................................................................... 45 Table 33. TSC Doses from MSLB ......................................................................................................... 45 Table 34. Design Inputs for pH Calculation .......................................................................................... 47 Table 35. Suppression Pool pH vs. Time ............................................................................................... 48 Table 36. Temporal Distribution of Turbine Building Ventilation Rates .................................................. 60 Page iv of viii Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Table 37. MCR Dose with Passive Ventilation ..................................................................................... 60 LIST OF FIGURES Figure 1. MSIV Alternate Leakage Treatment Pathways ........................................................................ 52 Page v of viii Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment ACRONYMS AND ABBREVIATIONS Ah Head Loss X, Removal Coefficient p Density X/Q Atmospheric Dispersion Factor ALT Alternate Leakage Treatment AST Alternative Source Term ASTM American Society for Testing and Materials Btu British Thermal Unit BWR Boiling Water Reactor BWROG Boiling Water Reactor Owners' Group CAD Containment Atmosphere Dilution CEDE Committed Effective Dose Equivalent cfh Cubic Feet per Hour cfm Cubic Feet per Minute CFR Code of Federal Regulations Ci Curie cm Centimeter Cp Pressure Coefficient CRDA Control Rod Drop Accident CS Core Spray CSB Containment Systems Branch CsI Cesium Iodide CST Condensate Storage Tank DBA Design Basis Accident DBE Design Basis Earthquake DDE Deep Dose Equivalent DE 1-131 Dose Equivalent 1-131 DF Decontamination Factor DNBR Departure from Nucleate Boiling Ratio DW Drywell EAB Exclusion Area Boundary

Enclosure I Page vi of viii Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment ACRONYMS AND ABBREVIATIONS ECCS Emergency Core Cooling System EDE Effective Dose Equivalent EPRI Electric Power Research Institute ESF Engineered Safety Feature F Fahrenheit FGR Federal Guidance Report FHA Fuel Handling Accident FSAR Final Safety Analysis Report ft Feet g Gram GDC General Design Criterion GIP Generic Implementation Procedure GL Generic Letter gpm Gallons per Minute HNP Edwin I. Hatch Nuclear Plant HPCI High-Pressure Coolant Injection HPT High Pressure Turbine hr Hour HVAC. Heating, Ventilation, and Air Conditioning ICRP International Commission on Radiological Protection ILRT Integrated Leak Rate Test in Inch INEEL Idaho National Engineering and Environmental Laboratory kg Kilogram KI Potassium Iodide kV Kilovolt Ibm Pounds (Mass)

LOCA Loss-of-Coolant Accident LOSP Loss of Offsite Power LPZ Low Population Zone m Meter MCC Motor Control Center

Enclosure 1 Page vii of viii Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment ACRONYMS AND ABBREVIATIONS MCR Main Control Room MCREC Main Control Room Environmental Control MeV Million Electron Volts min Minute mm Millimeter MOV Motor Operated Valve mph Miles per Hour MSIV Main Steam Isolation Valve MSL Main Steam Line MSLB Main Steam Line Break MWt Megawatt Thermal NRC Nuclear Regulatory Commission Pa Pascal psia Pounds Per Square Inch Absolute psig Pounds Per Square Inch Gauge PSW Plant Service Water PWR Pressurized Water Reactor Q Volumetric Flow Rate RB Reactor Building RBCCW Reactor Building Closed Cooling Water RCIC Reactor Core Isolation Cooling RCS Reactor Coolant System rem Roentgen Equivalent Man RG Regulatory Guide RHR Residual Heat Removal RWCU Reactor Water Cleanup scfh Standard Cubic Feet per Hour sec Second SER Safety Evaluation Report SGTS Standby Gas Treatment System SLC Standby Liquid Control SNC Southern Nuclear Operating Company Page viii of viii Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment ACRONYMS AND ABBREVIATIONS SPB Sodium Pentaborate (Na 20-5B 20 3-10H 20)

SQUG Seismic Qualification Utility Group SRP Standard Review Plan SSE Safe Shutdown Earthquake TB Turbine Building TEDE Total Effective Dose Equivalent TID Technical Information Document TIP Traversing Incore Probe TS Technical Specification TSC Technical Support Center USI, Unresolved Safety Issue wg Water Gauge

Enclosure 1 Page 1 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment

1. INTRODUCTION 1.1 Overview and Objectives The objective of this safety assessment is to document implementation of the Alternative Source Term (AST) for the Edwin I. Hatch Nuclear Plant (HNP) Units I and 2. The implementation of AST is governed by Title 10 of the Code of Federal Regulations (CFR), Section 50.67, the guidelines of the Standard Review Plan (SRP) Section 15.0.1, and Regulatory Guide (RG) 1.183.

Conformance to the positions of RG 1. 183 is closely adhered to for AST implementation. A RG 1.183 conformance matrix is included as Appendix A to this enclosure, providing the RG 1.183 positions, the corresponding HNP positions, and any clarifying comments. In addition, due consideration has been given to U.S. Nuclear Regulatory Commission (NRC) Regulatory Issue Summary 2006-04, "Experience with Implementation of Alternative Source Terms," March 7, 2006.

Southern Nuclear Operating Company (SNC) has elected to perform a full scope implementation of the AST for HNP Units I and 2 as defined in RG 1.183. The implementation consists of the following:

I. Identification of the core source term based on plant specific analysis of core fission product inventory. All characteristics of the AST are considered, including the composition and magnitude of the radioactive material, its chemical and physical form, and the timing of its release.

2. Determination of the release fractions for the four Final Safety Analysis Report (FSAR) Boiling Water Reactor (BWR) Design Basis Accidents (DBAs) that could potentially result in significant control room and offsite doses. These DBAs are the loss-of-coolant accident (LOCA), the fuel handling accident (FHA), the control rod drop accident (CRDA), and the main steam line break (MSLB).

3. Calculation of fission product deposition rates and removal efficiencies.

4. Calculation of offsite, control room, and technical support center (TSC) personnel total effective dose equivalent (TEDE).

5. Evaluation of suppression pool pH to ensure that the particulate iodine deposited into the suppression pool during a DBA LOCA does not re-evolve and become airborne as elemental iodine.

6. Evaluation of other related design and licensing bases such as NUREG-0737 requirements.

1.2 Summary of Technical Specification Changes The implementation of AST and the radiological dose consequence analyses includes several changes to HNP Technical Specifications (TSs). Table 1 identifies the TS changes proposed. For a more detailed description and justification of these TS changes, see Enclosure 2 of this submittal.

Enclosure I Page 2 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Table 1. Proposed Technical Specification Changes TS Number Scope and Description of Technical Specification Change The definition of DE 1-131 is revised to replace "thyroid dose" with 1.1 Definition of DE 1-131 "Committed Effective Dose Equivalent" and to reference FGR 11 for the dose conversion factors used in calculating 1-131 concentration.

Reactor Coolant Activity System Specific The maximum 4.0 pCi/g allowed DE 1-131 reactor to 2.0 pCi/g coolant specific activity is reduced from DE 1-131.

The maximum allowed bypass leakage rate for all secondary containment Secondary Containment bypass leakage paths is 2.0% of the maximum allowable primary 3.6.1.3 Bypass Leakage containment from 0.9% of leakage the maximum rate. This is a new allowable TS forcontainment primary Unit 1 and an increase leakage rate for Unit 2.

The maximum allowable combined MSIV leakage rates are revised by 3.6.1.3 MSIV Leakage increasing line leakageUnit limit.1 and decreasing In addition, two Unit 2 to 100 separate scfh and eliminating surveillance acceptancethe per criteria will be provided dependent on leakage rate test pressure.

DW Spray A new TS for RHR DW spray is added to reflect the crediting of DW spray 3.6.2.5 as part of the AST LOCA assumptions.

1.3 Changes to Main Control Room Unfiltered Inleakage Assumptions The HNP current licensing basis main control room (MCR) unfiltered inleakage limit is 110 cfm based on the administration of potassium iodide (KI) tablets to MCR occupants within 2 hr after the start of a design basis LOCA. The HNP Units I and 2 common MCR has a unique location. The MCR, as part of the control building, is located between the open end bays of the HNP Units I and 2 turbine buildings (TBs). The majority of the ductwork associated with the main control room environmental control (MCREC) system, which encompasses two independent filter trains for pressurizing the control room post-accident, is located external to the control room boundary on top of the control building within the confines of the HNP Units I and 2 TBs.

By letters dated August 4, 2003, March 29, 2004, October 27, 2004, and November 10, 2005, SNC submitted a course of action for developing responses to NRC Generic Letter (GL) 2003-01, "Control Room Habitability" information requests for HNP. GL 2003-01 was written to inform licensees that the design basis assumptions used for control room unfiltered inleakage, even with a pressurized control room, could be non-conservative. This was validated through testing at several power reactor facilities using the standard test method described in American Society for Testing and Materials (ASTM) consensus standard E74 1, "Standard Test Method for Determining Air Change in a Single Zone by Means of a Tracer Gas Dilution."

In order to address the possibility of unfiltered inleakage into the HNP control room, the incorporation of KI was approved on an interim basis as a measure to limit the thyroid dose to control room occupants in the event of a design basis LOCA. The incorporation of KI in the interim licensing basis is provided to assure that the 30-day thyroid dose remains within the regulatory limits of 10 CFR 50, Appendix A, General Design Criterion (GDC) 19, with MCR unfiltered inleakage up to 110 cfm. As a condition of the licensing basis, the crediting of KI in limiting post-LOCA doses to MCR personnel is for an interim period, expiring on May 31, 2010.

Enclosure 1 Page 3 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Tracer gas testing of the MCR envelope was completed in June 2006 using ASTM consensus standard E741. The most limiting results from testing revealed 5 cfm unfiltered inleakage into the MCR. With the completion of tracer gas testing, SNC will be completing its response to GL 2003-01 under a separate letter.

The change to the licensing basis by the implementation of AST is required to comply with control room habitability regulatory requirements without relying on the KI interim licensing basis. Approval of the AST license amendment request will ensure that the design basis radiological analysis for occupants of the MCR reflects the most limiting unfiltered inleakage into the MCR. There is significant margin between the measured unfiltered inleakage and the unfiltered inleakage assumptions used in the most limiting DBA dose consequence analysis for occupants of the MCR, which is the LOCA. Assumed unfiltered inleakage for the LOCA in the AST dose consequence analysis is 115 cfm.

1.4 Summary Implementation of the AST as the HNP radiological consequence analyses licensing basis requires a license amendment in accordance with the requirements of 10 CFR 50.67. The AST radiological consequence analyses demonstrate that the offsite, MCR, and TSC post-accident radiological doses remain within regulatory limits.

2. EVALUATION 2.1 Changes to Current Licensing Basis Implementation of AST includes several changes to the current licensing basis. These are summarized below.

2.1.1 MSIV Alternate Leakage Treatment RG 1.183, Appendix A, Section 6, allows credit for a reduction in main steam isolation valve (MSIV) releases due to holdup and retention in the main steam line (MSL) piping downstream of the MSIV and in the condenser for a DBA LOCA. This credit is based, in part, on the piping and components of the alternate leakage treatment (ALT) release path being capable of performing their safety functions during and after a design basis earthquake (DBE). The HNP AST implementation credits the ALT pathway for HNP Unit 1.1 Section 2.7.1 describes the ALT application.

2.1.2 MCR and TSC Inleakage The HNP current licensing basis contains an MCR unfiltered inleakage limit of 110 cfm based upon the administration of KI. Implementation of the AST would allow the interim licensing basis crediting KI to be retired while also increasing the MCR design basis unfiltered inleakage limit to 115 cfm. TSC inleakage is also considered in evaluating the dose consequences to occupants of the TSC.

2.1.3 Standby Liquid Control System The standby liquid control (SLC) system is credited for the injection of sufficient sodium pentaborate (SPB) solution to prevent the re-evolution of iodine from the suppression pool for a 30-day period 1NRC has previously approved MSIV ALT for Unit 2 (Reference 1).

Enclosure 1 Page 4 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment following a DBA LOCA. The pH buffering effect of SLC injection is sufficient to offset the effects of acids that are transported to the suppression pool and maintain suppression pool pH at or above 7, thus precluding the re-evolution of elemental iodine. NRC review guidelines, "Guidance on the Assessment of a BWR SLC System for pH Control," were addressed. An evaluation of the SLC system and its ability to perform the post-LOCA injection function is discussed in Section 2.7.2.

2.1.4 Turbine Building Ventilation The TB ventilation system is credited for the removal of activity from the TB beginning 9 hr after the start of a DBA, exhausting at a rate of 15,000 cfm. TB ventilation is credited for the LOCA, CRDA, and MSLB.

2.1.5 Secondary Containment Bypass Leakage The primary containment leakage that bypasses the secondary containment (reactor building) is assumed to be into the condenser for evaluating MCR doses for the DBA LOCA analysis. Activity holdup and deposition in the condenser from this secondary containment bypass leakage is credited in a manner similar to the treatment of MSIV releases and the MSIV ALT pathway.

An evaluation of the Unit 1 secondary containment system was performed using the guidance provided by Branch Technical Position CSB 6-3, "Determination of Bypass Leakage Paths in Dual Containment Plants." All primary containment penetrations were assessed to identify the leakage paths that do not terminate within the secondary containment and should be considered as potential secondary containment bypass leakage paths. Table 2 lists the piping systems identified as potential bypass leakage paths. This evaluation was performed consistent with the current licensing basis for Unit 2.

2.1.6 Drywell Sprays Drywell (DW) sprays are credited to help remove airborne particulates in the DW in the case of the DBA LOCA. DW sprays are also credited in the DBA LOCA analysis for primary containment atmosphere temperature and pressure reduction. This temperature and pressure reduction over time allows primary containment leakage and MSIV leakage to be reduced by 50% at 72 hr after the initiation of the LOCA.

The primary containment pressure and temperature profiles over time for the DBA LOCA were developed by GE, consistent with the current licensing basis containment analysis of the DBA LOCA.

Manual activation of sprays is required by control room operators, and is assumed to be manually initiated following a DBA LOCA (beginning of piping break). Initiation is based on radiation levels in the DW.

2.1.7 Atmospheric Dispersion Factors The onsite atmospheric dispersion factors (X/Q analysis) used for the radiological dose consequence analyses for both the MCR and the TSC are re-calculated for AST implementation. The current analysis is based on one year of meteorological data and the ARCON95 code. For AST, the analysis is based on a set of 3-year meteorological data and is performed with the ARCON96 code. No changes are made to the offsite atmospheric dispersion factors.

2.1.8 Cable Spreading Room Fans For HNP, the limits of unfiltered inleakage credited in the dose estimates to occupants of the MCR takes into account the operation of the MCREC system in pressurization mode. Currently, the cable spreading

Enclosure I Page 5 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment room supply and exhaust fans, 1Z41-C009 and IZ41-C010, are secured via operator action when the control room is pressurized to preclude a potential malfunction of those fans which could impact the capability to maintain the control room at a positive pressure.

A modification of the fan logic is planned to provide automatic securing of the cable spreading room supply and exhaust fans on automatic initiation of the pressurization mode of the MCREC system. This non-outage modification is scheduled to be completed by December 31, 2007.

2.1.9 FHA Decontamination Factor For the FHA, a new decontamination factor (DF) for iodine in the spent fuel pool is determined. The two regions considered for the FHA are the area over the reactor core and the spent fuel pool. The minimum depth of water over the core is 23 ft. The minimum depth of water over the fuel in the spent fuel pool is 21 ft. The iodine DF derived in RG 1.183 assumes a water depth of 23 ft. Because the depth of water over the fuel in the spent fuel pool is less than 23 ft, a DF consistent with 21 ft of water is determined for use in the FHA dose analysis in Section 2.5.3.

2.1.10 Dose Equivalent 1-131 The definition of dose equivalent 1-131 (DE 1-131) is revised to support AST implementation. "Thyroid dose" in the current definition is replaced with "committed effective dose equivalent" to more accurately depict the applicable dose component. Additionally, only Federal Guidance Report (FGR) 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1988, is referenced for the dose conversion factors used in calculating DE 1-131.

The current limit for DE 1-131 specific activity in the primary coolant is 4.0 ItCi/g. Analysis of the DBA MSLB uses a design input maximum DE 1-131 concentration of 2.0 jlCi/g. A TS change to reduce the maximum DE 1-131 concentration from 4.0 ItCi/g to 2.0 ItCi/g is part of the AST implementation. The MSLB analysis calculates doses to occupants of the MCR based upon a DE 1-131 concentration of 2.0 [tCi/g.

Table 2. Unit 1 Potential Secondary Containment Bypass Leakage Containment Line Pipe Service Isolation Size(1 ) Line System Name Description Valve Size (in) (in) Quantity(2 ) Remarks 24 (each) Design bases MSIV leakage has Nuclear Boiler Main Steam to 24 4 been considered in the dose System main turbine consequence analyses following an accident.

Condensate Drain 3 3 1

Enclosure I Page 6 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Containment Line 1

Pipe Service Isolation SizeM ) Line System Name Description Valve Size (in) (in) Quantity(2 ) Remarks Leakage through these lines Reactor 18 (each) 18 must flow through three 18-in Feedwater Supply check valves in series per line before release to the TB.

CS System Pump SupplyCondensate for Test 16 14 1 Note 3 Steam supply to 10 1 1 the HPCI turbine Pump Condensate 16 16 1 Note 3 HPCI System Suction Flow Test Leakage must pass through a Pump 18 10 1 normally closed MOV that directs Line flow to the CST via the test line.

Steam Supply to 4 1 1 the RCIC Turbine Pump Condensate 6 6 1 Note 3 RCIC System Suction Pump Flow Test Leakage must pass through a Line 18 10 1 normally closed MOV that directs flow to the CST via the test line.

RWCU Drainage to the main 6 4 1 Note 5 Condenser RWCU System Drainage to 6 4 1 Note 5 Radwaste Torus Drainage to Condenser or Normally isolated with a Suction to 8 3 1 minimum of 3 normally closed Condensate valves. Also, Note 6.

Pumps Torus Drainage and Purification Torus Drainage to Normally isolated with a System the Condensate 8 6 1 minimum of 3 normally closed Booster Pumps valves. Also, Note 6.

Torus Drainage to Normally isolated with 3 manual the waste Surge 8 4 1 normally closed valves. Also, Tank Note 6.

Enclosure I Page 7 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Containment Line Pipe Service Isolation Size(1 ) Line System Name Description Valve Size (in) (in) Quantity( 2 ) Remarks DW Equipment Drain Sump 3 3 1 Note 4 Radwaste Discharge DW Floor Drain Sump Discharge Nitrogen Supply from Nitrogen Storage Tanks to 2 (each) 2 1 Note 7 DW Pneumatic System DW Pneumatic System Suction path to the DW pneumatic system compressor DW Pneumatic has been permanently capped DWtPn 1 N/A N/A inside the RB. The drain line Suction drains into the RB equipment drain sump and does not bypass the secondary containment.

Neutron TIP Nitrogen Monitoring Purge Supply 3/8 2 1 Note 9 System DW purge Supply and Nitrogen 18, 6, 2 2, 6 2 Notes 7, 8 Make-up DW Exhaust 18, 2, 2, 2 18 1 Processed by the SGTS.

Primary Torus Purge Containment Supply and Torus 18, 6, 2 2, 6 2 Note 7, 8 Purge and Nitrogen Make-up Inerting System Torus Exhaust 18, 2, 2, 2 18 1 Processed by the SGTS.

The path is normally isolated by 1T48-F342A-L (normally closed).

Vacuum Breaker The line will be under instrument Air Supply air pressure (higher than the torus pressure) ifvalve 1T48-F342A-L is open.

Plant Service PSW Supply to 8 10 1 Closed loop system inside Water System DW Coolers primary containment.

Enclosure I Page 8 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Containment Line Pipe Service Isolation Size(1 ) Line System Name Description Valve Size (in) (in) QuantityC2 ) Remarks PSW Return from 8 30 1 Closed loop system inside DW Coolers primary containment.

RBCCW Supply 4 14 1 Closed loop system inside primary containment.

RBCCW System Closed loop system inside RBCCW Return 4 14 1 primary containment.

Demineralized Isolation valve located outside Demineralized Water Supply to 1-1/2 4 the DW is locked closed during Water System the Hose Stations normal plant operation.

Inside the DW Primary A minimum of 3 isolation valves Containment located outside the DW are Integrated ILRT Sample Line 3/4 3/4 1 closed during normal plant Leak Test System operation.

Notes:

1. Pipe size at the secondary containment wall.

2. Total number of lines that pass through the secondary containment.

3. The lines for HPCI and RCIC system pump suction piping, the CS system, CST suction piping, and the torus drainage and purification influent piping from the CST are continuously filled with water from the CST to the isolation valve, and with suppression pool water to the pump side of the isolation valve. Therefore, no leakage to the environment is postulated to occur.

4. The containment drainage sumps are located in the base of the DW and are flooded with coolant following the postulated LOCA. This flooding creates a water seal inside the containment up to the closed isolation valves. These valves are leak tested in accordance with 10 CFR 50, Appendix J, and their leakage rates form a part of the total bypass leakage fraction.

5. The RWCU system is isolated from the nuclear process through the closure of two 6-in. isolation valves in series on the influent line, and through the closure of the 18-in. feedwater system check valves at the system effluent, as well as a 3-in. RWCU system effluent check valve. The path to the RCIC is normally isolated. The leakage estimated is the combined leakage rate through the 6-in. isolation valves and the 18-in. feedwater check valves. Directing drainage to either the radwaste system or the main condenser does not affect the estimate of bypass leakage since both 4-in. lines connect to the RWCU system loop via the 6-in. and 18-in. isolation valves. See drawing nos. H16062, H16063, H16145, H16188, H16189.

6. The effluent torus drainage and purification system line, by virtue of its location with respect to the suppression chamber, is always provided with a water seal from the containment.

7. The containment gas purge supply piping is Seismic Category I piping, which is pressurized to a pressure of approximately 120 psig by the Seismic Category I nitrogen supply system, thus precluding the possibility of leakage to the environment from the containment through these lines.

8. The DW inerting piping is isolated during normal plant operation and is used only during plant startup for DW purge/inerting.

9. The TIP nitrogen supply piping is Seismic Category I piping, which is pressurized to a pressure of approximately 120 psig by the Seismic Category I nitrogen supply system, thus precluding the possibility of leakage to the environment from the containment through these lines.

2.1.11 Design Inputs and Assumptions Many of the design inputs and assumptions for the DBA radiological dose consequence analyses are

Enclosure I Page 9 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment different than those used in the current licensing basis. A listing of design inputs and assumptions used in AST analyses that are different from those used in the current licensing basis is provided in Appendix B.

2.2 Methodology 2.2.1 Accident Radiological Consequence Analysis Analyses are prepared for the radionuclide release, transport, removal, and doses for the LOCA, FHA, CRDA, and MSLB.

The ORIGEN2 code is used to calculate plant-specific fission product inventories which bound the use of 24-month fuel cycles, operation at maximum licensed power, and current fuel designs. Bounding values of fission product activity are determined for each radionuclide in the DBA radiological analyses by considering enrichment and burnup. Fission product activities are calculated for immediately after shutdown and decayed for the required times. The core inventory is multiplied by a factor of 1.1 to provide margin for future fuel changes or power uprates.

The LocaDose code is used for the DBA dose calculations. LocaDose is a proprietary code for performing multi-node radiation transport and dose calculations. It consists of inter-related modules. The modules used in the AST analyses are the Activity Transport Program, Dose Calculation Program, and Gamma Source Generation Program. The transport program calculates time-dependent isotopic activities within nodes, based on production and removal terms specified for each node. The dose program calculates doses within nodes and at offsite locations. Doses are calculated using the dose conversion factors from FGRs 11 and 12. The source program converts the isotopic activities within nodes into integrated energy release values.

LocaDose has been utilized previously at HNP. It was used for analysis in support of the Extended Power Uprate approved in 1998, and recently in the license amendment to incorporate the use of KI in the current licensing basis as an interim measure (approved May 25, 2006). LocaDose has also been used in the approved AST license amendments for Surry and Catawba.

Shield-SG is used to refine the control room immersion dose calculated by LocaDose. Shield-SG is also used to calculate the dose in the MCR from the airborne activity within the TB. Shield-SG is a point-kernel computer program which allows three dimensional modeling of shielding problems.

STARNAUA is a Polestar Applied Technology, Inc. (Polestar) proprietary code that analyzes activity transport in particulate form. STARNAUA is used to model aerosol removal in containment, taking into account a number of processes including gravitational settling, diffusion of particles to surfaces within the containment volume, removal by sprays, and leakage of particles from the containment.

STARNAUA has been utilized in the AST license amendments for which the NRC issued Safety Evaluations for Perry and the Westinghouse AP600 and API000 designs. STARNAUA was utilized for Oyster Creek and Columbia, whose AST license amendment requests are under review.

Table 3 summarizes the various applications of Polestar's AST methodology for various sites and reactor designs.

Enclosure I Page 10 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment 2.2.2 Suppression Pool pH Control Analysis The calculation methodology for suppression pool pH control is based on the approach outlined in NUREG-1465 and NUREG/CR-5950. Specifically, credit is taken for SPB addition to the suppression pool water as a result of SLC system operation. The pH of the suppression pool water is then calculated using the STARpH code. STARpH is a Polestar proprietary code that analyzes suppression pool pH as a function of time for systems containing radiolytically-produced strong acids and a variety of buffers and bases that may be used to control pH.

Calculations are performed to verify that sufficient SPB solution is available to maintain the suppression pool pH at or above 7 for 30 days post-accident. The design inputs are conservatively established to maximize the post-LOCA production of acids and to minimize the post-LOCA production and/or addition of bases. Other design input values such as initial suppression pool volume and pH are selected to minimize the calculated pH.

