NOC-AE-07002127, Request for License Amendment Related to Application of the Alternate Source Term

Request for License Amendment Related to Application of the Alternate Source Term

ML070890474
Person / Time
Site:	South Texas
Issue date:	03/22/2007
From:	Rencurrel D South Texas
To:	Document Control Desk, NRC/NRR/ADRO
References
NOC-AE-07002127
Download: ML070890474 (327)

Text

Attachment 8 is considered Not for Public Disclosure in accordance with 10CFR 2.390 Nuclear Operating Company South Texas Projed Electn'c GeneratnStation P.. Box 28 Wdsrth, Texas 77483 -

March 22, 2007 NOC-AE-07002127 10CFR50.67 10CFR50.90 U. S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738 South Texas Project Units 1 and 2 Docket Nos. STN 50-498, STN 50-499 Request for License Amendment Related to Application of the Alternate Source Term

References:

1. U. S. Nuclear Regulatory Commission, Regulatory Guide 1.183, "Alternative Radiological Source Terms for-Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000

2. U. S. Nuclear Regulatory Commission Standard Review Plan 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms,"

Revision 0, July 2000 In accordance with 10 CFR 50.67, "Accident Source Term," and 10 CFR 50.90, "Application for amendment of license or construction permit," STP Nuclear Operating Company (STPNOC) requests an amendment to Appendix A, Technical Specifications (TS), of Facility Operating Licenses NPF-76 and NPF-80 for South Texas Project (STP) Units 1 and 2. The proposed changes are requested to support application of an alternative source term (AST) methodology, with the exception that Technical Information Document (TID) 14844, "Calculation of Distance Factors for Power and Test Reactor Sites," will continue to be used as the radiation dose basis for equipment qualification. The proposed AST methodology is consistent with the guidance in References 1 and 2, including alternate methods for complying with the specified portions of the NRC's regulations as allowed by the guidance in Reference 1. Documentation of conformance to Reference 1 and the allowed alternate methods are presented in Tables A through G in of this submittal.

Application of the AST methodology is being used to resolve a non-conforming condition at STP where testing resulted in control room unfiltered in-leakage at a value greater than the current accident dose analysis assumption. The application of the AST methodology also allows for cost beneficial changes resulting from the relaxation of certain Technical Specification requirements.

Attachment 8 is considered Not for Public Disclosure in accordance with 10CFR 2.390 STI: 32126563 /f-ool

NOC-AE-07002127 Page 2 Implementation of an AST methodology changes the regulatory assumptions regarding the analytical treatment of the design basis accidents (DBAs). Approval of this proposed change will provide a source term for the STP facility that will result in a more accurate assessment of the DBA radiological doses.

By a separate action, STPNOC plans to submit a revision to TS 3/4.7.7, "Control Room Makeup and Cleanup Filtration System." This revision will be based on TS Task Force (TSTF) Traveler 448, Revision 3 that includes a surveillance program for measuring control room inleakage. The inleakage acceptance criterion in the surveillance is the value assumed in the design basis accident analyses. Implementation of the separate TS based on TSTF 448 will be predicated on approval and implementation of this AST amendment because the AST amendment establishes the revised inleakage assumption in the design basis accident analyses that is required for implementation of the TSTF 448 amendment.

STPNOC identified the following condition related to this amendment request. Westinghouse Electric Company Nuclear Safety Advisory Letter NSAL-06-15, dated December 13, 2006, advised operators of Westinghouse plants that the single failure scenario for the steam generator tube rupture (SGTR) analysis that licensees used in their accident analysis may not be limiting.

STPNOC has evaluated the applicability of this NSAL against the accident analysis assumptions.

'The current single failure assumption for the STP SGTR analysis is not limiting so that STPNOC is operating under compensatory measures to meet regulatory dose limits. STPNOC plans to resolve this condition at the earliest opportunity so that the assumptions, including the limiting single failure, for the SGTR accident analysis performed for this amendment request are consistent with the plant response to this event. Section 1.0 of Attachment 1 to this letter provides additional detail. The action to resolve this condition is provided as licensing commitment in Attachment 5 to this letter.

This request is subdivided as follows:

• Attachment 1 provides the Licensee's Evaluation for this change including a description of proposed changes, technical analysis, and regulatory analysis.

" Attachment 2 provides the markup of Technical Specification (TS) pages.Attachment 3 provides a markup of the associated TS Bases pages (for information only)

" Attachment 4 provides planned changes to the Technical Requirements Manual (for information only)

" Attachment 5 provides the List of Commitments resulting from the proposed changes.