STARpH was used for applications receiving NRC Safety Evaluations for Perry, Vermont Yankee, Hope Creek, Waterford 3, and Browns Ferry. AST license amendment requests utilizing STARpH which are under review include Salem, Oyster Creek, and Columbia.

2.2.3 NUREG-0737 Analysis An evaluation is performed to identify potential impacts of applying AST methodologies to NUREG-0737 items. This includes the following:

• Evaluation of the current radiological dose analyses for post-accident vital area access (NUREG-0737, Item II.B.2)

• Evaluation of the current radiological dose analyses for the post-accident containment high-range radiation monitors (NUREG-0737, Item II.F. I)

• Evaluation of control room post-accident radiological dose analyses for control room habitability, including habitability of the TSC (NUREG-0737, Items III.A.1.2 and III.D.3.4)

" Consideration of post-accident sources of radiation and radioactivity outside the primary containment in terms of impact on dose analysis related to integrity of systems outside containment likely to contain radioactive material (NUREG-0737, Item III.D. 1.1).

Enclosure 1 Page 11 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Table 3. Polestar Methods for Activity Removal Main NRC Safety DW Spray Condenser STARNAUA Evaluation AP-600/1000 X X Perry X X X Browns Ferry X X Vermont Yankee X X X Columbia X X X Oyster Creek X X Hatch X X X Notes:

1. AP-600 and AP-1000 are advanced PWR designs (passive plants)

2. Perry design (Mark III containment) uses containment spray, not DW spray

3. The STARNAUA application at Columbia was used to adjust documented NRC Office of Nuclear Regulatory Research acceptance for Perry to account for DW spray

4. The NRC Safety Evaluation for Columbia was completed by the Dose Assessment Branch 2.3 Environmental Qualification The radiation doses in the existing environmental qualification analyses were calculated using source terms determined by Technical Information Document (TID)-14844 methodology. Consistent with current regulatory allowance, the environmental qualification of equipment is bounded by, and will continue to be based on, TID-14844.

2.4 Atmospheric Dispersion Factors 2.4.1 Meteorological Data The existing MCR y/Q analysis was performed by use of one year meteorological data and ARCON95 in 1997 and has been approved by NRC. The postulated release locations from the reactor building (RB) and TB have been reviewed and approved by NRC. However, NRC recommended that any future analysis should be performed by use of ARCON96. Additionally, a Nuclear Energy Institute/NRC panel on "Control Room Habitability Analyses at Nuclear Power Plants" has recommended that 3-year meteorological data sets are more. representative of the atmospheric dispersion conditions at a specific site than the use of one year data and should be used in future applications. Therefore, the x/Q analysis for AST application is based on a set of 3-year meteorological data (1996-98) and is performed by use of ARCON96.

The maximum atmospheric dispersion factors (sec/m 3) for the MCR and TSC are shown in Table 4. The RB vent x/Q values apply to any leakage prior to RB drawdown as well as stack bypass leakage after drawdown. The RB vent X/Q values also apply to MSIV leakage in the TB.

Offsite atmospheric dispersion factors for containment, RB, and TB releases are shown in Table 5. The ground level x/Q values apply to any leakage from the containment or RB prior to RB drawdown as well Page 12 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment as stack bypass leakage after drawdown. The ground level X/Q values also apply to MSIV leakage in the TB. The elevated X/Q values apply to RB releases through the main stack after drawdown.

2.4.2 ARCON96 ARCON96 is a computer model developed by Pacific Northwest Laboratory for the NRC to estimate x/Q values for onsite receptors near building structures. The model uses hourly meteorological data and recently developed methods to estimate atmospheric dispersion. ARCON96 is capable of evaluating ground-level, vent, and elevated releases. A vent release is a release that takes place through a roof-top vent with an uncapped vertical opening. This model also treats diffusion more realistically under low wind conditions than previous NRC-issued models.

Table 4. Atmospheric Dispersion Factors - Main Control Room yIQ (sec/r 3 )

Release Point => RB Vent Main Stack Receptor => MCR Intake MCR Intake 0 - 2 hr 1.41 E-3 3.76E-6 2 - 8 hr 1.08E-3 2.88E-6 8 - 24 hr 4.70E-4 7.50E-7 1 - 4'day 3.54E-4 7.67E-7 4 - 30 day 2.67E-4 5.04E-7 Note: These x/Q values are also applied to the TSC.

Enclosure I Page 13 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Table 5. Atmospheric Dispersion Factors - Offsite 71Q (sec/m 3)

Release Point => Ground Elevated Receptor => EAB LPZ EAB LPZ 0 - 2 hr 3.1E-4 3.1E-4 1.7E-6 1.7E-6 2-8hr 3.1E-4 1.7E-4 1.7E-6 9.4E-7 8 - 24 hr 2.3E-5 3.9E-7 24 - 96 hr Not Applicable 1.1 E-5 Not Applicable 2.OE-7 96 - 720 hr 4.5E-6 8.OE-8 Notes:

1. Although the EAB dose is calculated for a 2-hr period only, the x/Q values are applied for 8 hr to determine the worst-case 2-hr dose.

2. The elevated x/Q values apply to RB releases through the main stack after drawdown. The ground level x/O values apply to other releases from the containment, the RB, and the TB.

2.5 Accident Radiological Consequence Analyses The DBA accident analyses documented in Chapter 15 of the HNP FSAR that could potentially result in control room and offsite doses are addressed using methods and input assumptions consistent with the AST. The following DBAs are addressed:

" LOCA, FSAR Section 15.3.3

" FHA, FSAR Section 15.3.5

• CRDA, FSAR Section 15.3.2

" MSLB Accident, FSAR Section 15.3.4 Radiological consequences in the analyses have changed due to the impact of the characteristics of the AST itself and licensing basis changes that are being made concurrent with the AST implementation.

Analyses are performed per RG 1.183. Documentation of conformance to RG 1.183 is included as Appendix A to this enclosure. The results of the radiological dose consequence analyses are evaluated to confirm compliance with the acceptance criteria presented in 10 CFR 50.67 and GDC 19 of 10 CFR 50, Appendix A.

Radiological dose consequences are evaluated for individuals located at two offsite locations, the exclusion area boundary (EAB) and the outer boundary of the low population zone (LPZ). Dose consequences are also evaluated for personnel occupying the MCR and the TSC. Although RG 1.183 does not address TSC dose limits, the dose limits that apply to the control room are assumed to apply to the TSC as well.

2.5.1 Common Inputs and Assumptions The design inputs, parameters, and assumptions that are common to multiple DBAs are presented in Table 6.

Enclosure I Page 14 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Table 6. Common DBA Radiological Consequence Analyses Inputs and Assumptions Input / Assumption Value Reactor Power 2818 MWt (current licensed rated thermal power level times an uncertainty factor of 1.005)

The equilibrium core inventory of fission products per unit power (Ci/MWt) has been generated using the ORIGEN2 computer program based on a 24-month fuel cycle. The inventory is limited to the Core Inventory radionuclide groups and elements specified in RG 1.183, Table 5. In addition, a 10% margin is incorporated into the core inventory to allow for future fuel changes or power uprates. The core inventory (without the 10% margin) is given in Table 7.

'Volumes MCR Free Volume 9.35E4 ft3 TSC Free Volume 1.56E4 ft3 TB Free Volume 6.50E6 ft3 Turbine / Condenser Volume 1.72E5 ft3 (combined volume of low-pressure turbine and condenser)

MCR/ TSC Ventilation The design value is 400 cfm, however, to allow for MCR Filtered Intake Rate the potential need for less flow to pressurize the MCR, the filtered intake rate conservatively assumed in the analyses is 250 cfm.

MCR Recirculation 2,100 cfm MCR Filter Efficiency (Intake and Recirculation) 95% for all radionuclides except noble gases MCR Unfiltered Inleakage 115 cfm (for LOCA, CRDA, and MSLB)

MCR Ingress / Egress 2 one-way trips per day, lasting 2 min (each way)

TSC Filtered Intake Rate 500 cfm TSC RecirculationJ 500 cfm TSC Filter Efficiency (Intake and Recirculation) j 90% for all radionuclides except noble gases Page 15 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Input I Assumption Value TSC Unfiltered Inleakage 10,000 cfm Breathing Rates and Occupancy Factors Breathing Rate (MCR, TSC, TB) 3.5E-04 m3/sec 0-8 hr: 3.5E-04 m3/sec Breathing Rate (Offsite) 8-24 hr: 1.8E-04 m3/sec 1-30 days: 2.3E-04 m3/sec 0-1 day: 100%

MCR and TSC Occupancy Factors 1-4 days: 60%

4-30 days: 40%

TB Ventiiation 9 hr after initiation of the accident (credited TB Fans - Time to Start operator action). TB exhaust is credited for the LOCA, CRDA, and MSLB.

15,000 cfm (capacity of one fan) unfiltered release TB Fan Exhaust Rate to the environment via the RB vent. TB exhaust is credited for the LOCA, CRDA, and MSLB.

Enclosure 1 Page 16 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Table 7. Core Inventory I

Isotope Source 1I/MWt)

(Ci/MWt) Isotope Source I ISource I ISource Isotope (Ci/MWt) Isotope (Ci/MWt) 1-129 1.23E-03 Se-79 1.66E-02 Ru-103 4.25E+04 Sm-1 51 1.90E+01 1-130 1.05E+03 Sb-124 4.54E+01 Ru-105 2.97E+04 Sm-1 53 1.37E+04 1-131 2.72E+04 Sb-125 6.1 OE+02 Ru-106 1.70E+04 Eu-152 4.70E-01 1-132 3.93E+04 Sb-126 3.58E+01 C) Rh-1 03m 3.83E+04 Eu-154 3.93E+02 1-133 5.52E+04 Sb-126m 1.36E+01 .Q Rh-105 2.72E+04 Eu-1 55 2.75E+02 1-134 6.05E+04 Sb-1 27 2.98E+03 0 Rh-106 1.82E+04 Eu-1 56 5.09E+03 0 z 1-135 5.16E+04 0 Sb-1 29 8.85E+03 Pd-107 4.11E-03 CO Am-241 6.86E+00 1-136 2.45E+04 Te-125m 1.33E+02 Pd-109 9.40E+03 ca

=C Am-242 2.85E+03

-4 1-4I 1-137 2.39E+04 2 Te-127 3.00E+03 Y-89m 6.27E-04 Am-242m 9.30E-01 1-138 1.18E+04 (D Te-127m 4.OOE+02 Y-90 3.20E+03 Am-243 7.46E-01 Kr-83m 3.30E+03 Te-129 8.71 E+03 Y-90m 4.61 E-01 Cm-242 1.68E+03 Kr-85 3.78E+02 Te- 129m 1.29E+03 Y-91 3.23E+04 Cm-243 8.99E-01 2.41 E+04 Y-91m 1.83E+04 Cm-244 1.05E+02 Kr-85m 6.92E+03 Te-131 Te-131m 3.94E+03 Y-92 3.44E+04 Cm-245 1.OOE-02 Kr-87 1.32E+04 Te-132 3.85E+04] Y-93 4.OOE+04 Cm-246 1.65E-03 Kr-88 1.86E+04 z) 3.24E+04 Y-94 4.05E+04 Br-82 1.81E+02 Kr-89 2.26E+04 Te-133 0

(1) Xe-131m 3.03E+02 Te-133m 1.99E+04 Y-95 4.36E+04 Br-83 3.30E+03 CA Br-84 0 5.69E+03 Xe-133 5.27E+04 Te-134 4.53E+041 Zr-93 9.53E-02 C) 0)

4 4-4 +

Xe-1 33m 1.58E+03 Sr-89 2.49E+04 Zr-95 4.77E+04 Br-85 6.83E+03 Iz Xe-135 1.89E+04 Sr-90 3.01 E+03 Zr-97 5.OOE+04 Br-87 1.11E+04 Xe-135m 1.09E+04 Sr-91 3.15E+04 Nb-93m 8.21 E-03 Br-88 1.19E+04 4 4-4 4.

Xe-1 37 4.81 E+04 Sr-92 3.42E+04 ca Nb-95 4.79E+04 Ce-141 4.48E+04

.2 Xe-138 4.52E+04 Sr-93 3.90E+04 Nb-95m 3.37E+02 Ce-143 4.13E+04 CO Rb-86 7.06E+01 Sr-94 3.69E+04 Nb-97 5.04E+04 Ce-144 3.69E+04

.J 0

Rb-88 1.89E+04 mt Sr-95 3.43E+04 La-140 4.88E+04 Np-237 1.61 E-02 Rb-89 2.42E+04 u) Ba-1 36m 3.60E+02 La-141 4.47E+04 Np-238 1.48E+04 Rb-90 2.34E+04 Ba-1 37m 3.92E+03 La-142 4.31 E+04 0- Np-239 5.71 E+05

.5 ,

Cs-134 6.83E+03 Ba-139 4.91E+04 La-143 4.11E+04 Pu-236 2.74E-02 (5

Cs-134m 1.65E+03 Ba-140 4.73E+04 Pr-143 4.04E+04 Pu-237 1.04E-01 E)

Cs-135 2.35E-02 Ba-141 4.45E+04 Pr-144 3.71 E+04 0D Pu-238 1.31 E+02 Cs-136 2.18E+03 Co-58 1.48E+02 Pr-144m 4.44E+02 Pu-239 1.29E+01 Cs-137 4.14E+03 M Co-60 4.27E+02 Nd-147 1.80E+04 Pu-240 1.78E+01 Cs-1 38 5.02E+04 Mo-99 5.15E+04 Pm-1 47 4.53E+03 Pu-241 5.23E+03 Cs-139 4.75E+04 ') Tc-99 5.24E-01 Pm-148 7.82E+03 Pu-242 6.19E-02 z.

z Tc-99m 4.53E+04 Pm-148m 1.16E+03 Pu-243 1.06E+04

+ 1-4 Tc-101 4.63E+04 Pm-149 1.60E+04 4 4-4 +

Pm-1 51 5.49E+03 1 ______________

___________________ A ____________________ 4. ______________ ____ ____________ 4. ______________

Note: In calculating doses, the progeny of these isotopes is also included and the activities are increased by 10% to allow for future fuel changes or power uprates.

Page 17 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment 2.5.2 Loss-of-Coolant Accident (LOCA)

The radiological dose consequences to occupants of the MCR and the TSC and to persons located at the EAB and LPZ following a postulated LOCA are determined using AST assumptions and methodology in accordance with RG 1.183.

2.5.2.1 Inputs and Assumptions Releasefrom Core In accordance with RG 1.183, Table 4, the core activity is assumed to be released into the containment in two phases: gap activity release (starts at 2 min and lasts 30 min) and early in-vessel release (starts at the conclusion of the gap activity release phase and lasts for 90 min). The core fractions of the radionuclide groups that are released during each phase are based on RG 1.183, Table 1. The release rates are shown in Table 8.

Chemical Form By adding boron to the suppression pool as a credited operator action, the pH of the torus water is maintained at a value of 7 or higher. This is based on full mixing of the boron within the pool water within the first 24 hr of the accident. With this pH, the chemical form of iodine released to the containment is assumed to be 95% cesium iodide (CsI), 4.85% elemental, and 0.15% organic per RG 1.183, Appendix A, Section 2. With the exception of elemental and organic iodine and noble gases, all isotopes are assumed to be in particulate form.

Containment Volume and Mixing It is assumed that the activity released from the fuel is instantaneously and homogeneously mixed throughout the free air volume of the DW (RG 1.183, Appendix A, Section 3.1). It is also assumed that the DW activity starts flowing into the torus air volume at the end of the early in-vessel release phase as shown in Table 9.

Containment Activity Removal Credit is taken for the reduction of airborne activity due to natural deposition and sprays (RG 1.183, Appendix A, Sections 3.2 and 3.3). The removal rates for elemental iodine and particulates within the DW due to natural deposition (sedimentation) and sprays (with the sprays assumed to start at 15 min as a credited operator action) are shown in Table 10. The removal rates are based on credit for one residual heat removal (RHR) pump. The manual initiation of DW sprays is a required operator action that is initiated based upon reaching a DW dose rate of 200,000 rem/hr. Analysis of the DW immersion dose rate following a DBA LOCA shows that this dose rate is reached within 15 min of accident initiation.

Modeling with STARNAUA Aerosol removal in containment is governed by a number of processes modeled by the STARNAUA computer code including gravitational settling (sedimentation) and removal by sprays. In addition, agglomeration (coagulation) of particles is modeled; removal by sprays may be considered a special case of agglomeration.

Agglomeration is more pronounced when the number density of particles in the containment atmosphere is large. It is apparent from Stokes Law that larger particles are removed more efficiently Page 18 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment than smaller ones (both for sedimentation and for spray removal); therefore, agglomeration can substantially increase the removal rate. Because large particles are more readily removed than smaller ones, the particle size distribution gets further depleted of large particles with time.

Agglomeration mitigates this trend, tending to be a source of large particles.

As noted, an important special case of interest occurs when containment sprays are present. Here the agglomeration takes place between the very large spray droplets and the aerosol particles, which results in a very efficient process for removing the particles.

Sedimentation is always an important removal mechanism for aerosol particles and is often the predominant one, even in the absence of sprays. When sprays are present, sedimentation of the compound spray droplet/particle accounts for almost all of the aerosol removal.

Sedimentation of aerosols (as opposed to that of spray droplets) is well understood in terms of the Stokes-Cunningham law which gives the terminal settling velocity for a single particle of actual radius rp as:

2Cg(pp - pgas)r2 9

Ptgas where:

v, = settling velocity, cm/sec g = gravitational acceleration, 980 cm/sec 2 3

pp = density of the particle, g/cm 3 Pgas = density of air (or the containment atmosphere), g/cm t

ags, = viscosity of air (or the containment atmosphere), g/cm-sec C = Cunningham slip factor = +- .246+ .42exp( 0.87rp rp, X = gas mean free path, cm The Stokes law expression above is valid for particles of radius less than - 50 microns (i.e., aerosols) which adequately covers the particle size range of interest. (It does not apply to spray droplets, which are considerably larger and are treated differently.)

Settling and spray removal rates are strongly dependent on the size (and material density) of the aerosol particles. In STARNAUA, the aerosol population is characterized by a size distribution which evolves in time from its initial (source) function to a time-dependent distribution as particles of different sizes are added, agglomerate, and/or are removed at different rates. The initial source size distribution is assumed to be characteristic of the aerosol released into the containment.

In so-called "discrete" codes (including STARNAUA) the size distribution is defined by choosing appropriate minimum and maximum particle sizes and dividing the interval between the two into a number of "bins." The initial size distribution is then entered as the fraction'of the total number of particles assigned to each bin. The population of each bin is then followed as a function of time. The larger the number of bins, the more accurately the distribution will be represented, but at the cost of

Enclosure I Page 19 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment increased computing time. The HNP analysis for DW spray was conducted using 30 bins covering 4.54E-03 microns to 90.8 microns aerodynamic radius for the initial distribution.

It has been observed that many aerosol distributions (number of particles, N, as a function of particle radius, r), both those occurring naturally and those resulting from industrial or other processes of human origin, are of the log-normal type. It should be noted that such a distribution is completely defined by two parameters: (ln(r))m and ag. (ln(r))m is the mean value of ln(r) for the aerosol population. To characterize the initial (source) distribution, STARNAUA replaces (ln(r))m by the closely related parameter rg, where rg is the (geometric) mean particle radius (ln(rg) = (ln(r))m).

Polestar analyzed the results of a number of large-scale fuel melt experiments in order to obtain values of rg and ag. The results were then averaged to obtain what Polestar believes to be the most representative values of the source size distribution in a reactor fuel melt accident. The representative values are rg = 0.22 gm and a = 1.81 ((g = 0.5933). These are the values that are normally input into STARNAUA calculations at Polestar (except for a small correction due to the void fraction of the particles).

An important consideration in sedimentation arises from the fact that the aerosol particles are not considered to be solid, but are assumed to have void fractions that are filled with gas if the containment atmosphere is dry or filled with water if the atmosphere is at or near saturation. For plants with sprays operating, it is generally assumed that the voids are water-filled.

The mechanisms that contribute to the spray collection efficiency modeled in STARNAUA include interception, impaction, Brownian diffusion of aerosol particles to the droplet, and diffusiophoretic deposition of particles to the droplets if the thermal-hydraulic conditions result in steam condensation on the droplets. The latter effect is neglected for HNP. The overall spray collection efficiency is the sum of the individual efficiencies of these processes, which are dependent on both the droplet and the particle sizes.

Thermal-Hydraulic Considerations The density and viscosity of the containment atmosphere are functions of input thermal-hydraulic conditions in the containment (temperature and gas composition (steam/nitrogen ratio)), which will vary with time. STARNAUA contains function statements that yield these quantities at each time step in the calculation. Three important phases in the assumed HNP accident sequence are:

1. Prior to DW spray actuation (0 to 15 min, with the activity release beginning at 2 min)

2. Release phase with sprays operating (15 min to 122 min)

3. Post-release phase (122 min to 24 hr)

During Phase (1), the only removal phenomenon that is credited is sedimentation. For sedimentation, the higher the gas temperature, the less effective the aerosol sedimentation. Consequently, for purpose of conservatism, the thermal-hydraulic inputs for this phase are chosen at a point in time when the DW temperature is the highest. 2 During Phases (2) and (3), DW sprays are operating. Unlike aerosol sedimentation (where the 2 Due to the conservative modeling for containment activity removal, the values for temperature and pressure used by the STARNAUA code (and presented in Table 11 and Table 12) are different from the temperature and pressure values presented in Table 16.

Enclosure I Page 20 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment conservatism of high temperature may correspond to a low gas density), the higher the gas density, the less effective the spray removal. Therefore, the thermal-hydraulic inputs for these two phases are chosen at a point in time when the DW pressure and gas density is the highest. The DW thermal-hydraulic data used by STARNAUA are shown in Table 11, Table 12, and Table 13.

Mass of Inert Aerosol Release In addition to fission products released as aerosols, non-fission product fuel and structural ("inert")

material aerosols are also released. Although these do not contribute significantly to radiation exposure, their presence in the total aerosol is important, since they do contribute to the aerosol number density in the containment atmosphere and take part in aerosol agglomeration. Thus, they influence the removal rates of the fission product aerosols from the containment atmosphere. It is thus essential to include them in the aerosol source term in STARNAUA calculations. The ratio of structural material to fission product aerosols should be at least 2.4. This value is considered to be conservative, i.e. it will result in less aerosol removal than a larger value would. For HNP, an even more conservative value of 1: 1 was used.

In a STARNAUA calculation the fission product aerosol release in each release period for each fission product is determined on the basis of its core inventory at the time of the accident and its release fraction from the core. The fission product release rates are summed and for the in-vessel release period the inert release rate is taken as equal to the sum of the total fission product releases (gap plus in-vessel release periods) times the assumed inert/fission product ratio. It is assumed that no inert release occurs during the gap release period.

The removal of elemental iodine is assumed to occur at the same rate and with the same degree of completeness as particulate. This is based on the propensity for elemental iodine to adsorb onto surfaces (in this case, the large surface area of the dispersed particulate). Once the iodine is dissolved in the spray water, for a suppression pool pH of 8.3 at 24 hr, the point in time when spray credit ceases (see Section 2.6), the ratio of iodine concentration in the liquid phase to that in the gas will exceed 30,000. For a primary containment gas-to-liquid volume of approximately 3 for HNP, this means that not more than 0.01% of the total iodine would remain airborne as elemental iodine at that time, and most likely, even less. This is negligible considering the residual 0.15% iodine airborne in organic form. With the suppression pool pH being greater than 7.7 even at 30 days (well above neutral pH), re-evolution of elemental iodine later in the accident does not need to be considered.

ContainmentLeakage The primary containment (DW and torus air) is assumed to leak at the peak pressure TS leak rate of 1.2% weight per day for the first 24 hr of the accident, reducing by 40% from 24 to 72 hr, and by 50%

thereafter, as justified in Section 2.5.2.2. Assuming the activity within the containment to be uniformly distributed, the volumetric leak rate is the same as the mass leak rate. Starting with the gap activity release phase, all the leakage from the primary containment (excluding MSIV leakage) enters the secondary containment except for 2% that is assumed to bypass the secondary containment. The bypass leakage is assumed to be directly to the environment at ground level in evaluating offsite and TSC doses and into the condenser in evaluating MCR doses.

Secondary Containment It is assumed that the RB (secondary containment) draws down to negative pressure within 2 min of the start of the accident. After secondary containment drawdown, RB activity is released to the environment through the plant stack at the maximum TS rate of 4,000 cfm per unit. It is possible for

Enclosure I Page 21 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment two standby gas treatment system (SGTS) fans to be in operation at the same time, taking suction from one RB and the refueling floor. In order to maximize the release to the environment, it is conservatively assumed that the entire flow of 8,000 cfm from two fans is from a single RB. The release is processed by the SGTS filters with an efficiency of 95% for all isotopes except noble gases.

The activity within the RB is assumed to be uniformly distributed within 50% of the volume (RG 1.183, Appendix A, Section 4.4). The total volume of the RB is obtained by adding the individual compartment volumes, conservatively neglecting the common refueling floor to minimize the dilution.

ESF Leakage In modeling engineered safety feature (ESF) leakage, it is conservatively assumed that all the isotopes that are released to the containment except noble gases are instantaneously transported to the torus water at the onset of the gap activity release phase and mixed uniformly (RG 1.183, Appendix A, Section 5.1). Although there is no TS limit on ESF leakage, a conservatively high leakage rate of 10 gpm (1.34 cfm) is assumed to start at the initiation of sprays, lasting for the duration of the accident. With the torus water temperature below 212'F, it is assumed that 10% of the iodine in the ESF leakage becomes airborne inside the RB while all other elements remain in the water; of the iodine that becomes airborne, 97% is assumed to be elemental and 3% organic (RG 1. 183, Appendix A, Section 5.3 to 5.6).

MSIV Leakage It is assumed that the maximum leakage from all four MSLs is 100 scfh,3 with no limit on the leakage per line. It is postulated that the inboard MSIV on one of the four steam lines fails to close, thus creating an unrestricted flow path to the outboard MSIV. It is assumed that the full leakage of 100 scfh is through this failed line with no MSIV leakage through any of the three intact lines. This is conservative as there is less activity removal due to deposition within a line with a failed MSIV than with intact MSIVs. The leakage rate is reduced by 40% at 24 hr and by 50% at 72 hr, as justified in Section 2.5.2.2.

The source of the leakage is the airborne activity in the DW (RG 1.183, Appendix A, Section 6.1).

The MSIV leakage could take place in either the RB or the TB. Since a leakage in the RB would have the benefits of filtration and dispersion due to an elevated release through the stack after RB drawdown, it is assumed that all MSIV leakage occurs in the TB. Since the MCR is located within the TB, it is conservative to calculate MCR doses assuming holdup in the TB as this provides a direct inleakage pathway from the TB to the MCR.

Main Steam Line Activity Removal The calculation of aerosol removal in the MSLs is also accomplished with STARNAUA. 4 It is assumed that particulate in the portion of the MSL (I) between the inboard and the outboard MSIVs for three lines and (2) between the outboard MSIV and the main stop valve for all lines is subject to removal by deposition, as allowed by RG 1.183, Appendix A, Section 6.5. Since it may be assumed that all MSIV leakage occurs in a single line (the limiting case being the line with only one MSIV closed), the limiting case involves deposition only between the outboard MSIV and the main stop 3 100 sch represents the maximum allowable MSIV leakage at reduced test pressure. For testing at or above accident pressure, the maximum allowable MSIV leakage is 144 scfh. See Section 2.5.2.2 for further discussion.

4 The methodology of AEB 98-03 (Reference 2) is not utilized.

Page 22 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment valve. Table 14 shows the time-dependent deposition rates for that limiting case. The particle size distribution used as input for the MSL analysis is that representing airborne activity from the DW analysis.

Because sedimentation is minimized by the assumption of high temperature, the steam line is conservatively assumed to remain at its maximum temperature (551VF) during the full duration of the analysis. Only horizontal runs of steam line are credited for sedimentation, and only the projected area of the steam line is used as sedimentation area.

It is also assumed that particulate mass and activity and elemental iodine activity are reduced by a factor of two due to particle impaction at the inboard or outboard MSIV, whichever is the first closed valve encountered (as noted, three lines are assumed to have both valves closed). Once the particulate enters the steam line beyond the first closed MSIV, however, no further elemental iodine removal is considered in the steam lines or main condenser. This is very conservative, because even if re-evolution from the particulate surfaces were to occur in the hot, dry conditions of the steam line, some deposition and retention would be expected on the metal surfaces in those volumes, as well.

Condenserand Turbines Most of the MSIV leakage reaches the condenser and the low pressure turbine while a small fraction bypasses the condenser/low pressure turbine and is released through the high pressure turbine (HPT).

The bypass fraction is calculated to be 0.005 per the methodology given in Reference 3. It is assumed that particulates in the condenser are removed by sedimentation as shown in Table 15. Although the entering flow is 100 scfh less the bypass fraction, the condenser leak rate is assumed to be 100 scfh during the first 24 hr, reducing by 40% at 24 hr and by 50% at 72 hr.

Particulate removal by sedimentation in the condenser is credited for both MSIV leakage and secondary containment bypass leakage.

Turbine Building In calculating MCR doses, it is assumed that releases from the condenser and the turbines are into the TB, where they are available for direct inleakage into the MCR. It is assumed that the released activity is uniformly mixed within the volume of TB elevation 164 ft floor, which is open to both units.

Atmospheric Dispersion All releases from the TB are assumed to be at ground level through the RB vent. MCR atmospheric dispersion factors (X/Q) are shown in Table 4. The MCR values are applied to the TSC since the MCR values are bounding. The EAB and LPZ X/Q values are shown in Table 5. The 0-2 hr EAB x/Q values are applied for the first 8 hr to ensure that the EAB dose is calculated over a 2-hr period that yields the maximum dose per RG 1.183, Section 4.1.5.

Accident Duration MCR, TSC, and LPZ doses are calculated assuming a release duration and exposure time of 30 days per RG 1.183, Table 6. The EAB dose is calculated over a two-hr period which yields the maximum dose per RG 1.183, Section 4.1.5.

Enclosure I Page 23 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Credited OperatorActions The following actions taken by operators are credited in the analysis:

1. MSIV ALT Pathway - Lining up the MSIV ALT pathway by opening a valve to establish the pathway and closing boundary valves to direct MSIV leakage to the condenser for holdup and retention of activity in MSL piping and the condenser.

2. Addition of pH Buffering Agent - Addition of SPB to the suppression pool via operation of the SLC system to maintain suppression pool pH at 7 or higher for the duration of the accident, thereby precluding the re-evolution of iodine from the suppression pool.

3. DW Sprays - Initiation of DW sprays based on radiation levels in the DW to help remove airborne particulate in the DW and reduce DW temperature and pressure.

4. TB Fans - Initiation of one TB exhaust fan within 9 hr of the start of the accident to remove activity from the TB.

2.5.2.2 Method of Evaluation Doses for the LOCA are calculated using the guidance provided in the main body and Appendix A of RG 1.183. The computer codes LocaDose and Shield-SG are utilized to calculate doses. Descriptions of these codes are provided in Section 2.2.1. The inputs and assumptions used to determine offsite, MCR, and TSC doses are described in the previous section. Assumptions requiring further elaboration are discussed below.

Activity transport models are developed for both the RB and the TB. There are three main activity release pathways:

• Containment leakage - This includes leakage of airborne activity in the DW and the torus air space. Initially any leakage is assumed to be directly to the environment at ground level. After RB drawdown to negative pressure, the leakage is assumed to be processed by the SGTS and released through the plant stack except for a small fraction that bypasses the SGTS.

• MSIV leakage - This leakage reaches the condenser via the steam lines. The condenser is then assumed to leak to the TB and eventually to the MCR and the environment.

• ESF leakage - This occurs in the RB after the DW sprays have been initiated and water from the torus is recirculated back into the DW.

To determine offsite doses, two models are developed. The first calculates doses from elevated releases from the RB via the SGTS. The second model is for ground releases, which include RB activity bypassing the SGTS and MSIV leakage activity leaking from the condenser. The ground level releases are through the RB vent. Total offsite doses are a combination of doses from elevated releases and ground releases.

In evaluating TSC doses, RB activity is released to the environment through the stack via the SGTS and the RB vent (for activity that bypasses the SGTS). MSIV leakage activity leaking from the condenser is released to the environment via the RB vent. Activity released to the environment reaches the TSC via outside air intake and unfiltered inleakage.

For MCR doses, RB activity is released to the environment through the stack Via the SGTS. RB bypass leakage is assumed to go into the condenser, where it leaks into the TB. MSIV leakage activity also

Enclosure 1 Page 24 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment collects in the TB via leakage from the condenser. Beginning at 9 hr after the accident, TB air is exhausted at a rate of 15,000 cfm. Activity released to the environment reaches the MCR via an outside air intake at a rate of 250 cfm. TB activity also leaks directly into the MCR (unfiltered inleakage).

SLC Injection HNP takes credit for the SLC system for the injection of a sufficient quantity of SPB solution into the reactor vessel, and ultimately mixing in the suppression pool, to meet the requirement for maintaining pH at or above 7. The mixing scenario is as follows:

1. It is conservatively assumed that there is no SLC injection during the first 2 hr of the accident.

Applicable plant procedures will be revised as needed to ensure that SLC injection is initiated within 2 hr of accident initiation.

2. Starting at 2 hr, SLC injection is initiated and lasts for up to 1.5 hr.

3. After SLC injection is completed at 3.5 hr, the core spray (CS) system floods the reactor vessel.

Assuming a single CS pump at a minimum flow rate of 4,000 gpm is used to fully replenish the entire primary system volume, the time required for the fill is 0.5 hr.

4. After the reactor vessel is filled at about 4 hr, the excess liquid will spill out of the break and reach the suppression pool, thereby initiating the mixing of the borate solution within the pool.

5. Suppression pool water is recirculated through the reactor vessel and the DW via the CS and RHR systems, respectively, resulting in further mixing. Based on the suppression pool volume, the flow rate of a single RHR pump, and the CS flow rate, the time required for one turnover of the suppression pool water is approximately 1 hr.

Therefore, it takes about 5 hr to complete the first turnover of the suppression pool water. As each subsequent turnover of the suppression pool water requires about I hr, there will be about 20 turnovers during the first 24 hr. It is expected that a few turnovers of the pool volume is all that is required to fully mix the SPB solution.

The calculation of the suppression pool pH is described in Section 2.6.

Reduction in Containmentand MSIV Leakage Rates Leakage may be reduced after the first 24 hr, if supported by plant configuration and analyses, to a value not less than 50% of the TS leakage rate (RG 1.183, Appendix A, Section 3.7 and 6.2).

For pressurized water reactors (PWRs), RG 1.183 allows the containment leakage rate to be reduced by 50% without further justification. The containment leakage rate is dependent on the containment pressure. For both BWRs and PWRs, after the initial transient conditions following a LOCA, the containment pressure steadily decreases with time unless some action is taken to increase the pressure.

Some BWRs, including HNP, have containment atmosphere dilution (CAD) systems to control the hydrogen and oxygen levels inside the containment. The CAD system works by injecting nitrogen into the containment, which has the potential to increase pressure. The severe accident guidelines and emergency operating procedures for HNP call for the DW and the torus to be vented while nitrogen is being injected such that the pressure does not increase. Therefore, without a pressure increase associated with the CAD system, the HNP containment leakage rate would be expected to decrease in the same manner as a BWR without a CAD system or a PWR.

Enclosure 1 Page 25 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment The containment and MSIV leakage rates are both based on the peak DW pressure and temperature. The volumetric flow rates are calculated at these peak conditions as well as the later time steps, when the pressure and temperature have decreased. By comparing the flow rates at different time steps, the reduction in leakage rate can be calculated at 24 hr or any other time step. Since the flow rate through the MSLs is driven by the pressure difference between the DW upstream and the condenser downstream, it is reasonable to assume that the MSIV leakage rate will be reduced by the same magnitude as containment leakage.

It is assumed that the containment and MSIV leakage paths are sufficiently restrictive that the flows are unchoked and may be treated as incompressible. The containment leakage is into the RB while the MSIV leakage is into the TB via the condenser. For both leakages, the downstream pressure in the RB and TB is assumed to be atmospheric at' 14.7 psia.

The use of sprays reduces the DW pressure from a peak value of 65.5 psia in Unit 1, and a peak value of 62.0 psia in Unit 2. A summary of parameters for containment leakage reduction is shown in Table 16.

The MSIV leakage rate is based on a peak pressure of 61.6 psia and a peak temperature of 340°F. The saturation steam pressure at 340°F exceeds 61.6 psia, meaning the steam is superheated. In applying the flow equation at 0 hr, increasing the density causes the flow rate to decrease, which results in higher flow ratios at 24 and 72 hr. Hence, to minimize the initial flow rate, the maximum possible air density is added to the superheated steam density. The air pressure used to calculate the air density is taken as the difference between 61.6 psia and the design DW pressure of 62 psig or 76.7 psia. A summary of parameters for MSIV leakage rate reduction for both units is shown in Table 17.

As shown in Table 16 and Table 17, the containment and MSIV leakage rates may be reduced to 60% of the initial values at 24 hr and to 50% at 72 hr.

MSIV Leakage The mass flow rate of 100 scfh (maximum allowable MSIV leakage at reduced test pressure) is converted to a true volumetric flow rate (cfh or cfm) for the appropriate conditions:

DW to MSL 100 scfh = 49.7 cfh = 0.828 cfm MSL to Condenser/HPT 100 scfh = 263 cfh = 4.38 cfm The methodology used in performing this conversion is as follows:

1. Determine an orifice size corresponding to MSIV leakage under test conditions

2. Determine a volumetric leak rate per unit of orifice surface area under accident conditions

3. Multiply the results of I and 2 above to determine volumetric leak rate under accident conditions For MSIV testing conducted at peak containment pressure or greater, the MSIV allowable leakage limit is 144 scfh for both units. The conversion of the 144 scfh to a true volumetric flow rate (cfh or cfm) for the appropriate conditions results in the same (or lower) cfh and cfm results as those given above for the reduced pressure testing case. That is, the leak path size corresponding to 100 scfh at reduced test pressure is equal to or larger than the leak path size corresponding to 144 scfh at peak containment pressure.

Enclosure I Page 26 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment The flow out of the MSL is apportioned between the condenser and the HPT based on the bypass fraction of 0.005. Although the entering flow isslightly less than 100 scfh, the flow out of the condenser is assumed to be 100 scfh. The flow rate out of the HPT is assumed to be equal to the entering flow. The leak rates are reduced by 40% at 24 hr and 50% at 72 hr. The flows are summarized in Table 18.

MCR Dose due to Airborne Activity in Turbine Building In calculating doses within regions, LocaDose only considers activities within the region. Although the MCR has 2 ft thick concrete walls, the radiation shine dose from the airborne activity within the TB could be significant because the MCR is located in the TB. The Shield-SG computer program is used to calculate the MCR dose due to TB activity.

A dose of 2.38E-03 rem is calculated for a conservatively low value of TB exhaust rate, and therefore the highest TB activity. This external shine dose from TB activity is conservatively applied to the MCR dose evaluation.

MCR Dose Due to Other ExternalSources In addition to the dose contributions from the MCR air, ingress and egress through the TB, and TB air, the shine from other external sources is evaluated. Other external shine sources considered are secondary containment, the cloud outside the TB, MSLs, condenser, and MCR filters. Analysis has determined that the shine dose from these sources is estimated to be 0.03 rem TEDE.

MCR Ingress/EgressDose Since the MCR is located in the TB, the MCR operator would walk through the TB when entering or leaving the MCR. A conservative maximum walking distance through the TB is estimated from TB dimensions. Using this distance, the transit time through the TB is estimated based on a walking speed of 3 miles/hr. An additional time of 45 sec for using stairwells and opening doors is added, and the total transit time is assumed to be 2 min.

The ingress/egress dose is calculated by determining the average TB dose rate during a time interval and multiplying by the exposure duration. Assuming two one-way trips per day, the doses from all trips over the 30-day duration of the accident are added to obtain the total ingress/egress dose.

2.5.2.3 Results Post-accident doses are the result of the following activity considerations:

1. Primary to secondary containment leakage - Activity is released directly into secondary containment and filtered by the SGTS prior to elevated release through the plant stack, except for a small amount that bypasses the SGTS. Prior to RB drawdown to negative pressure, this activity is assumed to be released directly to the environment.

2. MSIV leakage - Leakage through the MSIVs reaches the condenser through the steam lines, except for a small percentage that bypasses the condenser. The condenser is then assumed to leak to the TB and eventually to the MCR and the environment.

3. ESF leakage - This leakage occurs in secondary containment after the DW sprays have been initiated and water from the torus is recirculated back into the DW.

4. Shine from the radioactive cloud in the TB.

Enclosure I Page 27 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment

5. Shine from other external sources.

The results of the radiological consequences of a LOCA for offsite, TSC, and MCR are as follows:

• Offsite doses - The EAB and LPZ doses for the ground release pathway and elevated release pathway are shown in Table 19.

• MCR doses - The doses to occupants of the MCR are shown in Table 20.

" TSC doses - The dose to an individual inside the TSC is shown in Table 21.

In all cases, the doses are within regulatory limits.

Table 8. Core Release Rates Release Fraction Release Rate (Frac/hr)

Early Early Group Elements Gap In-Vessel Gap In-Vessel Halogens I,Br 0.05 0.25 0.1 0.167 Noble Gases Kr, Xe 0.05 0.95 0.1 0.633 Alkali Metals Cs, Rb 0.05 0.20 0.1 0.133 Tellurium Metals Sb, Se, Te 0 0.05 0 0.0333 Barium, Strontium Ba, Sr 0 0.02 0 0.0133 Noble Metals Co, Mo, Pd, Rh, Ru, Tc 0 0.0025 0 0.00167 Lanthanides Am, Cm, Eu, La, Nb, Nd, 0 0.0002 0 0.000133 Pm, Pr, Sm, Y,Zr Cerium Group Ce, Np, Pu 0 0.0005 0 0.000333 Release Duration (hr) 0.5 1.5 Note: Release rate is obtained by dividing the release fraction by the release duration.

Table 9. Flow Rates from DW to Torus Time (hr) Flow rate From To (cf m) 0 2.03 0 2.03 2.06 26457 2.06 2.39 685 2.39 3.00 349 3.00 720 0 Page 28 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Table 10. DW Activity Removal Rates Time (hr)

From To A (hW1) 0 0.250 0.0400 0.250 0.533 12.4 0.533 0.595 23.7 0.595 2.03 16.7 2.03 2.34 4.72 2.34 2.67 3.69 2.67 3.10 2.87 3.10 3.89 2.22 3.89 7.38 1.67 7.38 720 0 Table 11. DW Pressure Used in Containment Activity Removal Model Time Frame Pressure (accident hr) (psia) 0-0.250 44.7 0.250 - 2.033 18.1 2.033 - 24.00 25.3 Table 12. DW Temperature Used in Containment Activity Removal Model Time Frame Temp.

(accident hr) (OF) 0-0.250 343 0.250 - 2.033 152 2.033 - 24.00 92 Page 29 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Table 13. DW Steam Mole Fraction Used in Containment Activity Removal Model Time Frame Steam Mole (accident hr) Fraction 0-0.250 0.95 0.250 - 2.033 0.11 2.033 - 24.00 0.02 Table 14. Main Steam Line Deposition Rates Time (hr)

From To A (hV')

0 2.56 0.669 2.56 3.77 0.541 3.77 5.44 0.435 5.44 8.05 0.346 8.05 12.0 0.276 12.0 18.8 0.223 18.8 24.1 0.199 24.1 720 0 Table 15. Condenser Deposition Rates Time (hr).

From F To A (hW1) 0.0333 5.51 0.119 5.51 13.8 0.0949 13.8 24.0 0.0780 24.0 720 0.0724 Page 30 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Table 16. Containment Leakage Reduction Parameters Unit I Unit 2 Post-LOCA Time (hr) Post-LOCA Time (hr) 0 j24 172 0 24 1 72 DW Pressure (psia) 65.5 25.4 20.9 62.0 25.4 20.9 DW Temperature (OF) 293 188 160 292 188 160 Flow Ratio 1.00 0.60 0.48 1.00 0.60 0.48 Table 17. MSIV Leakage Reduction Parameters Post-LOCA Time (hr) 0 24 72 DW Pressure (psia) 76.7 25.4 20.9 DW Temperature (OF) 340 188 160 Flow Ratio 1.00 0.59 0.47

Enclosure I Page 31 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Table 18. Containment and MSIV Leakage Rates Leakage Leakage Pathway Flow Rate (cfm)

Source From To 2 min 24 hr 72 hr DW Condenser or Environment 0.0244 0.0146 0.0122 DW RB 1.20 0.720 0.600 Containment Leakage Torus Air Condenser or Environment 0.0183 0.0110 0.00915 Torus Air RB 0.899 0.539 0.450 DW MSL 0.828 0.497 0.414 MSL Condenser/ HPT 4.38 2.63 2.19 MSIV MSL Condenser 4.36 2.62 2.18 Leakage MSL HPT 0.0219 0.0132 0.0110 Condenser TB or Environment 2.29 1.37 1.15 HPT TB or Environment 0.0219 0.0132 0.0110 Table 19. Offsite Doses for LOCA Release Dose (rem TEDE)

Pathway EAB J LPZ I Limit Ground 0.307 0.644 Elevated 0.033 0.110 Total 0.34 0.75 25

Enclosure I Page 32 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Table 20. MCR Doses for LOCA Dose (rem TEDE)

MCR air 4.32 TB air (external shine) 0.0024 Ingress/egress (through TB) 0.57 Other external shine sources 0.03 Total 4.9 Limit 5 Table 21. TSC Dose for LOCA Dose (rem TEDE)

TSC air 3.9 Limit 5 2.5.3 Fuel Handling Accident (FHA)

The radiological dose consequences to occupants of the MCR and the TSC and to persons located at the EAB and LPZ following a postulated FHA are determined using AST assumptions and methodology in accordance with RG 1.183.

Two cases of radioactivity release paths from the RB are evaluated. The first case considers that the SGTS is in operation, so that all releases are filtered and elevated after drawdown of the RB. The second case does not take credit for operation of the SGTS.

2.5.3.1 Inputs and Assumptions Iodine Species in Pool The iodine released from the fuel into the pool is assumed to be composed of 99.85% elemental and 0.15% organic species, per RG 1.1 83, Appendix B, Section 2.

Activity Released The FHA is postulated to occur at the earliest possible time of fuel movement following shutdown from full power. The fuel is assumed to decay for 24 hr. The source terms are listed in Table 22.

Fuel Quantity There are 560 fuel bundles in the core. Each GE14 (lOx 10) fuel bundle contains an average of 87.3 fuel rods.

Enclosure I Page 33 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Fuel Damage The FHA is estimated to result in 172 damaged fuel rods.

Release Fractions As indicated in Appendix B of RG 1.183, all the gap activity in the damaged rods is assumed to be instantaneously released. Radionuclides that are considered include xenons, kryptons, halogens, cesiums, and rubidiums. The fractions of fission product inventory in the gap are 1-131 (0.08), other halogens (0.05), Kr-85 (0.10), other noble gases (0.05), and alkali metals (0.12) (RG 1.183, Table 3).

The release fractions are applied in Table 22 to determine the radioactivity released from the damaged fuel.

Radial Peaking Factor The maximum core radial peaking factor of 1.5 is applied to all of the damaged fuel rods.

Water Depth The minimum depth of water above the damaged fuel is 21 ft. Since this is less than the 23 ft depth assumed in the RG 1.183 derivation of the iodine DF, the reduced iodine DF is calculated in this analysis.

Pool DecontaminationFactors The retention of noble gases in the water in the spent fuel pool is negligible (i.e., DF of 1).

Particulate radionuclides are assumed to be retained by the water in the pool (i.e., infinite DF)

(RG 1.183, Appendix B, Section 3).

Secondary Containment The radioactive material that escapes from the pool is released to the environment over a 2-hr time period (RG 1.183, Appendix B, Section 5.3). Two cases of secondary containment isolation are considered. The first case assumes that the secondary containment isolates automatically in the event of a FHA. The second case assumes that isolation does not occur.

Case I The time necessary to drawdown the secondary containment (establish a 0.20 in wg negative pressure) is 120 sec. Prior to that time, airborne releases are assumed to be unfiltered at ground level.

After drawdown, all of the airborne activity is collected by the SGTS and released. The release is filtered (95% efficient filters for iodines and particulates) and elevated.

Case 2 No credit is taken for secondary containment isolation or operation of the SGTS. The release is assumed to be unfiltered at ground level for the duration of the accident.

MCR UnfilteredInleakage The unfiltered inleakage for the MCR is conservatively assumed to be 10,000 cfm, which is the same unfiltered inleakage value assumed for the TSC.

Enclosure I Page 34 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Accident Duration MCR and TSC doses are calculated for a 30-day period, since radioactivity that is brought into those rooms during the first 2 hr of the accident will continue to expose occupants until it is removed by air transfer or decay. Offsite doses are calculated for a 2-hr period, since the radioactive cloud is assumed to move past these locations during this time interval and no further exposure occurs (RG 1.183, Table 6).

Atmospheric Dispersion MCR, TSC, and offsite X/Q values are listed in Table 4 and Table 5. The MCR values are applied at the TSC since the MCR values are bounding.

2.5.3.2 Method of Evaluation Two cases of radioactivity release paths from the RB are evaluated. The first case considers that the SGTS is in operation, so that all releases are filtered and elevated after drawdown of the RB. The second case does not take credit for the SGTS.

Decontamination Factor RG 1.183 allows the use of an overall DF of 200 if the depth of water over the damaged fuel is 23 ft or greater (Appendix B, Section 2). If the depth of water is not 23 ft, the DF is determined on a case-by-case method.

For HNP, the minimum requirement for water level above the spent fuel is 21 ft. Thus, the DF as a function of water depth is derived so that an appropriately conservative value is applied to the iodine activity released from the fuel.

The overall effective DF represents a composite for the different iodine species. RG 1.183 stipulates an iodine species split of 99.85% elemental (inorganic) iodine and 0.15% organic iodine. The DF for organic iodine is 1, which means that organic iodine is not retained in water.

Using the DF values from RG 1.183 of 500 for inorganic iodine and I for organic iodine, the overall effective DF is 286. Since RG 1. 183 allows the use of an overall DF of 200, the factor of conservatism is therefore 286/200, or 1.43. This factor is applied to the effective DFs calculated as a function of water depth and is shown in Table 23. The last column shows the adjusted DF values.

For the FHA analysis, an overall effective DF of 142 (corresponding to 21 ft of water above the damaged fuel) is used.

2.5.3.3 Results The doses to MCR and TSC occupants and to persons at the EAB and LPZ are listed in Table 24. All doses are below the respective acceptance criteria.

Based on the above results, the following conclusions may be drawn:

• The MCR and TSC can tolerate significant unfiltered inleakage during a FHA.