" Attachment 6 contains "Regulatory Guide 1.183 Conformance Tables" providing detailed verification that the AST methodology conforms to the guidance in Regulatory Guide 1.183.

" Attachment 7 addresses information discussed in NRC Regulatory Issue Summary (RIS) 2006-04. A table provides a description of how each issue discussed in the RIS is addressed in the STPNOC application.

NOC-AE-07002127 Page 3

• Attachment 8 provides a markup of the affected pages to the STP Updated Final Safety Analysis Report (UFSAR) (for information only)

The Plant Operations Review Committee has approved the proposed change. STPNOC has notified the State of Texas in accordance with 10CFR50.91(b). A No Significant Hazards Consideration Determination is provided in Attachment 1.

STPNOC requests approval of the proposed amendment by March 30, 2008. Once approved, the amendment shall be implemented within 120 days due to the significant implementation scope of the subject changes. This implementation period will provide adequate time for the affected station documents to be revised using the appropriate change control mechanisms.

The NRC previously approved implementation of the AST methodology at a number of other nuclear power stations. This change is consistent with the guidance of NRC Regulatory Guide 1.183 and Standard Review Plan 15.0.1.

If there are any questions regarding this proposed amendment, please contact Mr. Ken Taplett at (361) 972-8416 or me at (361) 972-7867.

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

Executed on OQGQ N, 9A a.'067 0avid W: Rencurrel Vice President, Engineering and Strategic Projects

NOC-AE-67002127 Page 4 Attachments:

1. Licensee's Evaluation

2. Markup of Technical Specification pages

3. Markup of Technical Specification Bases pages (information only)

4. Planned Changes to the Technical Requirements Manual (information only)

5. List of Commitments

6. Regulatory Guide 1.183 Conformance Tables

7. NRC Regulatory Issue Summary 2006-04 Table

8. Markup of UFSAR pages (information only) (Not for Public Disclosure)

NOC-AE-07002127 Page 5 cc:

(paper copy) (electronic copy)

Regional Administrator, Region IV A. H. Gutterman, Esquire U. S. Nuclear Regulatory Commission Morgan, Lewis & Bockius LLP 611 Ryan Plaza Drive, Suite 400 Arlington, Texas 76011-8064 Mohan C. Thadani U. S. Nuclear Regulatory Commission Richard A. Ratliff Steve Winn Bureau of Radiation Control Christine Jacobs Texas Department of State Health Services Eddy Daniels 1100 West 49th Street Marty Ryan Austin, TX 78756-3189 NRG South Texas LP Senior Resident Inspector J. J. Nesrsta U. S. Nuclear Regulatory Commission R. K. Temple P. 0. Box 289, Mail Code: MN116 Kevin Pollo Wadsworth, TX 77483 Ed Alarcon City Public Service C. M. Canady City of Austin Jon C. Wood Electric Utility Department Cox Smith Matthews 721 Barton Springs Road Austin, TX 78704 C. Kirksey City of Austin David Leaver Polestar Applied Technology, Inc.

One First Street, Suite 4 Los Altos, CA 94022 Jim Metcalf Polestar Applied Technology, Inc.

4 High Street, Suite 2 Hampton, NH 03842

NOC-AE-07002127 Attachment 1 Page 1 of 219 Attachment 1 Licensee's Evaluation

NOC-AE-07002127 Attachment 1 Page 2 of 219 Table of Contents

1.0 DESCRIPTION

......................................................... ..............11 2.0 PRO PO SED CH A N G E .................................................................................................... 14 3.0 B A CK G R OU N D ..................................................................................................................... 19 3.1 Systems Affected by the Proposed Change .......... 1............................

19 3.2 The A ST R ule ...................................................................................................................... 21 3.3 STP Full Application of AST ....................................................................................... 22 3.4 Precedent ............................................................................................................................. 22