" The SGTS is not required to meet the dose criteria following a postulated fuel handling accident.

Enclosure I Page 35 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Table 22. Activity Releases from FHA Element Isotope Core Source Gap 2 Release from Water Release Term at t=1 Fraction2 Fuel (Ci)3 DF from Water4 day (Ci/MWt)' I (Ci)

Xenon Xe-133 5.10E+04 0.05 4.17E+04 1 4.17E+04 Xe-135 1.41 E+04 0.05 1.15E+04 1 1.15E+04 Xe-137 0.00E+00 0.05 0.00E+00 1 0.00E+00 Xe-138 0.OOE+00 0.05 0.00E+00 1 0.00E+00 Xe-131m 3.02E+02 0.05 2.47E+02 1 2.47E+02 CO Xe-133m 1.45E+03 0.05 1.19E+03 1 1.19E+03

___ Xe-1 35m 6.68E+02 0.05 5.46E+02 1 5.46E+02 Krypton Kr-85 3.78E+02 0.1 6.18E+02 1 6.18E+02 Z Kr-87 2.78E-02 0.05 2.27E-02 1 2.27E-02 Kr-88 5.31 E+01 0.05 4.34E+01 1 4.34E+01 Kr-89 0.00E+00 0.05 0.00E+00 1 0.00E+00 Kr-83m 1.31E+01 0.05 1.07E+01 1 1.07E+01 Kr-85m 1.71 E+02 0.05 1.40E+02 1 1.40E+02 Iodine 1-129 1.23E-03 0.05 1.01 E-03 142 7.08E-06 1-130 2.73E+02 0.05 2.23E+02 142 1.57E+00 1-131 2.52E+04 0.08 3.30E+04 142 2.32E+02 1-132 3.21E+04 0.05 2.63E+04 142 1.85E+02 1-133 2.54E+04 0.05 2.08E+04 142 1.46E+02 1-134 1.35E-03 0.05 1.10E-03 142 7.78E-06 M 1-135 4.17E+03 0.05 3.41 E+03 142 2.40E+01 1-136 0.00E+00 0.05 0.00E+00 142 0.00E+00 1-137 0.00E+00 0.05 0.00E+00 142 0.001+00

• 1-137 0.002+00 0.05 0.00E+00 142 0.0012+00 1-138 0.OOE+00 0.05 0.002+00 142 0.002+00 Bromine Br-82 1.13E+02 0.05 9.24E+01 142 6.512E-01 Br-83 3.37E+00 0.05 2.76E+00 142 1.94E-02 Br-84 1.48E-10 0.05 1.21E-10 142 8.53E-13 Br-85 0.OOE+00 0.05 0.OOE+00 142 0.001+00 Br-87 0.00E+00 0.05 0.OOE+00 142 0.00E+00 Br-88 0.00E+00 0.05 0.00E+00 142 0.002+00 Cesium Cs-134 6.82E+03 0.12 1.34E+04 Infinite O.OOE+00 Cs-135 2.36E-02 0.12 4.63E-02 Infinite 0.OOE+00 Cs-136 2.07E+03 0.12 4.06E+03 Infinite 0.00E+00 Cs-137 4.14E+03 0.12 8.13E+03 Infinite 0.001+00 Cs-138 2.97E-09 0.12 5.82E-09 Infinite 0.002+00 Cs-139 0.00E+00 0.12 0.00E+00 Infinite 0.001+00 Cs-134m 5.33E+00 0.12 1.05E+01 Infinite 0.OOE+00 Rubidium Rb-86 6.80E+01 0.12 1.33E+02 Infinite 0.00E+00 Rb-87 0.OOE+00 0.12 0.00E+00 Infinite 0.00E+00 Rb-88 5.93E+01 0.12 1.16E+02 Infinite 0.002+00 Rb-89 0.00E+00 0.12 0.00E+00 Infinite 0.002+00 Rb-90 0.00E+00 0.12 0.00E+00 Infinite 0.002+00 Total 1.69E+05 1.66E+05 5.66E+04 Notes:

1. Core inventory at t=1 day

2. Gap fractions from RG 1.183, Table 3

3. Release from fuel is core inventory times power level times fraction of core damaged times radial peaking factor times gap fraction times 10% margin

4. Release from water is release from fuel divided by water DF

Enclosure I Page 36 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Table 23. Decontamination Factors as a Function of Water Depth Water DF DF DF Adjusted Depth (ft) Inorganic Organic Effective DF 23 500 1 286 200 22.5 437 1 264 185 22 382 1 243 170 21.5 333 1 222 156 21 291 1 203 142 20.5 254 1 184 129 Table 24. Doses from FHA Dose (rem TEDE)

Case SGTS1: Case No SGTS2: Limit MCR 0.72 3.5 5.0 TSC 0.80 3.9 5.0 EAB 0.25 1.2 6.3 LPZ 0.25 1.2 6.3 2.5.4 Control Rod Drop Accident (CRDA)

The radiological dose consequences to occupants of the MCR and the TSC and to persons located at the EAB and LPZ following a postulated CRDA are determined using AST assumptions and methodology in accordance with RG 1.183.

Conservative estimates for the source of unfiltered inleakage are assumed for each dose receptor. For the MCR, the inleakage is assumed to come from the TB. Ingress and egress doses to MCR operators passing through the TB are included in the total dose. For the TSC, inleakage is assumed to come from the environment.

2.5.4.1 Inputs and Assumptions Fuel Quantity There are 560 fuel bundles in the core. Each GE14 (IOxlO) fuel bundle contains an average of 87.3 fuel rods.

Fuel Damage The control rod drop is estimated to result in 1,200 damaged fuel rods. Of those, 1,189 rods

Enclosure I Page 37 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment experience cladding failure and 11 rods experience melting.

Cladding Failure:

As indicated in Appendix C of RG 1.183, the activity in the gap of the fuel rods is assumed to be 10%

of the core inventory for noble gases and iodines. In addition, Table 3 of RG 1.183 lists fission product inventories in the gap of other groups of elements for non-LOCA events. Although it is not specifically stated that these additional elements should be included in the CRDA, Appendix C of RG 1.183 refers to "remaining radionuclides" and "particulate radionuclides." For completeness, these additional elements are conservatively included in the radiological analysis. Thus, 5% of halogens (other than iodine) and 12% of alkali metals (cesium and rubidium) are also assumed to be in the gap of the fuel rods. All of the gap activity of the 1,189 breached fuel rods is assumed to be released as a result of the accident.

Melting:

As indicated in Appendix C of RG 1.183, 100% of the noble gases and 50% of the iodines contained in the 11 fuel rods that melt are released to the reactor coolant. In addition, Table I of RG 1.183 lists fission product release fractions for other groups of elements for design basis LOCA events.

Although it is not specifically stated that these activities should be included in the CRDA, Appendix C of RG 1.183 refers to "remaining radionuclides" and "particulate radionuclides." For completeness, these additional elements are conservatively included in the radiological analysis for the rods that melt. Therefore, 30% of halogens (other than iodines), 25% of alkali metals (cesium and rubidium), 5% of tellurium, 2% of barium and strontium, 0.25% of noble metals, 0.05% of cerium, and 0.02% of lanthanides are also assumed to be released from the I1 melted fuel rods.

Radial Peaking Factor The maximum core radial peaking factor of 1.5 is applied to all of the damaged fuel rods.

Mixing in Vessel All of the activity released from the fuel is instantaneously mixed in the reactor coolant within the pressure vessel (RG 1.183, Appendix C, Section 3.1).

Removal in Vessel No credit is taken for partitioning in the pressure vessel or for removal by the steam separators (RG 1.183, Appendix C, Section 3.2).

Vessel Release Percentages Of the activity released from the reactor coolant within the pressure vessel, 100% of the noble gases, 10% of the iodine, and 1% of the remaining radionuclides are assumed to reach the turbine and condensers (RG 1.183, Appendix C, Section 3.3).

Turbine/CondenserRelease Percentages Of the activity that reaches the turbine and condenser, 100% of the noble gases, 10% of the iodine, and 1% of the particulate radionuclides are available for release to the environment (RG 1.183, Appendix C, Section 3.4).

Enclosure 1 Page 38 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Iodine Species of Release The species of iodine released from the turbine/condenser is assumed to be 97% elemental and 3%

organic (RG 1.183, Appendix C, Section 3.6). It is noted that RG 1.183, Appendix C, Section 3.6 also states that the iodine species released from the reactor coolant within the pressure vessel should be assumed to be 95% CsI as an aerosol, 4.85% elemental, and 0.15% organic. However, in this analysis no iodine species specific .mechanisms act upon the iodine prior to release from the turbine/condenser so the iodine species percentages are applied at the time of release from the fuel.

Thrbine/CondenserLeak Rate The turbine/condenser leaks at a rate of 1% per day for a period of 24 hr, at which time the leakage is assumed to terminate (RG 1.183, Appendix C, Section 3.4).

Mechanical Vacuum Putmp For offsite and TSC doses, a second release path is evaluated to address the possibility of a forced flow path from the turbine or condenser (RG 1.183, Appendix C, Section 3.4, footnote 2). The mechanical vacuum pump normally discharges to the plant stack through the gland-seal holdup line, which provides holdup for up to 2 min, but no filtration. The pump trips on high MSL radiation, but the release is conservatively assumed to continue for 24 hr. The mechanical vacuum pump flow is assumed to be 2,200 cfm for 24 hr, at which time the release is assumed to terminate.

Accident Duration MCR and TSC doses are calculated for a 30-day period, since radioactivity that is brought into those rooms during the first 24 hr of the accident will continue to expose occupants until it is removed by air transfer or decay. Also, unfiltered inleakage to the MCR from the contaminated TB is assumed to continue for the 30-day accident duration. EAB doses are calculated for a 2-hr period. LPZ doses are calculated for a 24-hr period, since the radioactive cloud is assumed to move past these locations during this time interval and no further exposure occurs.

Atmospheric Dispersion Design basis X/Q values for the MCR, TSC, and offsite are listed in Table 4 and Table 5. For MCR, the release point is assumed to be the RB vent. For the EAB and LPZ, ground level release is assumed. The MCR values are also applied at the TSC since the MCR values are bounding. Since the release terminates at 24 hr, X/Q values are not used at the LPZ after that time.

For the case of forced flow from the turbine/condenser, the release is elevated through the main stack.

Dose in MCRfromn External Sources Since the MCR walls are 2 ft concrete and the roof is 2.5 ft concrete, the dose due to external airborne activity is assumed to be negligible compared to the dose received from activity within the MCR.

The LOCA analysis demonstrates that the MCR dose due to TB activity is negligible compared to that due to activity within the MCR.

2.5.4.2 Method of Evaluation Doses are calculated using the guidance provided in the main body and Appendix C of RG 1.183.

LocaDose is used to calculate the dose in the MCR and TSC and at the EAB and LPZ. The two Page 39 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment LocaDose modules utilized in this analysis are Activity Transport Program and Dose Calculation Program. A description of LocaDose modules is given in Section 2.2.1.

Transportof Radioactivityand Doses Two models for the transport of radioactivity from the turbine/condenser to the MCR and to the environment are considered, to conservatively maximize the dose to each.

The first model considers holdup of radioactivity in the TB so that MCR doses are calculated conservatively, since the MCR is located in the TB. The TB exhausts to the environment starting at 9 hr with a flow of 15,000 cfm via the RB vent. Ingress and egress doses are also calculated for MCR operators who pass through the TB to enter and exit the MCR.

The second model does not consider holdup of radioactivity within the TB so that doses to occupants of the TSC and persons at the EAB and LPZ are calculated conservatively. Leakage from the turbine/condenser is released directly to the environment at ground level.

MCR Ingress/EgressDoses The total dose to MCR operators is the sum of the dose received while occupying the MCR for 30 days plus the dose received while traveling to and from the MCR through the TB. The methodology for calculating ingiress/egress dose is the same as that used in the dose calculations for the LOCA. This methodology is explained in Section 2.5.2.2.

2.5.4.3 Results Table 25 shows that the doses to persons at the EAB and the LPZ are below the acceptance criteria for a postulated CRDA. Table 26 shows that the dose to MCR personnel is within the regulatory limit. Table 27 shows that the TSC dose is well within the acceptance criterion.

Table 25. Offsite Doses from CRDA Dose (rem TEDE)

EAB LPZ Limit Design Basis 0.047 0.094 6.3 SForced Flow 0.333 0.540 6.3

Enclosure I Page 40 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Table 26. MCR Doses from CRDA Dose (rem TEDE)

MCR air 3.61 Ingress / Egress 0.23 Total 3.8 Regulatory Limit 5 Table 27. TSC Doses from CRDA Dose (rem TEDE)

TSC Limit Design Basis 0.32 5 Forced Flow 0.81 5 2.5.5 Main Steam Line Break (MSLB)

This analysis calculates the radiological dose consequences to occupants of the MCR and the TSC and to persons at the EAB and LPZ following a postulated MSLB using AST assumptions and methodology in accordance with RG 1.183.

A maximum primary coolant iodine concentration of 2.0 ItCi/g DE 1-13 1, which reflects a proposed change in the applicable TS, is assumed.

2.5.5.1 Inputs and Assumptions Fuel Damage The temperature and pressure transients resulting from this event are not severe enough to cause fuel damage.

PrimaryCoolant Iodine Activity during Accident For the case of a pre-accident spike, the iodine concentration in the primary coolant is assumed to be 2.0 gCi/g DE 1-131. For the case of maximum equilibrium value, the iodine concentration in the primary coolant is assumed to be 0.2 ItCi/g DE 1-131, the value allowed by the TSs for continued full power operation.

PrimaryCoolant Iodine Activities during Normal Operation During normal operation, the relative distribution of iodine isotopes in the primary coolant is given in Table 28.

Enclosure I Page 41 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Noble Gas Release Rate The noble gas release rates from the core are shown in Table 29 The values in Table 29 correspond to a decayed total release rate of 0.3 Ci/sec after a delay of 30 min for the offgas system. The TS limit for the delayed offgas release rate is 0.240 Ci/sec. Hence, the noble gas release rates in Table 29 bound the value in the TS. The noble gas release rates from the break are assumed to be the same as from the core as shown in Table 29, assuming no decay.

MSIV Closure Time The high-flow signal from the break is assumed to initiate MSIV closure within 0.5 sec of the break.

The maximum time required to isolate the MSIV is 5 sec. Therefore, total time from the break to MSIV isolation is assumed to be 5.5 sec.

Fluid Releasefrom ReactorCoolant The steam that is released expands due to the lower atmospheric pressure and becomes superheated.

The initial mass of the saturated liquid that is released equals the mass of saturated liquid and vapor at atmospheric pressure (14.7 psia) and 212'F. Enthalpy is constant through the phase change as part of the released saturated liquid flashes to vapor.

Steam Blowdown The blowdown rate through the break is given in Table 30.

Mixture Quality A mixture quality of 7% is assumed for the mixture portion of the blowdown.

Turbine Building In calculating the MCR dose, it is assumed that the activity released from the break is uniformly mixed within the total TB free volume. This is reasonable given the force with which the steam is released from the break.

Dose in MCR from External Sources Since the MCR walls are 2 ft concrete and the roof is 2.5 ft concrete, the dose due to external airborne activity is assumed to be negligible compared to the dose received from activity within the MCR.

The LOCA analysis demonstrates that the MCR dose due to TB activity is negligible compared to that due to activity within the MCR.

Atmospheric Dispersion MCR atmospheric dispersion factors (X/Q) for a ground level release through the RB vent are shown in Table 4. The MCR values are applied to the TSC since the MCR values are bounding. The EAB and LPZ X/Q values for a ground level release are shown in Table 5.

Accident Duration The release from the break is assumed to be an instantaneous puff. MCR, TSC, and LPZ doses are calculated assuming an exposure time of 30 days. The EAB dose is calculated over a 2-hr period which yields the maximum dose which, for a puff release, is the first 2 hr.

Enclosure F Page 42 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment-2.5.5.2 Method of Evaluation Doses are calculated using the Activity Transport Program and Dose Calculation Program modules of LocaDose. The following sequence of events is assumed to occur:

I. Before the accident, the reactor is assumed to be in hot standby mode to maximize the inventory lost through the break prior to isolation.

2. One of the MSLs outside the RB is completely severed, releasing steam.

3. Within 0.5 sec, a high-flow signal initiates MSIV closure.

4. Rapid depressurization of the reactor pressure vessel causes the water level to rise, releasing a steam-water mixture from the break.

5. The reactor scrams.

6. Within the maximum time allowed by the TSs, the MSIVs are fully closed, terminating the release.

7. The total mass of coolant released is that amount in the steam line and connecting lines at the time of the break plus the amount that passes through the valves prior to closure.

8. The released steam forms a large cloud.

The mass released includes the steam originally in the line as well as from a portion of the saturated liquid which spills from the break and flashes to steam. The flashing fraction of the released liquid assumes a constant enthalpy process.

Offsite Dose Analysis For the dose calculations for persons at the EAB and LPZ, it is assumed that the accident damages the TB such that it is not able to contain the steam that is released from the break. The steam is released to the environment as a puff, resulting in offsite doses. Two scenarios are evaluated. In the first case, the primary coolant iodine concentration corresponds to a pre-accident spike. In the second case, the primary coolant iodine concentration corresponds to an equilibrium value.

MCR Dose Analysis For doses to occupants of the MCR, it is assumed that the TB contains the steam that is released from the break. This contained source is available for direct leakage into the MCR. The primary coolant iodine concentration corresponds to a pre-accident spike, which bounds the equilibrium value case.

TSC Dose Analysis For doses to occupants of the TSC, the model used is the same as that used for calculating EAB and LPZ doses. The primary coolant iodine concentration corresponds to a pre-accident spike, which bounds the equilibrium value case.

MCR Ingress/Egress MCR ingress/egress doses apply to MCR operators as they walk across the TB deck when entering and leaving the MCR. These doses are calculated by determining the average TB dose rate during a time interval and multiplying by the exposure duration. The methodology for calculating ingress/egress dose

Enclosure I Page 43 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment is the same as that used in the dose calculations for the LOCA. This methodology is explained in Section 2.5.2.2.

Impact of Cesium An evaluation is performed to determine the impact of cesium released during the MSLB on MCR and offsite doses. The following cesium isotopes are considered: Cs-134, Cs-136, Cs-137, and Cs-138. For MCR doses, the cesium dose contribution represents a very small fraction of the dose received from iodine and noble gas isotopes. This difference is considered to be .within the accuracy of the MSLB calculation. For offsite doses, the cesium contribution represents a relatively larger fraction of the total dose since offsite releases do not have the benefit of particulate filters; however, the total offsite dose, including the cesium dose contribution, is less than one percent of the regulatory limit. It is therefore concluded that cesium has an insignificant impact on both MCR and offsite doses.

2.5.5.3 Results The EAB and LPZ doses are shown in Table 31 for both cases analyzed. The MCR doses are shown in Table 32. The TSC doses are shown in Table 33. For all cases analyzed, doses are within the regulatory limits.

Table 28. Primary Coolant Iodine Activities Io Primary Coolant Isotope j Activity (pCi/g) 1-131 0.018 1-132 0.16 1-133 0.12 1-134 0.31 1-135 0.17 Total 0.778

Enclosure I Page 44 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Table 29. Noble Gas Release Rates Isotope Release Rate (Ci/sec)

Kr-83m 1.02E-02 Kr-85m 1.83E-02 Kr-85 6.00E-05 Kr-87 6.00E-02 Kr-88 6.00E-02 Kr-89 3.90E-01 Xe-131m 4.50E-05 Xe-133m 8.70E-04 Xe-1 33 2.46E-02 Xe-135m 7.80E-02 Xe-1 35 6.60E-02 Xe-1 37 4.50E-01 Xe-138 2.67E-01 Total 1.43E+00 Table 30. Steam Blowdown Rate Time After Mass Flow Enthalpy Break (sec) (Ibm/sec) (Btu/Ibm) 0 5,300 1,191.5 2.75 4,500 1,191.5 2.76 19,600 589.3 4.0 19,500 589.5 5.5 0 589.5 Table 31. Offsite Doses from MSLB DE 1-131 Dose (rem TEDE)

Scenario A(pCivg) EAB LPZ Limit Pre-accident spike 2.0 0.15 0.1 5 25 Equilibrium iodine activity 0.2 0.015 0.015 2.5 Page 45 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Table 32. MCR Doses from MSLB Dose (rem TEDE)

MCR air 3.70 Ingress / Egress 0.22 Total 3.9 Regulatory limit 5 Table 33. TSC Doses from MSLB Dose (rem TEDE)

TSC air 0.43 Regulatory limit 5 2.6 Suppression Pool pH Control Suppression pool water will retain soluble gases and soluble fission products such as iodine and cesium.

Once deposited, the iodine will remain in solution as long as the suppression pool pH is maintained at or above 7 (RG 1.1 83, Appendix A, Section 2). The pH of the suppression pool water is calculated as a function of time to demonstrate that the pH remains at or above 7 for the duration of the DBA LOCA.

2.6.1 Inputs and Assumptions Core FissionProduct Inventory For radiolysis of water and the production of nitric acid, the fission product and actinide decay power is assumed to be a function of fission product/actinide mass grouped into eight categories per Table 3.4 of Reference 4. This power specification is inherently conservative since it was developed for relatively low burnup, and as burnup increases, the power per unit mass of fission products/actinides decreases. The group fission product/actinide mass inventories used for HNP are based on values for BWRs of similar thermal power with relatively high burnup (making the power specification very conservative), including a multiplication factor of 1.1 for additional conservatism.

For radiolysis of cable and the production of hydrochloric acid, the power specification is based on a conservative activity inventory and associated gamma and beta MeV/sec/MWt that is generic to the STARpH methodology (see Section 2.6.2).

Aerosol Fraction The fraction of aerosol in the source term in the suppression pool is 0.90. Given that the spray will tend to wash any aerosol which deposits on elevated surfaces into the sump, the actual fraction of aerosol in the water pool is expected to be essentially 100%. Thus use of 90% is conservative since it will overestimate the radiation level in the DW vapor space and thus overestimate the concentration of hydrochloric acid, [HCI], from radiolysis of chloride-bearing cable insulation.

Enclosure I Page 46 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment OrganicAcid fromn Paints Organic acid from paints can be neglected. The hydrogen ion concentration, [H+], from the production of organic acid in the suppression pool is expected to be a small fraction of the total [H+]

from the nitric and hydrochloric acid calculated to be produced by the Radiolysis of Water and Radiolysis of Cable models of the STARpH code.

Sodium PentaborateAddition The SLC system is actuated and the SPB solution is injected into the pool within several hours of accident initiation. A core damage event large enough to release the substantial quantities of fission products in the time frame considered for the AST in RG 1.183 will be very evident to the operators (e.g., core outlet temperature, radiation level in the DW, pressure and temperature in the DW, hydrogen level in the DW) within minutes of the initiating event. Thus it is reasonable to assume the HNP emergency operating procedures provide for SLC system actuation within approximately 2 hr of accident initiation.

If SLC injection is into the pool (i.e., into the reactor vessel with the vessel communicating with the pool as in a recirculation line break), significant mixing will occur quickly, on the order of I hr based on an RHR flow rate of about 10,000 gpm and pool volume of 7E+05 gallons.

If the reactor vessel is not immediately communicating with the pool, an additional few hours is assumed to assure communication with the pool or inject SPB to the pool via an alternate pathway.

Unbuffered Pool pH The unbuffered pH of the pool should remain above 7 for at least several hours. The acid added from radiolysis of water (HNO3) and radiolysis of cable (HCI) is not enough to neutralize the hydroxyl ion concentration, [OH-], from fission product cesium until approximately I day after accident initiation.

Pool Temperature The average temperature of the pool over 30 days is 155°F. The dissociation constant and starting pH are somewhat, but not strongly, temperature dependent. The average temperature of the pool over 30 days is calculated to be 155 0F for a BWR recently studied that operates at a 23% higher thermal power than the HNP units, but has a pool volume about 40% greater than the HNP units.

The design inputs to the pH calculation are given in Table 34.

2.6.2 Method of Evaluation The BWR version of the Radiolysis in Water model in the STARpH code calculates the hydroxyl ion concentration from fission product cesium, and nitric acid concentration in the containment water pool generated by radiolysis. The Radiolysis of Cable model in the STARpH code calculates the hydrochloric acid concentration as a result of radiolysis of electrical cable insulation. From these two calculations, the net hydrogen ion concentration added to the pool is calculated over time.

A calculation is performed to determine the amount of SPB buffer added to the pool from the SLC system. From this, the concentration of boron in the water pool is determined.

Enclosure I Page 47 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment The Add Acid model of STARpH is used to determine pH as a function of time using the [H÷] added, the concentration of Boron in the pool, the boron buffer dissociation constant, and the starting pH of the buffer solution.

The reliability of the SLC system to perform the post-LOCA function (injection of pH buffering agent) is discussed in Section 2.7.2.

2.6.3 Results The boron buffering is conservatively assumed to begin at 5 hr. Thus for times up to 5 hr, the pH is determined by the net [OH] resulting from the initial pH, HI, CsOH, and HNO 3 and the [H'] added to the pool from [HCI]. For time points 1 hr and 2 hr, pH is indicated simply as > 8.0 on the basis of [OH] from fission product cesium. From 5 hr on, the effect of cesium is neglected and pH is obtained by applying the addition of [H+]. The results are shown in Table 35.

It is determined that the calculated required quantity of SPB is met by the current TS limit. The pH of the containment water pool for the DBA LOCA is 7.7, or above, over a period of 30 days following accident initiation.

Table 34. Design Inputs for pH Calculation Input / Assumption Value Suppression Pool Volume (minimum/maximum) 85,110 ft3 / 89,670 ft3 RCS Inventory (volume in vessel and 3 recirculation loops) 9,965 ft , 18,000 Ibm steam Initial Suppression Pool pH 7.2 Electrical Cable Insulation (Hypalon) Mass 6,859 Ibm / 4,215 Ibm (Unit1 / Unit 2)

Fraction of Cable with Chloride-bearing 10%

Insulation in Conduit Unit 1 DW Free Volume 146,010 ft3 Unit 1 Minimum Pressure Suppression 112,900 ft3 Chamber Free Volume Unit 2 DW Free Volume 146,266 ft3 Unit 2 Minimum Pressure Suppression 109,800 ft3 Chamber Free Volume Mass of SPB Available for Injection 1,975 Ibm Boron Enrichment 60% Bio

Enclosure I Page 48 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Table 35. Suppression Pool pH vs.

Time pH Time Unit 1 Unit 2 1 hr >8.0 >8.0 2 hr >8.0 >8.0 5 hr 8.4 8.4 12 hr 8.3 8.4 1 day 8.3 8.3 3 day 8.2 8.2 10 day 7.9 8.1 20 day 7.8 8.0 30 day 7.7 7.9 2.7 Crediting of Non-Safety Related Systems The DBA analyses for the LOCA, FHA, CRDA, and MSLB use inputs that rely on the availability of many plant systems in order to mitigate the effects of the accidents. For the LOCA, CRDA, and MSLB, the analyses require the operation and structural integrity of a small number of systems that, although demonstrated to be highly reliable, are not safety related. These systems (including the applicable DBAs that credit them in the analyses) are:

0 MSIV ALT Pathway (LOCA) 0 SLC System (LOCA)

S TB Ventilation System (LOCA, CRDA, MSLB)

This section demonstrates the reliability of these systems in the various radiological dose consequence analyses, along with NRC-approved methodologies where applicable.

2.7.1 MSIV Alternate Leakage Treatment RG 1.183, Appendix A, Section 6, allows credit for reduction in MSIV releases due to holdup and deposition in the main steam piping downstream of the MSIVs and in the main condenser, including the treatment of air ejector effluent by offgas systems, if the components and piping systems used in the release path are capable of performing their safety function during and following a safe shutdown earthquake. Per RG 1.183, an acceptable model for evaluating reduction of MSIV releases is provided in General Electric Topical Report NEDC-31858P-A, "BWROG Report for Increasing MSIV Leakage Limits and Elimination of Leakage Control Systems" (Reference 3).

The NRC Safety Evaluation for NEDC-31858P-A (Reference 5) identified limitations to be addressed as part of a plant specific application of the ALT methodology. These limitations relate to assuring that the ALT pathway for MSIV leakage is functionally reliable commensurate with its intended safety function and assuring the pathway, including the main condenser, is seismically rugged.

Enclosure I Page 49 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment The use of MSIV ALT was previously approved for use on Unit 2 (Reference 1). The following discussion is with regard to Unit 1.

2.7.1.1 ALT Pathway Description The HNP ALT pathway for Unit I utilizes the large volume of the MSLs and the main condenser to provide holdup and plate-out of fission products that may leak through the closed MSIVs. The primary components of the ALT pathway are the main condenser, the MSLs from the MSIVs to the turbine stop and bypass valves, and the drain piping which originates downstream of the outboard MSIVs and terminates at the main condenser. The condenser forms the ultimate boundary of the ALT pathway.

Existing valves upstream of the condenser are used to establish the flow path and isolate the boundaries of the path and to limit the extent of seismic verification walkdown. The model for evaluating reduction of MSIV leakage is provided in Reference 3. Figure 1 shows a schematic of the primary and secondary ALT pathways.

The ALT pathway utilizes MSL drains to direct MSIV leakage to the main condenser. The ALT path is from the downstream side of the MSIVs through four 2-in lines which join a 3-in drain line to the main condenser. The path to the condenser is through motor operated valves (MOVs) 1B21-F020 and IB21-F021. MOV 1B21-F020 is normally open and will remain open. MOV IB21-F021 is normally closed and must be opened. Class I1E power is supplied to 1B21-F021 to assure the ability to open the valve. It will be opened by operator action from the MCR to initiate the flow path to the condenser.

Valve IB21-F019 is a normally closed MOV in the drain line upstream of IB21-F020 and 1B21-F021. It is a primary containment isolation valve and will close or remain closed to maintain the upstream boundary. MOV IB21-F038, a 2-in drain valve located upstream of IB21-F021 and downstream of 1B21-F019, is normally closed and will remain closed. As the flow path is via a 3-in line without an orifice, even in the case of loss of offsite power (LOSP), the drain path to the condenser is open and would be available.

For additional assurance that the ALT pathway boundary is isolated and the release is via the condenser, automatic and operator actions will be taken to close boundary valves downstream of the MSIVs and upstream of the condenser. In the event of a LOCA, the MSIVs, the turbine stop valves and the turbine bypass valves will automatically close. The reactor feed pump turbine stop valves, INI 1-F177 and INI 1-F178, which are hydraulically operated, close on an automatic or manual trip of the reactor feed pump turbines.

Operator action is required to isolate steam to the second stage moisture separator reheaters by closing MOVs 1N38-FIOIA and 1N38-FIO1B from the MCR. Steam to the steam jet air ejectors will be isolated by closing MOVs 1NI 1-FOO1A and 1N II-FOO1B at local panel 1H21-P216 and manual drain valves INI l-F039 and INI 1-F041 locally. The seal steam line that comes off of MSL "C" will be isolated by closing MOVs 1N33-F012 and 1N33-F013 from the MCR. Two manual drain valves, 1N1 1-F043 and INI 1-F044, on the steam lines to the reactor feed pump turbines, will be closed locally by operators.

The ALT pathway must be capable of performing its post-LOCA function during and following a DBE, assuming offsite power is not available. The valves required to be opened in order to establish the path to the condenser, and boundary valves required to be closed to establish the path boundary, are included in the plant's Inservice Inspection Program. The only active valVe required to open to establish the ALT pathway, MOV 1B21-F021, is powered from Class IE power sources and can be opened from the MCR in the event of LOSP.

Enclosure I Page 50 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment In the unlikely event that IB21-F021 fails to open, a secondary passive path with an orifice also exists. In this case, part of the flow would go through a normally open bypass with a 0.103 in diameter orifice around IB21-F021. The remainder would go to the condenser via the main steam stop and control valves before seat drain lines which contain a 0.850 in diameter restricting orifice.

2.7.1.2 ALT Boundary Seismic Evaluation The primary ALT boundary components relied upon for pressure boundary integrity are: (1) the main condenser, (2) the MSLs from the MSIVs to the turbine stop and bypass valves, and (3) the main steam turbine bypass and drain line piping to the condenser. The BWR Owners' Group (BWROG) has performed a comprehensive evaluation of the capability of similar components in actual earthquakes.

Based on this evaluation of earthquake experience data, the BWROG has developed an approach of verifying the seismic adequacy of the leakage path which is based on utilizing the earthquake experience-based methodology, supplemented by a plant-specific walkdown and analytical evaluations. This methodology is provided in General Electric proprietary report NEDC-31858P. In 1999, the NRC issued a Safety Evaluation Report (SER) (Reference 5) on the GE report. In this SER, the NRC staff stated it considers the BWROG report acceptable for use in individual plant submittals on MSIV leakage issues, subject to the conditions and limitations described in the SER.

A review and evaluation were performed for HNP Unit I following the BWROG methodology with appropriate consideration of the conditions and limitations of the NRC SER. Seismic hazard issues identified during the review and evaluation were identified and corrective actions specified. The results of the review and evaluation demonstrate, with the incorporation of the corrective actions, that the piping, supports, and equipment within the Unit I MSIV leakage control boundaries meet the appropriate acceptance criteria. The results of this review and evaluation are documented in Enclosure 8 of this submittal. Enclosure 8 provides a description of the ALT pathway including its boundaries, summary of the associated seismic evaluations, and how the criteria of the GE proprietary report NEDC-31858P and the conditions and limitations of the SER were applied.

2.7.1.3 Non-MSIV Leakage Bypass Pathway For the DBA LOCA, the design basis includes a maximum rate of containment leakage. This primary containment leakage, excluding MSIV leakage, enters the secondary containment (RB) except for 2% that is assumed to bypass the secondary containment. These lines are connected to the condenser, and thus for the evaluation of doses to occupants of the MCR, all of the secondary bypass leakage is assumed to be into the condenser. In a manner similar to the ALT pathway for MSIV leakage, the pathway for secondary containment leakage to the condenser must be able to withstand a design basis earthquake.

Seismic verification for the secondary containment bypass leakage pathway to the condenser has been completed for both units. The following bypass leakage piping is identified as being subject to the seismic verification:

• The portion of the reactor core isolation cooling (RCIC) steam drain line from outside the RB to the main condenser

• The portion of the high-pressure coolant injection (HPCI) steam drain line from outside the RB to the main condenser

• The reactor water cleanup (RWCU) blowdown line to condenser in the TB, from outside the RB to the main condenser

Enclosure I Page 51 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment All of the piping is located in the TB, primarily in the condenser bay below the operating floor. All of the identified piping provides a direct flowpath to the main condenser for secondary containment bypass leakage from containment. The piping does not have any branch lines. Therefore, there are no boundary components or isolation valves requiring seismic verification. Also, the piping does not contain any valves that must be positioned in order to provide a flowpath to the main condenser.

The reports for the seismic evaluation of the Units I and 2 bypass piping are included as Enclosures 9 and 10, respectively, of this submittal.

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Enclosure I Page 53 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment 2.7.2 Standby Liquid Control System The SLC system is a backup method of manually shutting down the reactor to the cold subcritical mode independent of the control rod system. Availability of SLC is governed by the TSs. The SLC system at HNP is considered a special safety system or safe shutdown system, and not an ESF system. Therefore, the NRC review guideline, "Guidance on the Assessment of a BWR SLC System for pH Control" is used to evaluate the SLC system for its ability to perform its AST function of post-LOCA suppression pool pH control.

Plant procedures will be revised as necessary so that upon detection of high DW radiation associated with the postulated activity release, manual initiation of SLC injection is executed for a LOCA to maintain suppression pool pH at or above 7.0.

SLC is suitably redundant in components and features to assure that its AST function can be accomplished assuming a single active failure. HNP has addressed two potential active failures that could impact the SLC system. The first potential failure is the SLC initiation control switch located on the Unit I and Unit 2 panels IHI 1-P603 and 2H1 1-P603, respectively, in the MCR. The second potential failure is one of the two check valves in series on the injection line that are credited to change state to inject the SPB solution.

The SLC initiation control switch is a key-locked, three position switch. The entire assembly is enclosed in a metal cover that provides protection for the contacts. The switch is commonly used in safety and non-safety related applications at HNP and throughout the industry. It is of simple construction with few parts vulnerable to failure. The typical mechanical service life for this switch is estimated to be in the range of 500,000 to 1,000,000 cycles. In the unlikely event of a failure of the control switch to initiate either SLC sub-system, a repair of the switch could be attempted in the MCR, considering that SLC injection is not required for the first 2 hr. An additional compensating action is the ability to install jumpers to overcome failure of the control switch. Procedures will be revised as necessary to address jumper installation for this application.

The injection line check valves are designed to open against full reactor pressure. For the AST function, there would be an even greater differential opening force on the check valves due to the depressurized reactor. Using database searches, no instances of failure to open were found for the injection line check valves. The check valves are considered highly reliable and no compensatory actions are considered necessary to address failure of the component.

Acceptable quality and reliability of the non-redundant active components and the corresponding compensatory actions in the event of failure is demonstrated for the SLC initiation control switch and the injection line check valves.

The environmental conditions for the SLC system have been evaluated with respect to the SLC post-LOCA mission. The SLC system mission time (i.e., the time at which SLC injection is complete) is approximately 6 hr post-LOCA. The post-LOCA RB environmental conditions of interest are temperature and radiation. Pressure and humidity are not environmental factors since the LOCA is in the DW. The post-LOCA RB temperature transient after switching from normal heating, ventilation, and air conditioning (HVAC) to SGTS is slow and the SLC area heat sources are relatively small. It is estimated that 150'F is a reasonable upper bound for the SLC temperature within the first 6 hr of a LOCA. This is considered a mild environment.

Enclosure I Page 54 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Evaluations are performed to determine the dose to SLC components following a LOCA. The total gamma dose, which is the accumulation of normal operating dose over the life of the plant plus 6 hr of dose from the LOCA, is less than 1.OE+04 rad. This is considered to be a mild environment for all components shielded from beta. The only component required for injection that is not shielded from beta is the SLC pump motor. The SLC pump motor has an open drip-proof type enclosure that permits limited exposure of the insulation system to the beta cloud. The total dose (gamma plus beta) to the motor insulation system is determined to be less than 1.OE+04 rad, and is therefore considered to be a mild environment. Cable that is associated with the SLC system is also evaluated and determined to be environmentally qualified for the SLC injection post-LOCA mission.

2.7.3 Turbine Building, Ventilation The MCR, as part of the control building, is located in the center of the Units I and 2 TBs. The Units I and 2 TB ventilation systems are credited in AST with purging the area around the MCR following a LOCA, CRDA, and MSLB. Although the TB ventilation system is not safety-related, a verification of the TB ventilation system is performed using a methodology similar to verifying the MSIV ALT pathway and SLC. From this verification, it can be concluded that the TB ventilation system is highly reliable, and that there exists a high degree of assurance that it will perform its intended function of purging the area around the MCR in the event of one of the abovementioned DBAs.

In addition, a defense-in-depth study on passive ventilation of the TB is conducted to determine expected exhaust flow rates in the absence of forced ventilation. Although passive ventilation is not credited in the analysis, it is shown that without forced TB exhaust, enough ventilation of the TB would exist to maintain MCR doses within regulatory limits, even with unfiltered inleakage into the MCR that is much greater than inleakage actually measured during recent testing.

2.7.3.1 Turbine Building Ventilation System Description The HNP TB ventilation system is important to operations, and is therefore required to be highly reliable.

It is designed to:

" . Provide temperature control and air movement control for personnel comfort.

• Optimize equipment performance by the removal of heat dissipated from plant equipment.

• Provide a sufficient quantity of filtered fresh air for personnel.

" Provide for air movement from areas of lesser potential airborne radioactivity to areas of greater potential airborne radioactivity prior to final exhaust.

" Minimize the possibility of exhaust air recirculation into the air intake.

• Minimize the escape of potential airborne radioactivity to the outside atmosphere during normal operation by exhausting air, through a suitable filtration system, from the areas in which a significant potential for radioactive particulates and radioactive iodine contamination exist.

For each unit, air is exhausted from the TB by a duct system to the outside environment via the RB vent plenum by two exhaust fans. The exhaust from the TB is filtered by two 50% :capacity filter trains. Each filter train consists of a bank of prefilters, carbon adsorbers, and high efficienc'y picuhite air filters. The carbon adsorber bank is provided with a water deluge system. Only one of the two 100% capacity

Enclosure I Page 55 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment exhaust fans is normally operating. If the operating exhaust fan fails, the standby exhaust fan starts automatically and an alarm is annunciated in the MCR.

The TB ventilation system incorporates redundancy and other features designed to assure TB operation for normal operation plant conditions. Such features for each unit include:

• A 100% standby supply air fan.

• A 100% standby exhaust air fan.

• Two 50% capacity normally operating charcoal filter trains.

" Provision to adjust supply and exhaust fan flowrates manually so that one filter (50% of normal airflow) can be used during filter maintenance periods.

In the event of a LOCA, CRDA, or MSLB, the appropriate operating procedures will be changed to ensure that TB exhaust ventilation (one of four TB exhaust fans) is initiated within 9 hr of the start of the accident, in accordance with the design basis assumptions for TB ventilation used in the DBA analyses.

The post-accident TB environment is evaluated for the DBAs that credit TB ventilation to ensure that the expected operating environment would not inhibit the functioning of the TB ventilation system. It is concluded that the post-accident environment in each case is a mild environment, with total dose less than 1.01E+04 rad.

2.7.3.2 Failure Analysis For the TB exhaust system to perform its post-accident function of purging the TB of activity, one of four TB exhaust fans must be able to operate, and the associated exhaust pathway must remain available. The pathway consists of ductwork and a small number of dampers.

Each unit has air-operated inlet isolation valves to the respective TB exhaust filter trains (two parallel filter trains per unit). These are normally open, fail closed valves. Possible active failure modes for these vales are loss of air and loss of power to the normally energized solenoid valve. Either loss of air or power would result in the valves failing closed and stopping flow through the 50% capacity filter trains.

Each unit utilizes air-operated variable pitch inlet dampers. These valves are controlled by normally energized solenoid valves that provide blade pitch control for the TB exhaust fans. Loss of air or power to the solenoid valves would result in closure of the associated inlet damper to allow minimum flow through the in-service TB exhaust fan.

Temperature switches are installed on each unit in the TB exhaust filter train. Upon reaching the high temperature setpoint, contacts.close to trip the TB exhaust fans. Loss of power to these switches would result in the contacts staying open, having no effect on the TB exhaust fans.

Loss of power to a single turbine building exhaust fan would result in a low flow annunciation in the MCR and the automatic start of the standby turbine building exhaust fan. Loss of power to both fans in one unit would only result in all turbine building forced exhaust flow being stopped if both exhaust fans in the other unit also failed.

Instruments are used in each TB exhaust system to detect low flow conditions. A low flow condition of the in-service TB ventilation fan would result MCR annunciation. Loss of power to these instruments

Enclosure 1 Page 56 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment would result in the loss of ability to automatically start the TB exhaust fans upon a low flow condition.

The in-service exhaust fan would continue to run.

Unit I utilizes TB exhaust fan outlet dampers. These valves are normally open, fail closed air-operated valves. Loss of air or loss of power to the normally energized solenoid valve would result in valve closure, stopping flow through the Uni I I TB exhaust filter trains. This failure mechanism is not applicable to Unit 2.

The air for all TB HVAC dampers on both units is supplied by interruptible service instrument air and a combination of three station service air compressors. Loss of instrument air can only occur as the result of one the following:

• A major line break in the compressed air system, or

• The mechanical or electrical failure of the normal instrument air supply, or

• A major dryer failure.

No single failures exist that would impact the TB exhaust capability of both units. The only failure mechanism that could affect both units is a seismic event. The air piping system supplies both safety and non-safety/non-seismic systems. A failure of the air systems of both Unit I and Unit 2 would render both TB HVAC systems incapable of performing their required exhaust functions. A modification will be completed to ensure that a loss of air event does not render both TB exhaust systems incapable of operating. This modification will be implemented by December 31, 2009.

No common power supplies exist through the start-up auxiliary transformers I D and 2D in which failure would result in the loss of the TB ventilation exhaust capability for both units. A discussion of offsite power reliability follows in Section 2.7.3.3.

The motor control center (MCC) panels utilized for TB ventilation are of a robust design. Included in the HNP program resolution to GL 87-02 for NRC Unresolved Safety Issue (USI) A-46, MCC panels 1R25-S037, 1R25-S065, and 2R25-S065 were verified as capable to function after a design basis earthquake using the Seismic Qualification Utility Group (SQUG) Generic Implementation Procedure (GIP),

Revision 2, corrected February 14, 1992, as clarified and interpreted by NRC Supplemental Safety Analysis Report No. 2. Using the same methodology, all conduits and cable raceways in the RBs, control buildings, and east cableway were verified as capable to function after a design basis earthquake.

MCC panels 1R25-S120, 1R24-S016, 2R25-S106 and 2R24-S016 were not included in the program resolution for NRC USI A-46, but are similar to the panels evaluated and are expected to perform similarly during a seismic event. The panels will be evaluated using the same SQUG methodology, beginning with walkdowns during the upcoming 2007 and 2008 outages, to verify that they are seismically adequate to withstand the appropriate DBE. The walkdowns will be completed by May 31, 2008.

2.7.3.3 Offsite Power Reliability HNP has a robust offsite power supply, consisting of four 500-kV transmission lines and four 230-kV transmission lines. A ring bus switching scheme is used for the 500-kV switchyard, and a breaker-and-a-half scheme is utilized for the 230-kV switchyard. Three physically independent 230-kV circuits are provided from the switchyard to startup auxiliary transformers IC, ID, 2C, and 2D.

Enclosure I Page 57 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Each transmission line is protected with two protective relaying systems: one primary system and one secondary system. Each power circuit breaker is equipped with two separate trip coils, primary and secondary. These components are connected so that each protective function is redundant, and the loss of any component in one relaying protective scheme, including loss of its battery, in no way affects the proper functioning of the other protective scheme. Each transmission line and both switchyards are equipped with overhead static wires as a designed lightning protection system.

Physical separation, the ring bus, breaker-and-a-half switching schemes, redundant switchyard protection systems, and transmission system design based on load flow and stability studies minimize simultaneous failure of all offsite power sources in compliance with GDC 17.

LOSP data were reviewed from an Idaho National Engineering and Environmental Laboratory (INEEL) document, Evaluation of Loss of Offsite Power Events at Nuclear Power Plants: 1986-2003 (Reference 6). Data for LOSP events are divided into five categories: Plant Centered, Switchyard Centered, Grid Related, Severe Weather Related, and Extreme Severe Weather Related. The weather related categories are primarily comprised of data from plants exposed to severe ice storms, heavy snow, extreme salt spray, direct hurricane damage, and high winds. Coastal plants, or plants residing near large bodies of water such as the Great Lakes, are much more likely to be affected by high winds than other plants. HNP is not subject to regular input from these severe weather phenomena, although it is subject to tornadoes. This skew on the LOSP data with regard to HNP would tend to make the conclusions in the INEEL report very conservative. Despite the conservatisms, data on LOSP duration that included all five categories of LOSP events were evaluated. The data conclude that for a LOSP event, the probability of the duration exceeding 8 hr is 0.122 (12.2%). Removing the inherent conservatism, the probability would be lower.

The mitigation function for the TB ventilation system is not required until 9 hr after the initiation of the accident (LOCA, CRDA, or MSLB). Assuming a LOSP coincident with the accident, there is ample time to restore offsite power and initiate TB purging via one exhaust fan of the TB ventilation system within the 9 hr.

2.7.3.4 Turbine Building Exhaust Ductwork Seismic Verification The HNP Units I and 2 TB HVAC exhaust ductwork systems were seismically verified to remain in place and maintain exhaust air flow from the TB though the TB exhaust filters to the exhaust stack for the HNP design basis earthquake. The seismic verification methodology is based on earthquake experience data.

This methodology is provided in the Electric Power Research Institute (EPRI) Technical Report 1007896 "Seismic Evaluation Guidelines for HVAC Duct and Damper Systems," April 2003. The major steps in this methodology are similar to those provided in the SQUG GIP for raceway systems. These steps are documentation review, in-plant screening walkdowns, analytical review of selected duct runs and supports, and identification and resolution of conditions that do not meet the screening or analysis criteria.

The SQUG methodology was previously used in the HNP resolution to USI A-46 and is an accepted verification methodology. A similar methodology, based on the application of earthquake experience data, was used to verify the seismic adequacy of the HNP TB HVAC system.:

SNC had both Units I and 2 TB HVAC exhaust ductwork seismic verifications performed by ABS Consulting, the contractor that developed this seismic verification methodology for HVAC systems and authored the EPRI technical report. The seismic verification reports, which consist of the scope of the verifications, methodology, walkdown summary, analytical review, and outlier summary, are included for

Enclosure 1 Page 58 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term

• Safety Assessment Units 1 and 2 as Enclosures I1 and 12, respectively, of this submittal. A summary of the resolution of outliers for both units is presented in the following paragraphs.

Table 4.1 in each of the TB exhaust ductwork seismic verification reports (Enclosures II and 12) summarizes the outliers identified as a result of the TB exhaust ductwork walkdowns.

The Unit I report (Enclosure 11) identifies four outliers. As detailed in Section 6 of the report, outliers Nos. I and 3 were resolved by analysis, outlier No. 2 has been resolved through repair, and outlier No. 4 has been resolved via modification.

The Unit 2 report (Enclosure 12) identifies seven outliers. As detailed in Section 6 of the report, outliers Nos. 2, 3, 5, 6 and 7 were resolved by analysis. Outliers Nos. I and 4 require repairs and have been entered into the Corrective Action Program. Resolution of outliers Nos. I and 4 is scheduled to be completed by November 28, 2006.

Since this is the first application of the EPRI guidelines and the approach requires the use of engineering judgment, SNC had an independent peer review performed on the EPRI guidelines (Reference 7) and later on the application of these guidelines for the seismic verification of the HNP Unit I TB exhaust HVAC system (Reference 8). Both peer reviews were performed by Dr. R. P. Kennedy, an acknowledged industry expert. Dr. Kennedy served as chairman of the five-member independent Senior Seismic Review and Advisory Panel, which provided considerable technical review and input during the development of the SQUG approach for evaluating the seismic adequacy of 20 classes of equipment plus cable and conduit raceway systems and their supports. The panel unanimously endorsed the SQUG approach for use on existing components in existing nuclear power plants. Dr. Kennedy also served on a four-member independent panel established by the NRC to provide advice on the use of the earthquake experience based approach for the seismic qualification of new equipment, cable trays, and HVAC duct systems in new plants.

Recommendations from the peer review of the EPRI guidelines were incorporated in the application of the guidelines for seismic verification of the HNP Unit I TB exhaust ductwork. The peer review of the seismic verification of the HNP Unit 1 TB exhaust ductwork concluded that the seismic verification was a very thorough and competent evaluation and fully concurs with the conclusions.