4.0 TECHNICAL ANALYSIS

................................................................................................ 23 4.1 Meteorology and Atmospheric Dispersion ......................................................................... 27 4.1.1 Analysis of the 2000-2004 Meteorology Data ................................ 30 4.1.2 Determination of the EAB and LPZ x/Qs ............................................................. 45 4.1.3 Control Room and Technical Support Center x/Q Analyses .................................. 59 4 .2 An alytical M odels ................................................................................................................ 70 4.2.1 Offsite D ose M odel ................................................................................................. 70 4.2.2 Control Room Analytical Model ....... ............................. 71 4.2.2.1 CRE Unfiltered In-leakage and Possible "Sneak" Paths ..................................... 77 4.2.3 Technical Support Center (TSC) Analytical Model .............................................. 78 4.2.4 R adiological Source Term s ......................................................................................... 81 4.2.4.1 Dose Conversion Factors and Physical Parameters ........................................... 81 4.2.4.2 Reactor Core Source Terms ........................................................................... 83 4.2.4.2.1 Peak Pin Evaluation for non-LOCA Fuel Gap Inventory .......................... 86 4.2.4.3 Dose Equivalent 1-131 and Coolant Activity .................................................. 88 4.2.4.4 Reactor Coolant System Source Terms ............................................................. 91 4.2.4.4.1 RC S at 1% Failed Fuel ............................................................................... 91 4.2.4.4.2 RCS lodines at Normal Tech Spec Limit of 1 [tCi/gm ............................... 93 4.2.4.4.3 RCS lodines at Spiking Tech Spec Limit of 60 p.Ci/gm ........................... 93 4.2.4.4.4 RCS Cs and Rubidium Concentrations ......... ...................... 95 4.2.4.5 Secondary System Source Terms ..................................................................... 96 4.2.4.5.1 Secondary System Iodine Concentrations ...................................................... 96 4.2.4.5.2 Secondary System Noble Gas Concentrations .......................................... 97 4.2.5 Iodine Species Released from Steam Generators ................................................... 99 4.3 Loss of Coolant Accident Radiological Assessment ......................................................... 100 4.3.1 M ethodology O verview ............................................................................................. 100 4.3.2 Radiological Source Term ........................................................................................... 100 4.3.3 R adiological R eleases ................................................................................................. 103 4.3.3.1 Radiological Releases from the Containment ..................................................... 103 4.3.3.1.1 Release from the Containment Supplemental Purge Subsystem .................. 104 4.3.3.1.2 Containment Sump pH and Iodine Re-evolution ......................................... 104 4.3.3.1:2.1 Determination of Sump pH .............................. 104 4.3.3.1.2.2 Iodine R e-evolution ............................................................................... 109 4.3.3.2 Radiological Releases from ESF Equipment ....................................................... 115 4.3.4 R adiological D ose M odels .......................................................................................... 1 16 4.3.4.1 C ontrol Room and T SC ....................................................................................... 122

- NOC-AE-07002127 Attachment 1 Page 3 of 219 4.3.4.1.1 CR/TSC Doses from Airborne Contaminants .............................................. 122 4.3.4.2 CR/TSC Doses from Gamma Shine .................................................................... 122 4.3.5 Inputs and Assumptions .. *........................................ 129 4.3.6 Summary and Conclusions ....................................................................................... 137 4.4 Fuel Handling Accident (FHA) Radiological Assessment ................................................ 138 4.4.1 M ethodology Overview .............................................................................................. 138 4.4.2 A nalytical M odel ........................................................................................................ 138 4.4.3 Radiological Source Term .................................................................................. ......... 139 4.4.4 Radiological Releases ............................................................................. 141 4.4.5 A ssumptions and Inputs .............................................................................................. 141 4.4.6 Summary and Conclusions ....................................................... I................ I................ 143 4.4.7 C ore A lterations ......................................................................................................... 143 4.4.8 Shutdown Safety Assessment/Defense-in-Depth ...................................................... 144 4.5 Main Steam Line Break Radiological Assessment ............................................................ 146 4.5.1 Methodology Overview .............................................................................................. 146 4.5.1.1 Comparison of Modeling with Regulatory Guide 1.183, Appendix E ................ 147 4.5.2 Analytical M odel ....................................................................................................... 148 4.5.3 Radiological Source Term .......................................................................................... 151 4.5.3.1 Reactor Coolant System Source Term ............................... .................................. 151 4.5.3.1.1 RCS Iodine Concentrations .......................................................................... 151 4.5.3.1.2 RCS Noble Gas Concentrations ................................................................... 152 4.5.3.1.3 RCS Cesium and Rubidium Concentrations ..................... 152 4.5.3.2 Secondary System Source Terms ........................................................................ 152 4.5.3.2.1 Secondary System Iodine Concentrations ................................................... 152 4.5.3.2.2 Secondary System Noble Gas Concentrations .................... 152 4.5.4 R adiological R eleases ........... ..................................................................................... 153 4.5.5 A ssum ptions and Inputs ............................................................................................. 154 4.5.6 Summary and Conclusions ....................................... 158 4.6 Steam Generator Tube Rupture Radiological Assessment ................................................ 159 4.6.1 M ethodology Overview .............................................................................................. 159 4.6.1.1 Comparison of Modeling with Regulatory Guide 1.183, Appendix E ................ 160 4.6.2 Analytical M odel ........................................................................................................ 162 4.6.3 Radiological Source Term .......................................................................................... 164 4.6.3.1 Reactor Coolant System Source Term ................................................................. 164 4.6.3.1.1 RCS Iodine Concentrations .......................................................................... 164 4.6.3.1.2 RCS Noble Gas Concentrations ................................................................. 165 4.6.3.1.3 RCS Cesium and Rubidium Concentrations ................................................ 165 4.6.3.2 Secondary System Source Terms ........................................................................ 165 4.6.3.2.1 Secondary System Iodine Concentrations .................................................... 165 4.6.3.2.2 Secondary System Noble Gas Concentrations ............................................. 165 4.6.4 Radiological R eleases ................................................................................................. 165 4.6.4.1 Thermal/Hydraulic Analysis of the SGTR .............................. 166 4.6.4.2 Modification of the Thermal/Hydraulic Data for Dose Analysis ........................ 170 4.6.5 A ssumptions and Inputs .............................................................................................. 173 4.6.6 Summary and Conclusions ...................................................................................... .. 177