2.7.3.5 Passive Ventilation of the Turbine Building A defense-in-depth study is performed to determine an estimate of the passive, wind-driven ventilation of the TB. The study is not prepared in conformance with the standards of 10 CFR 50, Appendix B, and as such, the results of the study are not used in any of the design basis analyses or licensing basis. However, analytical techniques are used that are consistent with industry practices for estimating HVAC requirements for commercial and industrial buildings.

The methodology applied for determining the magnitude of passive ventilation of the TB contains four elements:

• Preparation of a cumulative temporal distribution of wind speed for each of the sixteen cardinal, intercardinal, and bisecting wind directions.

• Determination of the pressure coefficient (Cp) spatial distribution for each of the sixteen cardinal, intercardinal, and bisecting wind directions.

Enclosure 1 Page 59 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment

" Calculation of the wind-generated building internal pressure and cumulative temporal distribution of passive ventilation rate for each of the sixteen cardinal, intercardinal, and bisecting wind directions.

" Combining the cumulative temporal distributions of passive ventilation rate for each of the sixteen cardinal, intercardinal, and bisecting wind directions into a single temporal distribution based on wind direction probability. This final distribution provides the fraction of time that the ventilation rate is less than a given value due to wind-generated pressures around the building.

"Stack effect" (i.e., the natural ventilation brought about by temperature differences between the inside and the outside of the building) is conservatively ignored.

HNP FSAR meteorological data are used for the study (Reference 9, Table 2.3-14, Joint Frequency Table of Wind Speed and Direction). The pressure coefficient (Cp) spatial distribution is determined using the CpCalc+ code developed by Politecnico di Torino (Turin, Italy) for the European Union. Once the distributed CP values are known for each wind direction, the ventilation flow is calculated for the partially-open railway doors on the east fagade of the TB (at the extreme north and south ends) and for leakage through the pre-cast concrete panel construction of the walls. Flow though the walls is based on the following expression:

ELA4 = a where Q(4) is the volumetric flow in m3/sec for a 4 Pa pressure difference (approximately one foot of air) across the faqade, ELA 4 is the equivalent leakage area (in cm 2/m 2 of wall area) under the same conditions, and p is the density of air in kg/mi3. ELA 4 values may be found in Persily, A.K. and Ivy, E. M., "Input Data for Multizone Airflow and IAQ Analysis," NISTIR 6585, January 2001, for many different types of construction. The value used for the HNP TB is 4.0 cm 2/m 2.

By way of illustrating the nature of the leak path through the turbine building faqade, one may note that the pre-cast wall panels used on the turbine building are approximately 7 m x 2 m or 14 m2 with a perimeter of approximately 1800 cm. For a total effective leakage area of approximately 56 cm2 per panel (4 cm 2/m x 14 m2), the effective average joint opening would be about 0.03 cm and the actual joint opening (assuming a head loss coefficient of 3) would be approximately 0.05 cm or 0.5 mm. For adjacent panels, the seam opening would be twice that or approximately 1 mm. Such small seam openings at the construction joints would be barely discernable. Flow through the wall is assumed to be proportional to Ah°0 65 where Ah is the head loss of air across the wall. This proportionality is typically assumed for this type of analysis.

The railway roll-up doors are 20 ft wide and are normally kept open to a height of 5 ft. They are blocked by grating that is assumed to be 80% free area. Assuming a head loss coefficient of 3, the effective area of each partially-open railway door is approximately 46 ft2. Flow through the partially-open railway roll-up doors is assumed to be proportional to the square-root of the head loss, as for an orifice.

Using this model, the following temporal distribution of ventilation rates in the absence of forced ventilation for the TB has been calculated:

Enclosure I Page 60 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Table 36. Temporal Distribution of Turbine Building Ventilation Rates Percentile Volumetric Flow Rate 5th 3,300 cfm 10th 4,900 cfm 20th 7,000 cfm 401h 10,700 cfm For study purposes, following the logic of X/Q development described in NRC regulatory guidance, it would be reasonable to apply the 5th percentile value of 3,300 cfm for the first 8 hr of the MCR dose analysis, 4,900 cfm for the next 16 hr of MCR dose analysis, 7,000 cfm for the next 72 hr of MCR dose analysis, and 10,700 cfm for the remainder of the 30 day duration of the MCR dose analysis. The radiological dose consequences for the MCR for the LOCA, CRDA, and MSLB, assuming only passive ventilation of the TB, are compared to the doses calculated in the design basis analyses using TB exhaust flow of 15,000 cfm (provided by one TB exhaust fan) in Table 37. For the passive ventilation cases, the maximum MCR unfiltered inleakage that could be tolerated and remain within regulatory dose limits is used, instead of the design basis value of 115 cfm.

The results of the defense-in-depth study conclude that significant air exhaust of the TB from natural, wind-driven ventilation would maintain the radiological dose consequences to occupants of the MCR within regulatory limits, with no forced TB exhaust for the full 30-day accident duration for the three DBAs (LOCA, CRDA, and MSLB) that credit TB ventilation. These assume MCR unfiltered inleakage results that are less conservative than the DBA analyses, but with significant margin to measured inleakage results from recent tracer gas inleakage testing.

Table 37. MCR Dose with Passive Ventilation Passive Ventilation Design Basis Limit Maximum MCR Dose MCR Dose (rem TEDE)

Inleakage (rem TEDE) (rem TEDE)

LOCA 70 cfm 4.9 4.9 5 CRDA 85 cfm 4.9 3.8 5 MSLB 130 cfm 4.9 3.9 5 Note: Design basis assumes 115 cf m unfiltered inleakage.

2.8 NUREG-0737 Evaluation The inputs and assumptions utilized in the NUREG-0737 evaluation include the AST plant-specific fission product inventories and other applicable inputs as described in Section 2.5.

Enclosure I Page 61 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment 2.8.1 Post-Accident Access Shielding Plant calculations used in support of plant post-accident vital area 5 access, prepared in accordance with NUREG-0737, Item II.B.2, are evaluated for impact by AST. The implementation of AST results in new activities to be performed post-accident in vital areas. The new activities, applicable post-LOCA, are isolation of the MSLs and establishment of a pathway for MSIV leakage to the main condenser. This requires access to several locations in the TB. A dose evaluation is performed using conservative estimates for walking speed and valve operation times.

The evaluation considered the comparative radiation levels from AST and the existing TID-14844 methodology source term (which is based on reactor power of 2,537 MWt). The results of the evaluation conclude that the current source term used for shielding remains bounding, even with the increase in power to 2,818 MWt and two-year fuel cycles. The NUREG-0737 Item II.B.2 review of plant shielding and environmental qualification of equipment previously completed remains applicable for AST. The dose evaluation of the new activities required to isolate the MSLs and establish the MSIV leakage pathway to the main condenser indicates that these activities can be completed with operator exposures of 5 rem TEDE or less. The evaluated doses continue to meet GDC 19 criteria and NUREG-0737 Item II.B.2.

2.8.2 Post-Accident Radiation Monitor The containment high-range radiation monitors used to monitor post-accident primary containment radiation levels are evaluated for the impact of AST. The monitors continue to provide their design function and envelop the projected radiation exposure rates. Accident radiation monitoring instrumentation continues to meet the requirements of NUREG-0737 Item II.F. 1.

In addition, the control room intake radiation monitor setpoint is evaluated for AST. The evaluation has determined that the current setpoint will alarm and initiate the MCREC system at the start of any of the four DBAs analyzed for AST.

2.8.3 Leakage Control The DBA LOCA control room and TSC dose analysis, as well as that for offsite doses, explicitly considers the effects of coolant leakage outside the primary containment (ESF leakage), satisfying the requirements of NUREG-0737, Item III.D. 1.1.

2.8.4 Control Room and TSC Radiation Protection The radiological dose impacts to the MCR and TSC have been specifically calculated for each of the four DBAs analyzed for AST implementation (NUREG-0737, Item III.A.1.2 and III.D.3.4). In addition, shine from contained sources is also evaluated. The current MCR dose from secondary containment shine is a bounding value for AST. The results of these analyses are presented in Sections 2.5.2.3 (LOCA), 2.5.3.3 (FHA), 2.5.4.3 (CRDA), and 2.5.5.3 (MSLB). The evaluated doses remain less than 5 rem TEDE.

5 As defined by Reference 10, a vital area is any area which will or may require occupancy to permit an operator to aid in the mitigation of or recovery from an accident.

Enclosure 1 Page 62 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment

3. CONCLUSIONS HNP is proposing a full-scope implementation of the AST. Application of the AST methodology for the four DBAs identified in the HNP FSAR that could result in significant control room and offsite doses has been completed using analysis methods and assumptions consistent with the conservative guidance of RG 1.183. This analysis has demonstrated that doses to occupants of the MCR and the TSC, and offsite (EAB and LPZ) doses remain within regulatory limits.

4. REFERENCES

1. Issuance of Amendment - Edwin I Hatch Nuclear Plant, Unit 2 (TAC No. M8785), March 17, 1994.

2. J. Schaperow et al., "Assessment of Radiological Consequences for the Perry Pilot Plant Application using the Revised (NUREG-1465) Source Term," U.S. Nuclear Regulatory Commission, AEB 98-03, December 9, 1998.

3. General Electric Topical Report NEDC-31858P-A, Volumes 1 and 2, "BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems," August 1999.

4. E. C. Beahm, C. F. Weber, T. S. Kress, and G. W. Parker, "Iodine Chemical Forms in LWR Severe Accidents," NUREG/CR-5732, ORNL/TM-1 1861, April 1992.

5. Safety Evaluation of GE Topical Report, NEDC-31858P, Revision 2, "BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems," U.S. Nuclear Regulatory Commission, March 3, 1999.

6. NUREG/CR-INEEL/EXT-04-02326, "Evaluation of Loss of Offsite Power Events at Nuclear Power Plants: 1986 - 2003" (Draft), October 2004.

7. Peer Review Comments on EPRI Seismic Evaluation Guidelines for HVAC Duct and Damper Systems, R. P. Kennedy, February 7, 2004.

8. Peer Review of Seismic Verification of the Turbine Building Exhaust Ductwork for Hatch Nuclear Plant Unit 1, R. P. Kennedy, November 24, 2004.

9. Hatch Unit 2 Final Safety Analysis Report, Revision 23H.

10. NRC NUREG-0737, "Clarification of TMI Action Plan Requirements," November 1980.

11. ASTM Consensus Standard E741, "Standard Test Method for Determining Air Change in a Single Zone by Means of a Tracer Gas Dilution."

12. Branch Technical Position Containment Systems Branch 6-3, "Determination of Bypass Leakage Paths in Dual Containment Plants," NRC SRP Section 6.2.3, July 1981.

13. Edwin I. Hatch Nuclear Plant, Unit No. I Renewed Facility Operating License DPR-57, issued January 15, 2002.

Enclosure I Page 63 of 63 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment

14. Edwin I. Hatch Nuclear Plant, Unit No. 2 Renewed Facility Operating License NPF-5, issued January 15, 2002.

15. EPRI Technical report 1007896, "Seismic Evaluation Guidelines for HVAC Duct and Damper Systems," Electric Power Research Institute, April 2003.

16. Federal Guidance Report No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1988.

17. Federal Guidance Report No. 12, "External Exposure to Radionuclides in Air, Water, and Soil,"

1993.

18. Generic Implementation Procedure (GIP) For Seismic Verification of Nuclear Plant Equipment, Revision 3A, Seismic Qualification Utility Group, December 2001.

19. Hatch Unit 1 Final Safety Analysis Report, Revision 23G.

20. Hatch Unit I Technical Specifications, Amendment 248.

21. Hatch Unit 2 Technical Specifications, Amendment 192.

22. Letter from U.S. NRC to H. L. Sumner, Jr. of Southern Nuclear Operating Company, Inc.,

Subject:

Edwin I. Hatch Nuclear Plant, Unit Nos. I and 2 - Issuance of Amendments Regarding Revise Operating Licenses to Support the Crediting of Potassium Iodide for an Interim Period (TAC Nos.

MD0525 and MD0526), May 25, 2006.

23. NRC Final Review Guidelines, "Guidance on the Assessment of a BWR SLC System for pH Control," February 12, 2004.

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

25. NRC NUREG-1465, "Accident Source Terms for Light Water Reactors for Light-Water Nuclear Power Plants," February 1995.

26. NRC Regulatory Guide 1.183, "Alternative Source Terms for Evaluating Design Basis Accidents At Nuclear Power Reactors," July 2000.

27. NRC Regulatory Issues Summary 2006-04, "Experience with Implementation of Alternative Source Terms," March 7, 2006.

28. NRC Standard Review Plan Section 15.0.1, "Radiological Consequences Analyses Using Alternative Source Terms," Revision 0, July 2000.

29. ORIGEN2 Computer Code, Oak Ridge National Laboratory.

30. Safety Evaluation on the Resolution of Unresolved Safety Issue A-46 at Edwin I. Hatch Plant, Units 1 and 2 (TAC Nos. N69451 and M69452), enclosure to letter from Leonard N. Olshan, Project Manager, NRC to Mr. H. L. Summer, Jr., Southern Nuclear Operating Company, September 24, 1998.

31. Technical Information Document - 14844, "Calculation of Distance Factors for Power and Test Reactor Sites," U.S. Atomic Energy Commission, March 23, 1962.

Enclosure I Page A-I Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment APPENDIX A: Regulatory Guide 1.183 Conformance Matrix Conformance with Regulatory Guide 1.183 - Main Sections RG Sec RG Position HNP Position Comments 3.1 The inventory of fission products in the reactor core and available for release to the Conforms. Core power accounts for containment should be based on the maximum full power operation of the core with, as a 0.5% uncertainty. 10%

minimum, current licensed values for fuel enrichment, fuel bumup, and an assumed core margin is added to core power equal to the current licensed rated thermal power times the ECCS evaluation fission product inventory uncertainty. The period of irradiation should be of sufficient duration to allow the activity to allow for future fuel of dose-significant radionuclides to reach equilibrium or to reach maximum values. The changes or power uprates.

core inventory should be determined using an appropriate isotope generation and depletion computer code such as ORIGEN 2 (Ref. 17) or ORIGEN-ARP (Ref. 18). Core inventory factors (Ci/MWt) provided in TID14844 and used in some analysis computer codes were derived for low burnup, low enrichment fuel and should not be used with higher burnup and higher enrichment fuels.

3.1 For the DBA LOCA, all fuel assemblies in the core are assumed to be affected and the Conforms.

_ _ core average inventory should be used.

3.2 The core inventory release fractions, by radionuclide groups, for the gap release and early Conforms. The release fractions from in-vessel damage phases for DBA LOCAs are listed in Table 1 for BWRs and Table 2 for Table 1 are used.

PWRs. These fractions are applied to the equilibrium core inventory described in Regulatory Position 3.1.

3.2 For non-LOCA events, the fractions of the core inventory assumed to be in the gap for the Conforms. Release fraction of 0.1 is various radionuclides are given in Table 3. The release fractions from Table 3 are used in used for 1-131, Other conjunction with the fission product inventory calculated with the maximum core radial Noble Gases, and Other peaking factor. Halogens in the CRDA analysis.

Table 3 Non-LOCA Fraction of Fission Product Inventory in Gap Group Fraction 1-131 0.08 Kr-85 0.10

Enclosure I Page A-2 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Conformance with Regulatory Guide 1.183 - Main Sections RG Sec RG Position HNP Position Comments 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 DBA LOCAs Conforms. The BWR durations from at PWRs and BWRs. The specified onset is the time following the initiation of the Table 4 are used for the accident (i.e., time = 0). The early in-vessel phase immediately follows the gap release LOCA.

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. 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.

3.4 Table 5 lists the elements in each radionuclide group that should be considered in design Conforms.

basis analyses.

3.5 Of the radioiodine released from the reactor coolant system (RCS) to the containment in a Conforms.

postulated accident, 95 percent of the iodine released should be assumed to be cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. This includes releases from the gap and the fuel pellets. With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form. The same chemical form is assumed in releases from fuel pins in FHAs and from releases from the fuel pins through the RCS in DBAs other than FHAs or LOCAs.

However, the transport of these iodine species following release from the fuel may affect these assumed fractions. The accident-specific appendices to this regulatory guide provide additional details.

3.6 The amount of fuel damage caused by non-LOCA design basis events should be analyzed Conforms.

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

Enclosure I Page A-3 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Conformance with Regulatory Guide 1.183 - Main Sections RG Sec RG Position HNP Position Comments deposition, for estimating fuel damage for the purpose of establishing radioactivity releases.

4.1 Offsite Dose Consequences Conforms.

4.1.1 The dose calculations should determine the TEDE. TEDE is the sum of the committed Conforms.

effective dose equivalent (CEDE) from inhalation and the deep dose equivalent (DDE) from external exposure. The calculation of these two components of the TEDE should consider all radionuclides, including progeny from the decay of parent radionuclides, that are significant with regard to dose consequences and the released radioactivity.

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

Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion" (Ref. 20), provides tables of conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the CEDE.

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

10-4 cubic meters per second. From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the breathing rate should be assumed to be 1.8 x 104 cubic meters per second. After that and until the end of the accident, the rate should be assumed to be 2.3 x 10. cubic meters per second.

4.1.4 The DDE should be calculated assuming submergence in semi-infinite cloud assumptions Conforms. Dose Conversion Factors with appropriate credit for attenuation by body tissue. The DDE is nominally equivalent from Federal Guidance to the effective dose equivalent (EDE) from external exposure if the whole body is Report 12 are used.

irradiated uniformly. Since this is a reasonable assumption for submergence exposure situations, EDE may be used in lieu of DDE in determining the contribution of external dose to the TEDE. Table 1I. 1 of Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil" (Ref. 21), provides external EDE conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the EDE.

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

Enclosure I Page A-4 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Conformance with Regulatory Guide 1.183 - Main Sections RG Sec RG Position HNP Position Comments EAB TEDE for any two-hour period following the start of the radioactivity release should be determined and used in determining compliance with the dose criteria in 10 CFR 50.67. The maximum two-hour TEDE should be determined by calculating the postulated dose for a series of small time increments and performing a "sliding" sum over the increments for successive two-hour periods. The maximum TEDE obtained is submitted.

The time increments should appropriately reflect the progression of the accident to capture the peak dose interval between the start of the event and the end of radioactivity release (see also Table 6).

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

ground.

4.2.1 The TEDE analysis should consider all sources of radiation that will cause exposure to Conforms.

control room personnel. The applicable sources will vary from facility to facility, but typically will include:

• Contamination of the control room atmosphere by the intake or infiltration of the radioactive material contained in the radioactive plume released from the facility,

• Contamination of the control room atmosphere by the intake or infiltration of airborne radioactive material from areas and structures adjacent to the control room envelope,

" Radiation shine from the external radioactive plume released from the facility,

" Radiation shine from radioactive material in the reactor containment,

• Radiation shine from radioactive material in systems and components inside or external to the control room envelope, e.g., radioactive material buildup in recirculation filters.

4.2.2 The radioactive material releases and radiation levels used in the control room dose Conforms. Main condenser release analysis should be determined using the same source term, transport, and release (from MSIV leakage and assumptions used for determining the EAB and the LPZ TEDE values, unless these secondary containment Page A-5 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Conformance with Regulatory Guide 1.183 - Main Sections RG Sec RG Position HNP Position Comments assumptions would result in non-conservative results for the control room. bypass leakage) is to turbine building, to maximize control room dose since the control room is within the turbine

_ building.

4.2.3 The models used to transport radioactive material into and through the control room, and Conforms.

the shielding models used to determine radiation dose rates from external sources, should be structured to provide suitably conservative estimates of the exposure to control room personnel.

4.2.4 Credit for engineered safety features that mitigate airborne radioactive material within the Conforms. Pressurization and intake control room may be assumed. Such features may include control room isolation or filtration are credited.

pressurization, or intake or recirculation filtration. Refer to Section 6.5.1, "ESF Atmospheric Cleanup System," of the SRP (Ref. 3) and Regulatory Guide 1.52, "Design, Testing, and Maintenance Criteria for Post-accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants" (Ref. 25), for guidance.

4.2.5 Credit should generally not be taken for the use of personal protective equipment or Conforms.

prophylactic drugs. Deviations may be considered on a case-by-case basis.

4.2.6 The dose receptor for these analyses is the hypothetical maximum exposed individual Conforms.

who is present in the control room for 100% of the time during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the event, 60% of.the time between I and 4 days, and 40% of the time from 4 days to 30 days. For the duration of the event, the breathing rate of this individual should be assumed to be 3.5 x 104 cubic meters per second.

4.2.7 Control room doses should be calculated using dose conversion factors identified in Conforms. A rigorous analysis Regulatory Position 4.1 above for use in offsite dose analyses. The DDE from photons treating the control room may be corrected for the difference between finite cloud geometry in the control room and as a 148' x 66' x 16' the semi-infinite cloud assumption used in calculating the dose conversion factors. The volume was performed.

following expression may be used to correct the semi-infinite cloud dose, DDEO., to a Based on this rigorous

Enclosure I Page A-6 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Conformance with Regulatory Guide 1.183 - Main Sections RG Sec RG Position HNP Position Comments finite cloud dose, DDEinkte, where the control room is modeled as a hemisphere that has a analysis, an additional 0.5 volume, V, in cubic feet, equivalent to that of the control room (Ref. 22). correction factor is applied DDEý0..338 to the correction factor

• 173 calculated from the r1173 expression to the left.

4.3 The guidance provided in Regulatory Positions 4.1 and 4.2 should be used, as applicable, Conforms. NUREG-0737 analysis in re-assessing the radiological analyses identified in Regulatory Position 1.3.1, such as using AST was completed.

those in NUREG-0737 (Ref. 2). Design envelope source terms provided in NUREG-0737 should be updated for consistency with the AST. In general, radiation exposures to plant personnel identified in Regulatory Position 1.3.1 should be expressed in terms of TEDE.

Integrated radiation exposure of plant equipment should be determined using the guidance of Appendix I of this guide.

4.4 The radiological criteria for the EAB, the outer boundary of the LPZ, and for the control Conforms.

room are in 10 CFR 50.67. These criteria are stated for evaluating reactor accidents of exceedingly low probability of occurrence and low risk of public exposure to radiation, e.g., a large-break LOCA. The control room criterion applies to all accidents. For events with a higher probability of occurrence, postulated EAB and LPZ doses should not exceed the criteria tabulated in Table 6.

The acceptance criteria for the various NUREG-0737 (Ref. 2) items generally reference General Design Criteria 19 (GDC 19) from Appendix A to 10 CFR Part 50 or specify criteria derived from GDC-19. These criteria are generally specified in terms of whole body dose, or its equivalent to any body organ. For facilities applying for, or having received, approval for the use of an AST, the applicable criteria should be updated for consistency with the TEDE criterion in 10 CFR 50.67(b)(2)(iii).

5.1.1 The evaluations required by 10 CFR 50.67 are re-analyses of the design basis safety Conforms.

analyses and evaluations required by 10 CFR 50.34; they are considered to be a significant input to the evaluations required by 10 CFR 50.92 or 10 CFR 50.59. These analyses should be prepared, reviewed, and maintained in accordance with quality assurance programs that comply with Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50.

Page A-7 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Conformance with Regulatory Guide 1.183 - Main Sections RG Sec RG Position HNP Position Comments 5.1.2 Credit may be taken for accident mitigation features that are classified as safety-related, Does not Credit is taken for the are required to be operable by technical specifications, are powered by emergency power conform. alternate leakage treatment sources, and are either automatically actuated or, in limited cases, have actuation pathway via the main requirements explicitly addressed in emergency operating procedures. The single active condenser, the standby component failure that results in the most limiting radiological consequences should be liquid control system (a assumed. Assumptions regarding the occurrence and timing of a loss of offsite power safe shutdown system, not should be selected with the objective of maximizing the postulated radiological an ESF system), and consequences. turbine building ventilation. See Section 2.7 for the justification of this credit.

5.1.3 The numeric values that are chosen as inputs to the analyses required by 10 CFR 50.67 Conforms.

should be selected with the objective of determining a conservative postulated dose. In some instances, a particular parameter may be conservative in one portion of an analysis but be nonconservative in another portion of the same analysis.

5.1.4 Licensees should ensure that analysis assumptions and methods are compatible with the Conforms.

AST and the TEDE criteria.

5.2 Licensees should analyze the DBAs that are affected by the specific proposed applications Conforms.

of an AST.

5.3 Atmospheric dispersion values (X/Q) for the EAB, the LPZ, and the control room that Conforms. Only control room were approved by the staff during initial facility licensing or in subsequent licensing atmospheric dispersion proceedings may be used in performing the radiological analyses identified by this guide. factors were re-calculated Methodologies that have been used for determining X/Q values are documented in for AST. ARCON96 was Regulatory Guides 1.3 and 1.4, Regulatory Guide 1.145, "Atmospheric Dispersion used for updating these Models for Potential Accident Consequence Assessments at Nuclear Power Plants," and values. These values are the paper, "Nuclear Power Plant Control Room Ventilation System Design for Meeting also applied to the TSC, General Criterion 19" (Refs. 6, 7, 22, and 28). since the control room The methodology of the NRC computer code ARCON96 (Ref 26) is generally acceptable values are bounding.

to the NRC staff for use in determining control room X/Q values.

Enclosure I Page A-8 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Conformance with Regulatory Guide 1.183 - Appendix A (LOCA)

App Sec RG Position HNP Position Comments 1 Acceptable assumptions regarding core inventory and the release of radionuclides from Conforms. See main Sections 3.1 to the fuel are provided in Regulatory Position 3 of this guide. 3.4 for more information.

2 If the sump or suppression pool pH is controlled at values of 7 or greater, the chemical Conforms.

form of radioiodine released to the containment should be assumed to be 95% cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. Iodine species, including those from iodine re-evolution, for sump or suppression pool pH values less than 7 will be evaluated on a case-by-case basis. Evaluations of pH should consider the effect of acids and bases created during the LOCA event, e.g., radiolysis products. With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form.