NOC-AE-07002127 Attachment 1 Page 4 of 219 4.7 Control Rod Ejection Radiological Assessment ............................ 178 4.7.1 M ethodology O verview .............................................................................................. 178 4.7.1.1 Comparison of Modeling with Regulatory Guide 1.183, Appendix E ....... 178 4.7.2 Analytical Model .............................................. 179 4.7.3 R adiological Source Term ......................................................................................... 180 4.7.3.1 R eactor C ore Releases ......................................................................................... 182 4.7.3.1.1 Release from Cladding Failures ................................... ......................... 182 4.7.3.1.2 Release from Fuel M elt ............................................................................... 183 4.7.3.2 Reactor Coolant System Source Terms ....................... ........................................ 183 4.7.3.2.1 RCS Iodine Concentrations ....... .......................... 183 4.7.3.2.2 RCS N oble Gas Concentrations ................................................................... 184 4.7.3.2.3 RCS Cesium and Rubidium Concentrations ..................... 184 4.7.3.3 Secondary System Source Term s ........................................................................ 184 4.7.3.3.1 Secondary System Iodine Concentrations .................. ..... 184 4.7.3.3.2 Secondary System Noble Gas Concentrations .................... 184 4.7.3.3.3 Secondary System Cesium and Rubidium Concentrations ............ 184 4.7.4 Radiological R eleases ................................................................................................. 184 4.7.4.1 Release from Containment Building Scenario .................................................... 185 4.7.4.2 Release via Secondary Side Scenario ............. ................ 186 4.7.5 A ssum ptions and Inputs .............................................................................................. 188 4.7.6 Sum m ary and Conclusions ......................................................................................... 193 4.8 Locked Rotor Accident Radiological Assessment ........................... 195 4.8.1 M ethodology O verview .............................................................................................. 195 4.8.1.1 Comparison of Modeling with Regulatory Guide 1.183, Appendix E ................ 196 4.8.2 Analytical Model .............................................. 197 4.8.3 Radiological Source Term .......................................................................................... 198 4.8.3.1 Reactor Coolant System Source Term ....... ..................... 199 4.8.3.1.1 RCS Iodine C oncentrations ............................................................  :!............ 199 4.8.3.1.2 RCS Noble Gas Concentrations .................................................................. 199 4.8.3.1.3 RCS Cesium and Rubidium Concentrations ................................................. 199 4.8.3.2 Secondary System Source Term s ....................................................................... 199 4.8.3.2.1 Secondary System Iodine Concentrations .................................................... 199 4.8.3.2.2 Secondary System Noble Gas Concentrations ............................................. 199 4.8.3.2.3 Secondary System Cesium and Rubidium Concentrations .......................... 199 4.8.3.3 Fuel Pin G ap Source ............................................................................................ 199 4.8.3.4 Total Source Available for Release..................................................................... 201 4.8.4 Radiological Releases ....................................................................................... I........... 202 4.8.5 A ssum ptions and Inputs .............................................................................................. 203 4.8.6 Summ ary and Conclusions ......................................................................................... 205 4.9 NUREG -0737 Evaluations ................................................................................................ 206 4 .10 C on clu sion ....................................................................................................................... 207 5.0 REGU LATO RY AN ALY SIS ............................................................ ................................... 208 5.1 N o Significant H azards Consideration .............................................................................. 208 5.1.1 Overview ........... 208 5 .1.2 C riteria ........................................................................................................................ 209