3.1 The radioactivity released from the fuel should be assumed to mix instantaneously and Conforms. Flow from the DW to the homogeneously throughout the free air volume of the primary containment in PWRs or torus prior to the assumed the drywell in BWRs as it is released. This distribution should be adjusted if there are core quench at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is internal compartments that have limited ventilation exchange. The suppression pool free conservatively ignored.

air volume may be included provided there is a mechanism to ensure mixing between the Post-reflood flow from drywell to the wetwell. The release into the containment or drywell should be assumed DW to torus is considered to terminate at the end of the early in-vessel phase. for a period of time to bring about a near uniform distribution of activity.

This modeling is conservative relative to the assumption of a well-mixed containment post-reflood.

3.2 Reduction in airborne radioactivity in the containment by natural deposition within the Conforms. STARNAUA is used containment may be credited. Acceptable models for removal of iodine and aerosols are instead of NUREG/CR-Page A-9 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Conformance with Regulatory Guide 1.183 - Appendix A (LOCA)

App Sec RG Position HNP Position Comments described in Chapter 6.5.2, "Containment Spray as a Fission Product Cleanup System," 6189 to predict natural of the Standard Review Plan (SRP), NUREG-0800 (Ref. A-I) and in NUREG/CR-6189, deposition of aerosol prior "A Simplified Model of Aerosol Removal by Natural Processes in Reactor to the start of sprays. Also Containments" (Ref. A-2). The latter model is incorporated into the analysis code refer to item 3.3 below.

RADTRAD (Ref. A-3).

3.3 Reduction in airborne radioactivity in the containment by containment spray systems that Conforms. STARNAUA is used have been designed and are maintained in accordance with Chapter 6.5.2 of the SRP instead of the methods (Ref. A-i) may be credited. Acceptable models for the removal of iodine and aerosols cited. STARNAUA are described in Chapter 6.5.2 of the SRP and NUREG/CR-5966, "A Simplified Model combines the effects of of Aerosol Removal by Containment Sprays"l (Ref. A-4). This simplified model is both natural deposition and incorporated into the analysis code RADTRAD (Refs. A-I to A-3). sprays.

3.3 The evaluation of the containment sprays should address areas within the primary Conforms. DW is assumed to be well-containment that are not covered by the spray drops. The mixing rate attributed to mixed based on the fact natural convection between sprayed and unsprayed regions of the containment building, that the DW is sufficiently provided that adequate flow exists between these regions, is assumed to be two turnovers small and the spray of the unsprayed regions per hour, unless other rates are justified. The containment flowrate is sufficiently building atmosphere may be considered a single, well-mixed volume if the spray covers large (i.e., the ratio of at least 90% of the volume and if adequate mixing of unsprayed compartments can be spray flow to volume shown. sprayed is 20-40 times larger for the HNP DW than for a typical sprayed region of a PWR) that mixing by momentum exchange alone (between the droplets and the atmosphere) will keep the DW well-mixed; i.e.,

natural convection will play no noticeable role.

Enclosure I Page A-10 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Conformance with Regulatory Guide 1.183 - Appendix A (LOCA)

App Sec RG Position HNP Position Comments 3.3 The SRP sets forth a maximum decontamination factor (DF) for elemental iodine based Conforms. Maximum iodine DF is on the maximum iodine activity in the primary containment atmosphere when the sprays based on projected pH of actuate, divided by the activity of iodine remaining at some time after decontamination. suppression pool, not on The SRP also states that the particulate iodine removal rate should be reduced by a factor the SRP. STARNAUA of 10 when a DF of 50 is reached. The reduction in the removal rate is not required if the does not explicitly reduce removal rate is based on the calculated time-dependent airborne aerosol mass. There is the removal rate for no specified maximum DF for aerosol removal by sprays. The maximum activity to be particulate by a factor of used in determining the DF is defined as the iodine activity in the columns labeled 10 when a DF of 50 is "Total" in Tables I and 2 of this guide multiplied by 0.05 for elemental iodine and by reached, but the code does 0.95 for particulate iodine (i.e., aerosol treated as particulate in SRP methodology), take into account the small size of the remaining particles, and the same removal rate reduction effect is realized.

3.4 Reduction in airborne radioactivity in the containment by in-containment recirculation Not applicable.

filter systems may be credited if these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. A-5 and A-6). The filter media loading caused by the increased aerosol release associated with the revised source term should be addressed.

3.5 Reduction in airborne radioactivity in the containment by suppression pool scrubbing in Conforms. Suppression pool BWRs should generally not be credited. However, the staff may consider such reduction scrubbing not credited.

on an individual case basis. The evaluation should consider the relative timing of the blowdown and the fission product release from the fuel, the force driving the release through the pool, and the potential for any bypass of the suppression pool (Ref. 7).

Analyses should consider iodine re-evolution if the suppression pool liquid pH is not maintained greater than 7.

3.6 Reduction in airborne radioactivity in the containment by retention in ice condensers, or other engineering safety features not addressed above, should be evaluated on an individual case basis. See Section 6.5.4 of the SRP (Ref. A-1).

Enclosure I Page A-11 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Conformance with Regulatory Guide 1.183 - Appendix A (LOCA)

AppSec RG Position HNP Position Comments 3.7 The primary containment (i.e., drywell for Mark I and II containment designs) should be Conforms.

assumed to leak at the peak pressure technical specification leak rate for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. For PWRs, the leak rate may be reduced after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 50% of the technical specification leak rate. For BWRs, leakage may be reduced after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, if supported by plant configuration and analyses, to a value not less than 50% of the technical specification leak rate. Leakage from subatmospheric containments is assumed to terminate when the containment is brought to and maintained at a subatmospheric condition as defined by technical specifications.

3.8 If the primary containment is routinely purged during power operations, releases via the Not applicable.

purge system prior to containment isolation should be analyzed and the resulting doses summed with the postulated doses from other release paths. The purge release evaluation should assume that 100% of the radionuclide inventory in the reactor coolant system liquid is released to the containment at the initiation of the LOCA. This inventory should be based on the technical specification reactor coolant system equilibrium activity. Iodine spikes need not be considered. If the purge system is not isolated before the onset of the gap release phase, the release fractions associated with the gap release and early in-vessel phases should be considered as applicable.

4.1 Leakage from the primary containment should be considered to be collected, processed Conforms.

by engineered safety feature (ESF) filters, if any, and released to the environment via the secondary containment exhaust system during periods in which the secondary containment has a negative pressure as defined in technical specifications. Credit for an elevated release should be assumed only if the point of physical release is more than two and one-half times the height of any adjacent structure.

4.2 Leakage from the primary containment is assumed to be released directly to the Conforms.

environment as a ground-level release during any period in which the secondary containment does not have a negative pressure as defined in technical specifications.

4.3 The effect of high wind speeds on the ability of the secondary containment to maintain a Conforms. The 9 5th percentile wind negative pressure should be evaluated on an individual case basis. The wind speed to be speed for HNP is assumed is the 1-hour average value that is exceeded only 5% of the total number of approximately 12 mph.

Enclosure I Page A- 12 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Conformance with Regulatory Guide 1.183 - Appendix A (LOCA)

App Sec RG Position HNP Position Comments hours in the data set. Ambient temperatures used in these assessments should be the 1- The dynamic pressure for hour average value that is exceeded only 5% or 95% of the total numbers of hours in the such a wind speed is 0.07 data set, whichever is conservative for the intended use (e.g., if high temperatures are in wg. Even if the limiting, use those exceeded only 5%). minimum wind pressure coefficient on the reactor building faqade were as negative as -1.0, the minimum pressure on the surface of the reactor building (-0.07 in wg relative to ambient static pressure) would still be greater than the Technical Specification surveillance limit of at least -0.2 in wg (with approximately 200%

margin).

4.4 Credit for dilution in the secondary containment may be allowed when adequate means Conforms. 50% of the reactor

,to cause mixing can be demonstrated. Otherwise, the leakage from the primary building volume is containment should be assumed to be transported directly to exhaust systems without credited for dilution.

mixing. Credit for mixing, if found to be appropriate, should generally be limited to 50%. This evaluation should consider the magnitude of the containment leakage in relation to contiguous building volume or exhaust rate, the location of exhaust plenums relative to projected release locations, the recirculation ventilation systems, and internal walls and floors that impede stream flow between the release and the exhaust.

4.5 Primary containment leakage that bypasses the secondary containment should be Conforms.

evaluated at the bypass leak rate incorporated in the technical specifications. If the bypass leakage is through water, e.g., via a filled piping run that is maintained full, credit for retention of iodine and aerosols may be considered on a case-by-case basis.

Similarly, deposition of aerosol radioactivity in gas-filled lines may be considered on a Page A- 13 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Conformance with Regulatory Guide 1.183 - Appendix A (LOCA)

App Sec RG Position HNP Position Comments case-by-case basis.

4.6 Reduction in the amount of radioactive material released from the secondary Conforms.

containment because of ESF filter systems may be taken into account provided that these systems meet the guidance of Regulatory Guide 1.52 (Ref. A-5) and Generic Letter 99-02 (Ref. A-6).

5.1 With the exception of noble gases, all the fission products released from the fuel to the Conforms.

containment (as defined in Tables 1 and 2 of this guide) should be assumed to instantaneously and homogeneously mix in the primary containment sump water (in PWRs) or suppression pool (in BWRs) at the time of release from the core. In lieu of this deterministic approach, suitably conservative mechanistic models for the transport of airborne 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.

5.2 The leakage should be taken as two times the sum of the simultaneous leakage from all Conforms. As there is no technical components in the ESF recirculation systems above which the technical specifications, or specification limit, a licensee commitments to item III.D. 1.1 of NUREG-0737 (Ref. A-8), would require conservatively high declaring such systems inoperable. The leakage should be assumed to start at the earliest leakage rate of 10 gpm is time the recirculation flow occurs in these systems and end at the latest time the releases assumed. ESF leakage is from these systems are terminated. Consideration should also be given to design leakage assumed to begin at the through valves isolating ESF recirculation systems from tanks vented to atmosphere, time DW sprays are e.g., emergency core cooling system (ECCS) pump miniflow return to the refueling started.

water storage tank.

5.3 With the exception of iodine, all radioactive materials in the recirculating liquid should Conforms.

be assumed to be retained in the liquid phase.

5.4 If the temperature of the leakage exceeds 212'F, the fraction of total iodine in the liquid Not applicable.

that becomes airborne should be assumed equal to the fraction of the leakage that flashes to vapor. I I__

Page A- 14 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Conformance with Regulatory Guide 1.183 - Appendix A (LOCA)

App Sec RG Position HNP Position Comments 5.5 If the temperature of the leakage is less than 212'F or the calculated flash fraction is less Conforms. A release fraction of 10%

than 10%, the amount of iodine that becomes airborne should be assumed to be 10% of is assumed.

the total iodine activity in the leaked fluid, unless a smaller amount can be justified based on the actual sump pH history and area ventilation rates.

5.6 The radioiodine that is postulated to be available for release to the environment is Conforms. Credit is taken for holdup assumed to be 97% elemental and 3% organic. Reduction in release activity by dilution and dilution of ESF or holdup within buildings, or by ESF ventilation filtration systems, may be credited leakage in reactor building where applicable. Filter systems used in these applications should be evaluated against and for release through the guidance of Regulatory Guide 1.52 (Ref. A-5) and Generic Letter 99-02 (Ref. A-6). SGTS filters via plant stack.

6.1 For the purpose of this analysis, the activity available for release via MSIV leakage Conforms.

should be assumed to be that activity determined to be in the drywell for evaluating containment leakage (see Regulatory Position 3). No credit should be assumed for activity reduction by the steam separators or by iodine partitioning in the reactor vessel.

6.2 All the MSIVs should be assumed to leak at the maximum leak rate above which the Conforms. The full technical technical specifications would require declaring the MSIVs inoperable. The leakage specification maximum should be assumed to continue for the duration of the accident. Postulated leakage may combined leakage for all be reduced after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, if supported by site-specific analyses, to a value not MSIVs is conservatively less than 50% of the maximum leak rate. assumed through the failed line.

6.3 Reduction of the amount of released radioactivity by deposition and plateout on steam Conforms. Impaction credited as well system piping upstream of the outboard MSIVs may be credited, but the amount of as sedimentation reduction in concentration allowed will be evaluated on an individual case basis. (calculated with Generally, the model should be based on the assumption of well-mixed volumes, but STARNAUA). Well-other models such as slug flow may be used if justified. mixed volumes are assumed.

6.4 In the absence of collection and treatment of releases by ESFs such as the MSIV leakage Conforms. Since the control room is control system, or as described in paragraph 6.5 below, the MSIV leakage should be I located within the turbine

Enclosure I Page A-15 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Conformance with Regulatory Guide 1.183 - Appendix A (LOCA)

App Sec RG Position HNP Position Comments assumed to be released to the environment as an unprocessed, ground- level release. building, it is more Holdup and dilution in the turbine building should not be assumed. conservative to assume holdup in the turbine building than to assume direct release to the environment when calculating control room dose.

6.5 A reduction in MSIV releases that is due to holdup and deposition in main steam piping Conforms. Particulate and elemental downstream of the MSIVs and in the main condenser, including the treatment of air iodine deposition is ejector effluent by offgas systems, may be credited if the components and piping systems credited in the piping and used in the release path are capable of performing their safety function during and in the main condenser.

following a safe shutdown earthquake (SSE). The amount of reduction allowed will be Particulate deposition is evaluated on an individual case basis. References A-9 and A-10 provide guidance on calculated using acceptable models. STARNAUA.

7 The radiological consequences from post-LOCA primary containment purging as a Not applicable.

combustible gas or pressure control measure should be analyzed. If the installed containment purging capabilities are maintained for purposes of severe accident management and are not credited in any design basis analysis, radiological consequences need not be evaluated. If the primary containment purging is required within 30 days of the LOCA, the results of this analysis should be combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA. Reduction in the amount of radioactive material released via ESF filter systems may be taken into account provided that these systems meet the guidance in Regulatory Guide 1.52 (Ref. A-5) and Generic Letter 99-02 (Ref. A-6).

Conformance with Regulatory Guide 1.183- Appendix B (FHA)

App Sec I RG Position I HNP Position Comments

Enclosure I Page A- 16 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Conformance with Regulatory Guide 1.183 - Appendix B (FHA)

App Sec RG Position HNP Position Comments 1 Acceptable assumptions regarding core inventory and the release of radionuclides from Conforms.

the fuel are provided in Regulatory Position 3 of this guide.

1.1 The number of fuel rods damaged during the accident should be based on a conservative Conforms.

analysis that considers the most limiting case. This analysis should consider parameters such as the weight of the dropped heavy load or the weight of a dropped fuel assembly (plus any attached handling grapples), the height of the drop, and the compression, torsion, and shear stresses on the irradiated fuel rods. Damage to adjacent fuel assemblies, if applicable (e.g., events over the reactor vessel), should be considered.

1.2 The fission product release from the breached fuel is based on Regulatory Position 3.2 of Conforms.

this guide and the estimate of the number of fuel rods breached. All the gap activity in the damaged rods is assumed to be instantaneously released. Radionuclides that should be considered include xenons, kryptons, halogens, cesiums, and rubidiums.

1.3 The chemical form of radioiodine released from the fuel to the spent fuel pool should be Conforms.

assumed to be 95% cesium iodide (Csl), 4.85 percent elemental iodine, and 0.15 percent organic iodide. The CsI released from the fuel is assumed to completely dissociate in the pool water. Because of the low pH of the pool water, the iodine re-evolves as elemental iodine. This is assumed to occur instantaneously. The NRC staff will consider, on a case-by-case basis, justifiable mechanistic treatment of the iodine release from the pool.

2 If the depth of water above the damaged fuel is 23 feet or greater, the decontamination Conforms. Water depth is 21 ft. A factors for the elemental and organic species are 500 and 1, respectively, giving an smaller decontamination overall effective decontamination factor of 200 (i.e., 99.5% of the total iodine released factor is calculated using from the damaged rods is retained by the water). This difference in decontamination the guidance of the factors for elemental (99.85%) and organic iodine (0.15%) species results in the iodine regulatory guide.

above the water being composed of 57% elemental and 43% organic species. If the depth of water is not 23 feet, the decontamination factor will have to be determined on a case-by-case method (Ref. B-I).

3 The retention of noble gases in the water in the fuel pool or reactor cavity is negligible Conforms.

(i.e., decontamination factor of 1). Particulate radionuclides are assumed to be retained

Enclosure I Page A- 17 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Conformance with Regulatory Guide 1.183 - Appendix B (FHA)

App Sec RG Position HNP Position Comments by the water in the fuel pool or reactor cavity (i.e., infinite decontamination factor).

4 Fuel handling accidents within the fuel building. Not applicable.

5.1 If the containment is isolated during fuel handling operations, no radiological Conforms. Containment is not consequences need to be analyzed. isolated. Radiological consequences analyzed.

5.2 If the containment is open during fuel handling operations, but designed to automatically Not applicable.

isolate in the event of a fuel handling accident, the release duration should be based on delays in radiation detection and completion of containment isolation. If it can be shown that containment isolation occurs before radioactivity is released to the environment, no radiological consequences need to be analyzed.

5.3 If the containment is open during fuel handling operations (e.g., personnel air lock or Conforms.

equipment hatch is open), the radioactive material that escapes from the reactor cavity pool to the containment is released to the environment over a 2-hour time period.

5.4 A reduction in the amount of radioactive material released from the containment by ESF Conforms. ESF filter systems not filter systems may be taken into account provided that these systems meet the guidance credited.

of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. B-2 and B-3). Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system should be determined and accounted for in the radioactivity release analyses.

5.5 Credit for dilution or mixing of the activity released from the reactor cavity by natural or Conforms. No credit taken for dilution forced convection inside the containment may be considered on a case-by-case basis. or mixing in the reactor Such credit is generally limited to 50% of the containment free volume. This evaluation building.

should consider the magnitude of the 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 stream flow between the surface of the reactor cavity and the exhaust plenums.

S---------.impede Page A- 18 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Conformance with Regulatory Guide 1.183 - Appendix C (CRDA)

App Sec RG Position HNP Position Comments I Assumptions acceptable to the NRC staff regarding core inventory are provided in Conforms.

Regulatory Position 3 of this guide. For the rod drop accident, the release from the breached fuel is based on the estimate of the number of fuel rods breached and the assumption that 10% of the core inventory of the noble gases and iodines is in the fuel gap. The release attributed to fuel melting is based on the fraction of the fuel that reaches or exceeds the initiation temperature for fuel melting and on the assumption that 100% of the noble gases and 50% of the iodines contained in that fraction are released to the reactor coolant.

2 If no or minimal fuel damage is postulated for the limiting event, the released activity Conforms. Substantial fuel damage is should be the maximum coolant activity (typically 4 jiCi/gm DE 1-131) allowed by the postulated. Coolant technical specifications. activity neglected.

3.1 The activity released from the fuel from either the gap or from fuel pellets is assumed to Conforms.

be instantaneously mixed in the reactor coolant within the pressure vessel. I 3.2 Credit should not be assumed for partitioning in the pressure vessel or for removal by the Conforms.

steam separators.

3.3 Of the activity released from the reactor coolant within the pressure vessel, 100% of the Conforms.

noble gases, 10% of the iodine, and 1% of the remaining radionuclides are assumed to reach the turbine and condensers.

3.4 Of the activity that reaches the turbine and condenser, 100% of the noble gases, 10% of Conforms.

the iodine, and 1% of the particulate radionuclides are available for release to the environment. The turbine and condensers leak to the atmosphere as a ground- level release at a rate of 1% per day for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, at which time the leakage is assumed to terminate. No credit should be assumed for dilution or holdup within the turbine building. Radioactive decay during holdup in the turbine and condenser may be assumed.

3.5 In lieu of the transport assumptions provided in paragraphs 3.2 through 3.4 above, a Not applicable.

___E_____ more mechanistic analysis may be used on a case-by-case basis. Such analyses account III

Enclosure I Page A-19 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Conformance with Regulatory Guide 1.183 - Appendix C (CRDA)

App Sec RG Position HNP Position Comments for the quantity of contaminated steam carried from the pressure vessel to the turbine and condensers based on a review of the minimum transport time from the pressure vessel to the first main steam isolation (MSIV) and considers MSIV closure time.

3.6 The iodine species released from the reactor coolant within the pressure vessel should be Conforms.

assumed to be 95% CsI as an aerosol, 4.85% elemental, and 0.15% organic. The release from the turbine and condenser should be assumed to be 97% elemental and 3% organic.

Conformance with Regulatory Guide 1.183 - Appendix D (MSLB)

App Sec RG Position HNP Position Comments 1 Assumptions acceptable to the NRC staff regarding core inventory and the release of Conforms. No fuel damage. Release radionuclides from the fuel are provided in Regulatory Position 3 of this guide. The estimate based on coolant release from the breached fuel is based on Regulatory Position 3.2 of this guide and the activity.

estimate of the number of fuel rods breached.

2 If no or minimal fuel damage is postulated for the limiting event, the released activity Conforms.

should be the maximum coolant activity allowed by technical specification. The iodine concentration in the primary coolant is assumed to correspond to the following two cases in the nuclear steam supply system vendor's standard technical specifications.

2.1 The concentration that is the maximum value (typically 4.0 [tCi/gm DE 1-131) permitted Conforms. 2.0 [tCi/g DE l-131 for and corresponds to the conditions of an assumed pre-accident spike, and the pre-accident spike; concentration that is the maximum equilibrium value (typically 0.2 [Ci/gm DE 1-131) corresponds to maximum permitted for continued full power operation. technical specification limit.

3 The activity released from the fuel should be assumed to mix instantaneously and Conforms.

homogeneously in the reactor coolant. Noble gases should be assumed to enter the steam phase instantaneously.

4.1 The main steam line isolation valves (MSIV) should be assumed to close in the Conforms.

Enclosure I Page A-20 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Conformance with Regulatory Guide 1.183 - Appendix D (MSLB)

App Sec RG Position HNP Position Comments maximum time allowed by technical specifications.

4.2 The total mass of coolant released should be assumed to be that amount in the steam line Conforms.

and connecting lines at the time of the break plus the amount that passes through the valves prior to closure.

4.3 All the radioactivity in the released coolant should be assumed to be released to the Conforms for For doses to control room, atmosphere instantaneously as a ground-level release. No credit should be assumed for offsite doses. which is located within the plateout, holdup, or dilution within facility buildings. turbine building, it is conservatively assumed that activity is released directly into the turbine building, thereby providing a direct inleakage pathway to the control room.

4.4 The iodine species released from the main steam line should be assumed to be 95% CsI Conforms.

as an aerosol, 4.85% elemental, and 0.15% organic.