NOC-AE-07002127 Attachment 1 Page 5 of 219 5.2 Applicable Regulatory Requirements/Critenia .................................................................. 212

6.0 ENVIRONMENTAL CONSIDERATION

........................................................................... 215 7.0 REFEREN C E S ...................................................................................................................... 2 17

NOC-AE-07002127 Attachment 1 Page 6 of 219 Table of Tables Table 4.0-1: Frequently Used Acronyms ............................................................................... 26 Table 4.1 -1: Joint Frequency Distribution for 2000-2004 Summary of All Stability Classes ................................................................................................................... 32 Table 4.1-2: Joint Frequency Distribution for 2000-2004 Stability Class: A Extremely U nstab le ................................................................................................................. 32 Table 4.1-3: Joint Frequency Distribution for 2000-2004 Stability Class: B Moderately Unstable ................................................................................................................. 33 Table 4.1-4: Joint Frequency Distribution for 2000-2004 Stability Class: C Slightly Un stab le ................................................................................................................. 33 Table 4.1-5: Joint Frequency Distribution for 2000-2004 Stability Class: D Neutral ....... 34 Table 4.1-6: Joint Frequency Distribution for 2000-2004 Stability Class: E Slightly Stab le ..................................................................................................................... 34 Table 4.1-7: Joint Frequency Distribution for 2000-2004 Stability Class: F Moderately Stab le ..................................................................................................................... 35 Table 4.1-8: Joint Frequency Distribution for 2000-2004 Stability Class: G Extremely Stab le ..................................................................................................................... 35 Table 4.1-9: Stability Class Distribution ............................................................................... 36 Table 4.1-10: Average Wind Speed and Peak Wind Direction ............................................... 36 Table 4.1 -11: Wind Direction Distribution ............................................................................. 37 Table 4.1-12: Defined Wind Speed Category Ranges For PAVAN Modeling ....................... 47 Table 4.1-13: PAVAN Input: Joint Frequency Distribution for 2000-2004 Stability Class: A Extremely Unstable .......................................................................... 48 Table 4.1-14: PAVAN Input: Joint Frequency Distribution for 2000-2004 Stability Class: B Moderately Unstable ........................................................................... 49 Table 4.1-15: PAVAN Input: Joint Frequency Distribution for 2000-2004 Stability Class: C Slightly Unstable ............................................................................... 50 Table 4.1-16: PAVAN Input: Joint Frequency Distribution for 2000-2004 Stability Class: D N eutral ................................................................................................ 51 Table 4.1-17: PAVAN Input: Joint Frequency Distribution for 2000-2004 Stability Class: E Slightly Stable ................................................................................... 52 Table 4.1-18: PAVAN Input: Joint Frequency Distribution for 2000-2004 Stability Class: F Moderately Stable ............................................................................... 53 Table 4.1-19: PAVAN Input: Joint Frequency Distribution for 2000-2004 Stability Class: G Extremely Stable ............................................................................... 54 Table 4.1-20: PAVAN Input: Joint Frequency Distribution for 2000-2004 Summary of All Stability C lasses ........................................................................................... 55 Table 4.1-21: Relative Concentration (X/Q) Values (sec/mi3) Versus Averaging Time

@ E AB ................................................................................................................... 56 Table 4.1-22: Relative Concentration (X/Q) Values (sec/m 3) Versus Averaging Time

@ L P Z .................................................................................................................... 57

NOC-AE-07002127 Attachment 1 Page 7 of 219 Table of Tables (cont.)