Enclosure I Page B-I Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment APPENDIX B: Design Inputs and Assumptions for DBA Analyses Parameter Current Value AST Design Basis Value Design iInputsad-Assumpti -Com-Common to MultipleDBAAnalyses -

Source Terms e Fo HNP 24-month bounding fission product inventory Core Fission Product Inventory GE generic fission product inventory calculated using ORIGEN2 and multiplied by 1.1 Volumes and Dimensions Minimum Torus Air Volume 1.1 3E5 ft3 for Unit 2 1.10E5 ft3 for Unit 2 RB Volume Not used 1.38E6 ft3 (Unit 1) 1.30E6 ft3 (Unit 2)

TB Free Volume Not used 6.5E6 ft3 Condenser Volume 8.32E4 ft3 1.72E5 ft3 combined volume of low-pressure turbine and condenser Control Room External Not used 148ftx66ftx 16ft Dimensions Control Room External Shielding Not used 2 ft thick concrete walls, 2.5 ft thick concrete roof MCR / TSC Ventilation MCR Filtered Intake Rate 400 cfm 250 cfm Limiting MCR Unfiltered Inleakage (LOCA is Limiting 110 cfm 115 cfm (for LOCA, CRDA, and MSLB)

DBA)

Page B-2 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Parameter Current Value AST Design Basis Value 99% for all isotopes except noble gases for intake 90% for all isotopes except noble gases for intake and TSC Filter Efficiency and recirculation recirculation TSC Unfiltered Inleakage 0 cfm 10,000 cfm TB Ventilation TB Fan Start Time Not used 9 hr after accident initiation TB Exhaust Rate Not used 15,000 cfm Atmospheric Dispersion Unit 1:

9.90E-4 sec/m 3 (0 - 2 hr) 3.97E-4 (2- 8 hr) 4.30E-4 (8 - 24 hr) Both Units:

3.22E-4 (24 - 96 hr) 1.41 E-3 sec/m 3 (0 - 2 hr)

MCR and TSC x/Q for RB Vent 2.62E-4 (96 - 720 hr) 1.08E-3 (2 - 8 hr)

Release at Ground Level Unit 2: 4.70E-4 (8 - 24 hr) 1.26E-3 sec/M 3 (0 - 2 hr) 3.54E-4 (24 - 96 hr) 3.87E-4 (2 - 8 hr) 2.67E-4 (96 - 720 hr) 4.17E-4 (8 - 24 hr) 3.56E-4 (24 - 96 hr) 2.37E-4 (96 - 720 hr)

Both units: Both units:

4.85E-6 sec/m 3 (0 - 2 hr) 3.76E-6 sec/m 3 (0 - 2 hr)

MCR and TSC X/Q for Release 1.17E-6 (2- 8 hr) 2.88E-6 (2 - 8 hr)

Through Plant Stack 9.69E-7 (8 - 24 hr) 7.50E-7 (8 - 24 hr) 8.27E-7 (24 - 96 hr) 7.67E-7 (24 - 96 hr) 5.49E-7 (96 - 720 hr) 5.04E-7 (96 - 720 hr)

LOCA Inputs. - -----------

Enclosure I Page B-3 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Parameter Current Value AST Design Basis Value SGTS Flow Rate Not used 8,000 cfm (assumes 2 fans aligned to single RB volume) 1.2% per day (0-24 hr)

Primary Containment Leakage 1.2% per day

• Reduce by 40% (24-72 hr)

• Reduce by 50% (>72 hr)

Secondary Containment Bypass 0.9% of primary containment leakage 2.0% of primary containment leakage Leakage ESF Leakage 0 gpm (not modeled) 10 gpm 250 scfh 100 scfh total (all modeled from failed line)

MSIV Leakage

• Reduce by 40% (24-72 hr)

• Reduce by 50% ( > 72 hr)

Condenser Leakage 6.8% per day Mass balance based on flow into condenser DW Spray Start Time Not used 15 min after accident initiation Holdup in TB Not used Yes (for MCR doses)

Use of KI Tablets Taken by operators to lower thyroid dose Not used FHA Inputs----------.-- -

Number of Fuel Rods per Bundle 62 87.3 Number of Fuel Rods with Cladding Failure 125 172 Minimum Depth of Water Above - 21 ft Damaged Fuel Holdup and Filtration in Secondary Yes Both cases evaluated; holdup and filtration in secondary Containment containment is not required Page B-4 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Safety Assessment Parameter Current Value AST Design Basis Value EAB and LPZ X/Q for Leakage at 4.1 E-4 sec/r 3 3 3.1E-4 sec/r Ground Level (0 - 120 sec)

CRDA-In puts - --- ------ --------

Number of Fuel Rods per Bundle 62 87.3 Number of Fuel Rods with Cladding Failure Number of Fuel Rods with Melting 7 11 Radionuclide Percentage Released 100% noble gases 100% noble gases from Vessel 10% iodines 10% iodines 1% particulates Radionuclide Percentage Released 100% noble gases 100% noble gases from Turbine / Condenser 10% iodines 10% iodines 1% particulates Turbine / Condenser Leak Rate 0.5% per day 1% per day Holdup in TB Not used Yes (for MCR doses)

MSLB Inputs- -

Dose Conversion Factors Used to RG 1.109 inhalation thyroid FGR 11 inhalation effective Calculate DE 1-131 Iodine Activity 4.0 lICi/g DE 1-131 (Pre-accident I spike) 2.0 [tCi/g DE 1-131 (Pre-accident I spike) 0.2 ItCi/g DE 1-131 (Equilibrium I activity) 0.2 gCi/g DE 1-131 (Equilibrium I activity)

Holdup in TB Not used Yes (for MCR doses)

EAB and LPZ x/Q for Leakage at 4.1 E- sec/rn 3 3.1E-4 sec/r 3 Ground Level (0 - 1 hr)

Enclosure 2 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Description and Justification for Proposed Change

Enclosure 2 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Description and Justification for Proposed Change Proposed Change Southern Nuclear Operating Company (SNC) requests a revision of the Edwin I. Hatch Nuclear Plant (HNP) Operating License by revising the Technical Specifications (TSs) and incorporating an alternative source term (AST) methodology into the facility's licensing basis. The proposed license amendment involves a full scope implementation of an AST methodology by revising the current accident source term and replacing it with an accident source term as prescribed in 10 CFR 50.67.

AST analyses were performed using the guidance provided by Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000, and Standard Review Plan (SRP) Section 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms." The AST analyses include determination of the onsite, specifically the main control room and technical support center, and offsite radiological doses resulting from the HNP limiting design basis accidents (DBAs). The four DBAs considered were the loss-of-coolant accident (LOCA), the fuel handling accident, the control rod drop accident, and the main steam line break accident. The analyses demonstrate that, using AST methodologies, the post-accident onsite and offsite doses remain within regulatory acceptance limits.

As a result of the application of a revised accident source term, the following changes to the Technical Specifications are proposed:

1. The definition of dose equivalent 1-131 (DE 1-131) is revised to incorporate the updated reference for the dose conversion factors used in the DE 1-131 calculation on Units 1 and 2.

2. The maximum allowed reactor coolant specific activity is revised from 4.0 ptCi/gm DE 1-131 to 2.0 tCi/grn DE 1-131 on Units I and 2.

3. A Unit 1 Technical Specification on secondary containment bypass leakage is added, consistent with the current licensing basis on Unit 2, and the maximum allowed bypass leakage rate is conservatively increased from 0.9%

to 2.0% of the maximum allowable primary containment leakage rate (La) to allow for newly identified secondary containment bypass leakage paths on both Units 1 and 2.

4. Maximum allowed combined main steam line isolation valve (MSIV) leakage rates are revised by increasing the Unit I limit to 100 scfh and decreasing the Unit 2 limit to 100 scfh, and by eliminating the per line leakage rate limit. In addition, two separate surveillance acceptance criteria are provided dependant on leakage rate test pressure. Finally, the requirement to restore MSIV leakage to 11.5 scfh or less following discovery of MSIV leakage not meeting the acceptance criterion has been eliminated.

Enclosure 2 Page 2 of 7 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Description and Justification for Proposed Change

5. A new Technical Specification for residual heat removal drywell spray is added on Units 1 and 2. Drywell sprays are credited for the reduction of activity in the containment atmosphere as well as pressure and temperature reduction following a LOCA.

6. Changes are being made to the Technical Specification Bases to reflect AST implementation.

Background

On December 23, 1999, the NRC published 10 CFR 50.67, "Accident Source Term," in the Federal Register. This regulation provides a mechanism for licensed power reactors to replace the current accident source term used in DBA analyses with an alternative source term. 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 must apply for a license amendment under 10 CFR 50.90.

Regulatory Guide 1.183 and SRP Section 15.0.1 were used by SNC in preparing the AST analyses. These documents were prepared by the NRC staff to address the use of ASTs at existing operating power reactors. The regulatory guide establishes the parameters of an acceptable AST and identifies the significant attributes of an AST acceptable to the NRC staff. In this regard, the regulatory guide provides guidance to licensees on acceptable applications for an AST; the scope, nature, and documentation of associated analyses and evaluations; consideration of impacts on risk; and acceptable radiological analysis assumptions. The SRP provides guidance to the staff on the review of AST submittals.

Acceptance criteria consistent with that required by 10 CFR 50.67 were used to replace the HNP current design basis source term acceptance criteria. The AST analyses were performed for the four limiting DBAs that could potentially result in control room and offsite doses.

The HNP current licensing basis allows for the administration of potassium iodide (KI) to be credited to reduce the 30-day post-accident thyroid radiological dose to the operators in the main control room (MCR) for an interim period of approximately 4 years. This interim licensing basis was requested in preparation for MCR tracer gas inleakage testing, in order to accommodate a range of potential tracer gas inleakage test results.

Implementation of the AST will allow the administration of KI in the HNP interim licensing basis to be retired, as well as provide a significant increase in margin for MCR unfiltered inleakage assumed in the radiological dose DBA analyses.

Justification of Technical Specification Changes

1. Technical Specification 1.1, Definitions: DOSE EQUIVALENT 1-131 Current Technical Specification Current Technical Specification 1.1 provides a definition of "Dose Equivalent I-131," and includes references to dose conversion factors listed in Table III of TID-1 4844, AEC, 1962, "Calculation of Distance Factors for Power and Test Page 3 of 7 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Description and Justification for Proposed Change Reactor Sites"; Table E-7 of Regulatory Guide 1.109, Rev. 1, 1977; and ICRP 30, Supplement to Part 1, pages 192-212, Table titled "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity," for calculating thyroid DE 1-131.

ProposedChange The current Technical Specification 1.1 definition of "Dose Equivalent 1-131" is revised to replace the "thyroid dose" with "committed effective dose equivalent" and to reference only Federal Guidance Report (FGR) 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1988 for the dose conversion factors used in calculating DE 1-131.

The proposed revised Technical Specification 1.1 definition is:

DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same Committed Effective Dose Equivalent as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The dose conversion factors used for this calculation shall be those listed in Federal Guidance Report (FGR) 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1988.

Justilication The existing definition is revised to conform to the implementation of AST. The new citation of dose conversion factors is cited in Regulatory Guide 1.183, which has been found to be acceptable by the NRC for AST applications. Inhalation committed effective dose equivalent dose conversion factors used in this calculation are from FGR 11.

With the implementation of AST, the accident dose guidelines of 10 CFR 100 are superseded by the dose criteria of 10 CFR 50.67. The whole body and thyroid doses of 10 CFR 100 are replaced by the total effective dose equivalent (TEDE) criteria of 10 CFR 50.67. A conforming change to the definition is to replace "thyroid dose" in the definition with "committed effective dose equivalent." The analyses performed in support of this amendment request determined radiological consequences in terms of the TEDE dose quantity and were shown to be in compliance with the dose criteria of 10 CFR 50.67. These changes to the definition are acceptable because they reflect adoption of the dose conversion factors and the dose consequences of the revised radiological analyses.

2. Technical Specification 3.4.6, Reactor Coolant System Specific Activity Current TechnicalSpecification The maximum limit for RCS specific activity in current Technical Specification Page 4 of 7 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Description and Justification for Proposed Change 3.4.6 Conditions A and B is 4.0 pCi/gm DE 1-131.

ProposedChange The maximum limit for RCS specific activity in Technical Specification 3.4.6 Conditions A and B is revised from 4.0 [tCi/gm DE I-131 to 2.0 pCi/gm DE I-131.

Justification This revision is required to meet control room dose regulatory limits for the main steam line break accident. A lower maximum allowable RCS specific activity is more conservative than the existing limit. Typical values of DE 1-131 RCS specific activity are well below the new limit, and adequate operating margin is maintained.

3. Technical Specification 3.6.1.3, Primary Containment Isolation Valves Current Technical Specification As stated in surveillance requirement (SR) 3.6.1.3.10 for Unit 2, the maximum combined leakage rate for all secondary containment bypass leakage paths is 0.00 9 La, where La is the maximum allowable primary containment leakage rate,.

For Unit 1, there is no Technical Specification on maximum leakage rate for secondary containment bypass leakage.

Proposed Change For Unit 2, the current Technical Specification SR is revised to increase the maximum combined leakage rate for all secondary containment bypass leakage paths from 0.009La to 0.02L.. For Unit 1, new Technical Specification SR 3.6.1.3.13 is added. SR 3.6.1.3.13 establishes a maximum combined leakage rate for all secondary containment bypass leakage paths of 0.02L,.

Justification The secondary containment bypass leakage rate assumptions in the radiological dose consequence analysis for the LOCA form the basis for the revised Technical Specification limits. The proposed secondary bypass leakage rate limit of 0.02La is acceptable since this value was assumed in the accident analysis and regulatory criteria have been met. Because calculated doses are below the regulatory limits of 10 CFR 50.67, additional leakage margin exists.

The increase in bypass leakage is necessary to allow for newly identified bypass leakage paths. The addition of this Technical Specification SR to Unit I reflects a required Regulatory Guide (RG) 1.183 assumption in the accident analyses and standardizes the Technical Specifications between units.

4. Technical Specification 3.6.1.3, Primary Containment Isolation Valves Current Technical Specification Unit I Technical Specification SR 3.6.1.3.10 specifies a maximum leakage rate of 11.5 scfh through each MSIV when tested at 28.0 psig or greater. Unit 2 Page5 of`7 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Description and Justification for Proposed Change Technical Specification SR 3.6.1.3.11 specifies a maximum leakage rate of 100 scfh through each MSIV, and a combined maximum pathway leakage rate of 250 scfh, when tested at 28.8 psig or greater.

Unit 2 Technical Specification SR 3.6.1.3.11 imposes the leakage rate acceptance criteria of 11.5 scfh maximum leakage through an MSIV upon discovery of leakage not meeting the 100 scfh limit.

ProposedChange The per line MSIV leakage rate limits are eliminated from the Technical Specification SR for both units (SR 3.6.1.3.10 for Unit 1, and SR 3.6.1.3.11 for Unit 2). The Unit I combined maximum leakage rate is established at 100 scfh when tested at > 28.0 psig and < 50.8 psig. The Unit 2 combined maximum leakage rate is reduced from 250 scfh to 100 scfh when tested at > 28.8 psig and

< 47.3 psig. The pressure values of 50.8 psig and 47.3 psig represent calculated peak drywell pressures for Unit I and Unit 2, respectively, in the event of a LOCA.

A second test pressure range, with a corresponding leakage rate criterion, is established for both units when test pressure exceeds the peak calculated drywell pressure during a LOCA. This is in addition to the 100 scfh combined maximum leakage rate specification when tested within the specified test pressure range that is below the calculated peak drywell pressure. For Unit 1, a combined maximum leakage rate of 144 scfh is established when tested at > 50.8 psig. For Unit 2, a combined maximum leakage rate of 144 scfh is established when tested at > 47.3 psig.

For Unit 2, the requirement to restore MSIV leakage to 11.5 scfh upon discovery of leakage not meeting the 100 scfh leakage rate limit is eliminated.

Justification The revised values for MSIV combined maximum leakage rates are used in the radiological dose consequence analysis for the LOCA. The contribution to total combined leakage from any individual MSIV is not considered in the analysis.

Conservatively, the analysis assumes that the maximum allowed combined leakage rate is entirely through one MSIV. The proposed leakage rates are acceptable since this value was assumed in the revised accident analysis, and calculated doses are below the regulatory criteria of 10 CFR 50.67. In addition, because calculated doses are below the regulatory criteria, additional leakage margin exists.

The addition of a second MSIV leakage rate criterion for testing at or above calculated peak drywell pressure provides a more accurate leakage rate acceptance criterion for test pressures that are higher than calculated post-LOCA peak drywell pressures. This facilitates testing the MSIVs in the accident direction at peak accident drywell pressure as preferred by 10 CFR 50 Appendix J, as opposed to testing the MSIVs in the reverse direction at a lower test pressure as allowed by existing HNP Appendix J exemptions. A higher pressure results in Page 6 of 7 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Description and Justification for Proposed Change a higher mass flow rate through a given leakage area. The higher leakage rate (mass flow rate) acceptance criterion is based on a pressure and mass flow rate analysis. This allows for the use of different MSIV Appendix J test configurations as dictated by plant configuration during the outage, while also ensuring that the appropriate acceptance criterion exists for the actual test pressure used.

The elimination of the requirement to restore MSIV leakage to 11.5 scfh is acceptable since it is not an input or assumption in the radiological dose consequence analysis. In addition, this restoration is an overly-restrictive maintenance burden. The disadvantages of increased maintenance and higher worker radiation exposure associated with restoring MSIV leakage rates to relatively low values are not justified by any additional conservatism that might apply.

5. Technical Specification 3.6.2.5, Residual Heat Removal Drywell Spray Current TechnicalSpecification There is currently no Technical Specification for drywell spray.

Proposed Change Technical Specification 3.6.2.5, Residual Heat Removal (RHR) Drywell Spray, is added for both Unit I and Unit 2. The Technical Specification, limiting condition for operation (LCO), applicability, action statements, and SRs are patterned after existing Technical Specification 3.6.2.4, RHR Suppression Pool Spray. The proposed LCO requires two RHR drywell spray subsystems to be operable.

Justification The LOCA radiological dose analysis credits the use of drywell sprays for the reduction of airborne activity in the drywell by scrubbing inorganic iodines and particulates from the primary containment atmosphere. In addition, drywell spray is credited in the LOCA analysis for reducing the temperature and pressure in the drywell over time, thereby reducing the post-LOCA primary containment and MSIV leakage to within the assumptions of the dose analysis. Drywell spray is assumed to be manually initiated. Initiation is based on radiation levels in the drywell. By requiring two RHR drywell spray subsystems to be operable, this will ensure that in the event of a design basis LOCA, at least one subsystem will be operable assuming the worst case single active failure.

It should also be noted that the surveillance frequency for new Technical Specification SR 3.6.2.5.2 is "following maintenance which could result in nozzle blockage," unlike the 10 year frequency used for the similar surveillance for RHR suppression pool spray. Given the location of the spray headers in the drywell and prior demonstration of system operability, nozzle blockage is considered unlikely except as a consequence of maintenance or repair. This SR frequency has been approved by the NRC for Perry for containment spray.

Page 7 of 7 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Description and Justification for Proposed Change

6. Technical Specification Bases Current TechnicalSpecification Bases The Technical Specification Bases provide explanation and rationale for associated Technical Specification requirements, and in some cases, how they are to be implemented.

ProposedChang-e In addition to the Technical Specification Bases changes associated with the above mentioned Technical Specification changes, other Technical Specification Bases changes are associated with changing the radiological dose limits reference from 10 CFR 100 to 10 CFR 50.67, changing the accident analysis methodology reference to AST RG 1.183, and reflecting the additional standby liquid control system function of buffering the suppression pool to preclude the re-evolution of iodine from the suppression pool water following a DBA LOCA.

Justification The accident dose guidelines of 10 CFR 100 are superseded by the dose criteria of 10 CFR 50.67. The whole body and thyroid doses of 10 CFR 100 are replaced by the TEDE criteria 10 CFR 50.67, and references to 10 CFR 100 are replaced with 10 CFR 50.67. This is a conforming change.

The buffering of the suppression pool is credited in the radiological dose analysis for the LOCA and the resultant calculated doses are below the regulatory criteria of 10 CFR 50.67. The remaining Bases changes reflect the use of the AST RG 1.183 accident analysis methodology.

Other changes were made to the Technical Specification Bases to conform to the changes being made to the associated Technical Specifications. The revisions to the Technical Specification Bases incorporate supporting information for the proposed Technical Specification changes. Bases do not establish actual requirements, and as such do not change technical requirements of the Technical Specifications. The Bases changes are therefore acceptable, since they administratively document the reasons and provide additional understanding for the associated Technical Specification requirements.

Enclosure 3 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term 10 CFR 50.92 Significant Hazards Evaluation and Environmental Assessment

Enclosure 3 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term 10 CFR 50.92 Significant Hazards Evaluation and Environmental Assessment Proposed Change Southern Nuclear Operating Company (SNC) requests a revision of the Edwin I. Hatch Nuclear Plant (HNP) Operating License by revising the Technical Specifications and incorporating an alternative source term (AST) methodology into the facility's licensing basis. The proposed license amendment involves a full scope implementation of an AST methodology by revising the current accident source term and replacing it with an accident source term as prescribed in 10 CFR 50.67.

AST analyses were performed using the guidance provided by Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000, and Standard Review Plan, Section 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms." The AST analyses include determination of the onsite, specifically the main control room and technical support center, and offsite radiological doses resulting from the HNP limiting design basis accidents (DBAs). The four DBAs considered were the loss-of-coolant accident (LOCA), the fuel handling accident, the control rod drop accident, and the main steam line break accident. The analyses demonstrate that, using AST methodologies, the post-accident onsite and offsite doses remain within regulatory acceptance limits.

As a result of the application of a revised accident source term, changes to the Technical Specifications are proposed that revise the definition of dose equivalent 1-131, revise the maximum allowed reactor coolant specific activity, revise secondary containment bypass leakage rates, revise main steam line isolation valve leakage rates, and add a requirement for residual heat removal drywell sprays. Changes are also proposed to Technical Specification Bases to reflect AST implementation.

In addition to Technical Specification changes, the AST implementation adds a post-LOCA suppression pool pH control function for the standby liquid control system, credits main steam piping and the main condenser to provide an alternate leakage treatment pathway for main steam isolation valve leakage for Unit 1 (previously approved for Unit 2), and credits the turbine building ventilation system for removing activity from the turbine building post-accident.

10 CFR 50.92 Evaluation In 10 CFR 50.92(c) the Nuclear Regulatory Commission (NRC) provides the following standards to be used in determining the existence of a significant hazards consideration:

.a proposed amendment to an operating license for a facility licensed under 50.21(b) or 50.22, or for a testing facility involves no significant hazards consideration, if operation of the facility in accordance with the proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in the margin of safety.

Enclosure 3 Page 2 of 3 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term 10 CFR 50.92 Significant Hazards Evaluation and Environmental Assessment SNC has reviewed the proposed amendment request and determined that its adoption does not involve a significant hazards consideration based upon the following discussion:

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

Adoption of the AST and those plant systems affected by implementing AST do not initiate DBAs. The AST does not affect the design or manner in which the facility is operated; rather, once the occurrence of an accident has been postulated, the new accident source term is an input to analyses that evaluate the radiological consequences. The implementation of the AST and changed Technical Specifications have been incorporated in the analyses for the limiting DBAs at HNP.

The structures, systems, and components affected by the proposed change are mitigative in nature, and relied upon after an accident has been initiated. Based on the revised analyses, the proposed changes to the Technical Specifications (including revised leakage limits) impose certain performance criteria which do not increase accident initiation probability. The proposed changes do not involve a revision to the parameters or conditions that could contribute to the initiation of a DBA discussed in Chapter 15 of the Unit 2 Final Safety Analysis Report. Therefore, the proposed change does not result in an increase in the probability of an accident previously identified.

Plant specific AST radiological analyses have been performed and, based on the results of these analyses, it has been demonstrated that the dose consequences of the limiting events considered in the analyses are within the regulatory guidance provided by the Nuclear Regulatory Commission for use with the AST. This guidance is presented in 10 CFR 50.67, Regulatory Guide 1.183, and Standard Review Plan, Section 15.0.1. Therefore, the proposed change does not result in a significant increase in the consequences of an accident previously evaluated.

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

Implementation of AST and associated changes does not alter or involve any design basis accident initiators. These changes do not affect the design function or mode of operations of systems, structures, or components in the facility prior to a postulated accident. Since systems, structures, and components are operated essentially no differently after the AST implementation, no new failure modes are created by this proposed change. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant decrease in the margin of safety?

The changes proposed are associated with a revision to the licensing basis for HNP.

Approval of the licensing basis change from the original source term to the AST is requested by this application for a license amendment. The results of the accident analyses revised in support of the proposed change are subject to the acceptance Page 3 of 3 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term 10 CFR 50.92 Significant Hazards Evaluation and Environmental Assessment criteria in 10 CFR 50.67. The analyzed events have been carefully selected, and the analyses supporting these changes have been performed using approved methodologies and conservative inputs to ensure that analyzed events are bounding and safety margin has been retained. The dose consequences of these limiting events are within the acceptance criteria presented in 10 CFR 50.67, Regulatory Guide 1.183, and Standard Review Plan 15.0.1. Therefore, because the proposed changes continue to result in dose consequences within the applicable regulatory limits, the changes are considered to not result in a significant reduction in the margin of safety.

Environmental Assessment SNC has evaluated the proposed changes and determined the changes do not involve (1) a significant hazards consideration, (2) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (3) a significant increase in the individual or cumulative occupational exposure. Accordingly, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9), and an environmental assessment of the proposed changes is not required.

Enclosure 4 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Regulatory Safety Analysis

Enclosure 4 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Regulatory Safety Analysis Southern Nuclear Operating Company (SNC) requests a revision of the Edwin I. Hatch Nuclear Plant (HNP) Operating Licenses to revise the Technical Specifications and incorporate an alternative source term (AST) methodology into the facility's licensing basis. The proposed license amendment involves a full scope implementation of an AST methodology by revising the current accident source term and replacing it with an accident source term as prescribed in 10 CFR 50.67.

The AST analyses include determination of the onsite, specifically the main control room and technical support center, and offsite radiological doses resulting from the HNP limiting design basis accidents (DBAs). The four DBAs considered were the loss-of-coolant accident, the fuel handling accident, the control rod drop accident, and the main steam line break accident. The analyses demonstrate that, using AST methodologies, the post-accident onsite and offsite doses remain within regulatory acceptance limits.

The application of a revised accident source term and the application of an AST methodology have resulted in several proposed changes to the Technical Specifications, as well as changes to inputs and assumptions in the current design basis analyses.

The proposed license amendment will comply with 10 CFR 50.67. The radiological dose consequence analyses have shown that the dose criterion, 25 rem total effective dose equivalent (TEDE), prescribed for (1) an individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, and (2) an individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), has been met. Additionally, the analyses have demonstrated that adequate radiation protection would be provided to 'permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem TEDE for the duration of the accident.

The proposed license amendment will comply with 10 CFR 50 Appendix A General Design Criterion (GDC) 19. GDC 19 requires maintaining the control room in a safe, habitable condition under accident conditions, including loss-of-coolant accidents. Radiological dose consequence analyses have demonstrated that adequate radiation protection is provided to permit access and occupancy of the control room under postulated accident conditions, and that the radiation exposures would not exceed 5 rem TEDE for the duration of the accident.

The proposed license amendment will conform to Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000. RG 1.183 provides regulatory guidance on acceptable applications of ASTs; the scope, nature, and documentation of associated analyses and evaluations; consideration of impacts on analyzed risk; and content of submittals. This guide establishes an acceptable AST and identifies the significant attributes of an AST that may be found acceptable by the NRC staff. RG 1.183 also identifies acceptable radiological analysis assumptions for use in conjunction with the accepted AST.

The analyzed events have been carefully selected. The analyses supporting the proposed changes have been performed using approved methodologies and conservative inputs to ensure that analyzed events are bounding and safety margin has been retained. The dose consequences of these limiting events are within the acceptance criteria presented in 10 CFR 50.67, RG 1.183, and Standard Review Plan Section 15.0.1.

Page 2 of 2 Edwin I. Hatch Nuclear Plant Request to Implement an Alternative Source Term Regulatory Safety Analysis Therefore, the proposed changes to Technical Specifications and a full scope implementation of an AST methodology contained within the license amendment request comply with the applicable regulatory requirements and guidance.

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) issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.