Table 4.1-23: X/Q Values Based on DT(60M-10M) Stability Data and 10 Meter Winds ..... 58 Table 4.1-24: x/Q Values for Radiological Dose Calculations - EAB and LPZ ..................... 58 Table 4.1-25: Key to Figure 4.1-13: Release and Receptor Locations .................. 60 Table 4.1-26: Geometric Relationships Between Release Locations and Receptors ........ 62 Table 4.1-27: Data Used to Generate ARCON96 Inputs ........................................................ 63 Table 4.1-28: ARCON96 Input: Unit I Releases to Unit 1 Control RoomrTSC ..................... 64 Table 4.1-29: ARCON96 Input: Unit 2 Releases to Unit 2 Control Room/TSC .................... 65 Table 4.1-30: ARCON96 Results Unit 1 Containment to Unit 1 Control Room/TSC ........... 66 Table 4.1-3 1: ARCON96 Results Unit 1 Plant Vent to Unit 1 Control Room/TSC ............... 66 Table 4.1-32: ARCON96 Results Unit 1 East PORV to Unit 1 Control Room/TSC ............. 67 Table 4.1-33: ARCON96 Results Unit 2 Containment to Unit 2 Control Room/TSC ........... 67 Table 4.1-34: ARCON96 Results Unit 2 Plant Vent to Unit 2 Control Room/TSC ......... 68 Table 4.1-35: ARCON96 Results Unit 2 East PORV to Unit 2 Control Room/TSC ............. 68 Table 4.1-36: Control Room/TSC 95 Percentile X/Qs ............................................................ 69 Table 4.1-37: Summary of Control Room and TSC z/Q Values ............................................. 69

-Table 4.2-1: O ffsite Breathing Rates ..................................................................................... 71 Table 4.2-2: Control Room HVAC Flow Rates .................................................................... 76 Table 4.2-3: Parameters Used in Modeling the Control Room ...................... 77 Table 4.2-4: TSC H V AC Flow Rates ................................................................................... 79 Table 4.2-5: Parameters Used in M odeling the TSC ............................................................ 80 Table 4.2-6: Dose Conversion Factors .......................................................... 81 Table 4.2-7: Isotopic Half Lifes, Parent-to-Daughter Decay Isotopes and Fractions ........... 82 Table 4.2-8: Comparison of Revisions to the RadiationAnalysis Design Manual ............... 84 Table 4.2-9: Comparison of CLB and AST Reactor Core Sources ..................................... 84 Table 4.2-10: Isotopic Concentrations for 60 MCi/gm Representing An Equivalent Concentration of DE 1-131 for the Current Licensing Basis ............................. 88 Table 4.2-11 A Isotopic Concentrations Representing An Equivalent Concentration of DE 1-131 Using Updated Iodine RCS Concentrations and Thyroid DCFs ........ 89 Table 4.2-1 1B Isotopic Concentrations Representing An Equivalent Concentration of DE 1-131 Using Updated Iodine RCS Concentrations and CEDE DCFs ..... 89 Table 4.2-12: Proposed Isotopic Concentrations Representing An Equivalent Concentration of D E 1-131 .............................................................................. 90 Table 4.2-13: Proposed Isotopic Concentrations Representing An Equivalent Concentration Of DE 1-131 Using a TEDE DCF ............................................ 90 Table 4.2-14: Comparison of CLB and AST Reactor Coolant Sources @ 1% Failed Fuel ........ 91 Table 4.2-15: RCS Iodine Concentrations for 1% Failed Fuel ............................................... 94 Table 4.2-16: RCS Iodine Concentrations and DCFs ............................................................. 94 Table 4.2-17: RCS Iodine Concentrations for a Pre-existing Iodine Spike to 60 pCi/gm ..... 95 Table 4.2-18: Total RCS Cs and Rb Activity for a Pre-Accident Iodine Spike ...................... 95 Table 4.2-19: Secondary Iodine Concentrations at 0.1 pCi/gm............................................. 97 Table 4.2-20: Initial RCS (@60 pCi/gm) and Secondary Concentrations (@ 0.1 pCi/gm DEI) ............................................................................ 98 Table 4.3-1: LOCA: Reactor Core Fission Product Inventory @ t=0 ..................................... 101

NOC-AE-07002127 Attachment 1 Page 8 of 219 Table of Tables (cont.)

Table 4.3-2: Containment Sump pH Control Inputs ................................................................ 107 Table 4.3-3: Sump Concentrations and pH as a Function of Time .......................................... 109 Table 4.3-4: RA DTRAD M odels for LOCA ........................................................................... 117 Table 4.3-5: CR and TSC Gamma Shine Dose Analysis Inputs for DBA LOCA ................... 125 Table 4.3-6: Current Licensing Basis Activity on One Control Room Make-Up Filter .......... 126 Table 4.3-7: Gamma Power from Activity on Two Control Room Make-Up Filters .............. 127 Table 4.3-8: Alternative Source Term Gamma Power from Activity on Two Control R oom C leanup Filters ........................................................................................... 128 Table 4.3-9: Gam m a Shine Com ponent D oses ........................................................................ 128 Table 4.3-10: LOCA Time-Dependent Release Fractions ......................................................... 133 Table 4.3-11: Dose A nalysis Inputs for LO CA ....................................................................... 134 Table 4.3-12: CLB Spray Removal Param eters ......................................................................... 136 Table 4.3-13: A ST Spray Rem oval Param eters ......................................................................... 136 Table 4.3-14: LOCA Dose Results ............. 137 Table 4.4-1: Base Fission Product Gap Inventory for the FHA .......................... 140 Table 4.4-2: Fuel H andling Accident Inputs ............................................................................ 142 Table 4.4-3: Fuel Handling Accident Dose Results ............................ ..................................... 143 Table 4.5-1: RCS Iodine Inventory Due to Accident-Induced Spike (500x Production R ate).............................................................................................................. ...... 15 1 Table 4.5-2: Total RCS Cs and Rb Activity for an Accident-Induced Iodine Spike ............... 152 Table 4.5-3: Inputs for M SLB A nalysis ................................................................................... 156 Table 4.5-4: M SL B Dose R esults ............................................................................................ 158 Table 4.6-1: RCS Iodine Inventory Due to an 8-Hour Accident-Induced Spike (335x Produ ction Rate) .................................................................................................. 164 Table 4.6-2 Thermal Hydraulics Analysis Sequence of Events .............................................. 167 Table 4.6-3: Thermal Hydraulics Analysis Time Points used in the SGTR Dose A n alysis ................................................................................................................ 168 Table 4.6-4: Thermal Hydraulics Analysis Total Break Flow ................................................. 168 Table 4.6-5: Thermal Hydraulics Analysis Flashed Break Flow .................... 169 Table 4.6-6: Thermal Hydraulics Analysis Total Intact SG Steam Flow to Atmosphere ........ 169 Table 4.6-7: Thermal Hydraulics Analysis Total Ruptured SG Steam Flow to A tm osph ere ......................................................................................................... 170 Table 4.6-8: Modified Time Sequence of Events for SGTR Dose Analysis ........................... 170 Table 4.6-9: Total Break Flow used in SGTR Dose Analysis ................................................. 171 Table 4.6-10: Total Flashed Break Flow used in SGTR Dose Analysis .................................... 171 Table 4.6-11: Total Intact SG Flow to Atmosphere Used in SGTR Dose Analysis ................. 172 Table 4.6-12: Total Ruptured SG Flow to Atmosphere Used in SGTR Dose Analysis ............ 172 Table 4.6-13: Inputs for SG TR A nalysis ................................................................................... 175 Table 4.6-14: SG T R Dose R esults ............................................................................................. 177 Table 4.7-1: 10% Failed Fuel G ap Release Source .................................................................. 182 Table 4.7-2: 0.25% C ore M elt Source ..................................................................................... 183

NOC-AE-07002127 Attachment 1 Page 9 of 219 Table of Tables (cont.)

Table 4.7-3: Release From the RCB Scenario: Total Activity Releasedinto the RCB ........... 186 Table 4.7-4: Release From the Secondary Side Scenario: Total Activity in the Steam Generators (RCS+SG) ........................................ 187 Table 4.7-5: Steam Released to the Environment ................................................................... 188 Table 4,7-6: Inputs for CREA Analysis: Release from the RCB Scenario .............................. 190 Table 4.7-7: Inputs for CREA Analysis: Release from the Secondary Side Scenario ............. 192 Table 4.7-8: CREA Doses from Containment Leakage ........................................................... 193 Table 4.7-9: CREA Doses from Secondary Side Release ........................................................ 193 Table 4.7-10: Total CR EA D ose Results ................................................................................... 194 Table 4.8-1: 10% G ap Release Source .................................................................................... 200 Table 4.8-2: Total Source Available for Release (RCS+Sec) .................................................. 201 Table 4.8-3: Steam Released to the Environment .................................................................... 202 Table 4.8-4: Inputs for LRA A nalysis ...................................................................................... 204 Table 4.8-5: L RA D ose Results ............................................................................................... 205

NOC-AE-07002127 Attachment 1 Page 10 of 219 Table of Figures Figure 4.1-1: Comparison of Wind Speed Distribution for STPEGS .................................... 38 Figure 4.1-2: Wind Rose for 2000 .......................................... 38 Figure 4.1-3: W ind Rose for 2001 .......................................................................................... 39 Figure 4.1-4: W ind Rose for 2002 ........................................................................................... 39 Figure 4.1-5: W ind Rose for 2003 .......................................................................................... 40 Figure 4.1-6: W ind Rose for 2004 ........................................................................................... 40 Figure 4.1-7: Wind Rose for 2000-2004 ......................... ............. 41 Figure 4.1-8: Delta-T Frequency for 2000 ............................................................................ 42 Figure 4.1-9: Delta 7T Frequency for 2001 ............................................................................ 42 Figure 4.1-10: Delta-T Frequency for 2002 ............................................................................ 43 Figure 4.1-11: Delta-T Frequency for 2003 ............................................................................ 43 Figure 4.1-12: Delta-T Frequency for 2004 ............................................................................ 44 Figure 4.1-13: Simplified Plot Plan with Release Points and Receptors ................................. 59 Figure 4.2-1: Control Room Envelope ................................................................................... 75 Figure 4.2-2: Control Room HVAC Analytical Model ........................................................... 76 Figure 4.2-3: TSC HVAC Analytical M odel .............................................................................. 79 Figure 4.3-1: 12 Fraction vs pH ............................................................................................. 111 Figure 4.3-2: Iodine Decontamination Factor as a Function of pH ........................................... 112 Figure 4.3-3: Iodine Decontamination Factor in the Containment Sump ................................. 114 Figure 4.3-4: RADTRAD Model for [LOCA] Case 1 ............................................................. 118 Figure 4.3-5: RADTRAD Model for [LOCA] Case 1pen......................................................... 118 Figure 4.3-6: RADTRAD Model for [LOCA] Case lesf .......................................................... 119 Figure 4.3-7: RADTRAD Model for [LOCA] Case 1pur ......................................................... 119 Figure 4.3-8: RADTRAD Model for [LOCA] Case 2 .............................................................. 120 Figure 4.3-9: RADTRAD Model for [LOCA] Case 2pen ......................................................... 120 Figure 4.3-10: RADTRAD Model for [LOCA] Case 2esf ............................... 121 Figure 4.3-11: RADTRAD Model for [LOCA] Case 2pur ......................................................... 121 Figure 4.5-1: M SLB RAD TRA D M odel .................................................................................. 150 Figure 4.6-1: SGTR RADTRA D M odel ................................................................................... 163 Figure 4.7-1: CREA RADTRA D M odel ................................................................................... 180 Figure 4.8-1: LRA RADTRA D M odel ..................................................................................... 198

NOC-AE-07002127 Attachment 1 Page 11 of 219 LICENSEE'S EVALUATION

1.0 DESCRIPTION

In accordance with 10 CFR 50.67, "Accident Source Term," and 10 CFR 50.90, "Application for an amendment of license or construction permit," STP Nuclear Operating Company (STPNOC) requests an amendment to Appendix A, Technical Specifications (TS), of Facility Operating Licenses NPF-76 and NPF-80 for South Texas Project (STP) Units 1 and 2. The proposed changes are requested to support application of an alternative source term (AST) methodology, with the exception that Technical Information Document (TID) 14844, "Calculation~of Distance Factors for Power and Test Reactor Sites," (Reference 1) will continue to be used as the radiation dose basis for equipment qualification. The proposed AST methodology is consistent with the guidance in Standard Review Plan 15.0. 1, "Radiological Consequence Analyses Using Alternative Source Terms," (Reference 2) and Regulatory Guide (RG) 1. 183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,"

(Reference 3) except where alternate methods for complying with the specified portions of the NRC's regulations have been used as allowed by RG 1.183. Documentation of conformance to RG 1.183 and the allowed alternate methods are presented in Tables A through G in Attachment 6 of this submittal.

NRC Generic Letter (GL) 2003-01, 'Control Room Habitability," required licensees, in part, to confirm that the most limiting unfiltered in-leakage into the control room envelope (CRE) is no more than the value assumed in the current design basis radiological analyses. The current analyses at STPNOC assume no unfiltered in-leakage other than 10 cfmn for ingress and egress into the CRE. The American Society for Testing and Materials (ASTM) E741 tracer gas test was performed in Unit 1 in March 2004 and in Unit 2 in March 2007. The test results in both units were greater than the unfiltered in-leakage assumed in the current licensing basis accident analyses. Therefore, STPNOC opted for full-scope implementation of the AST methodology to address the test results and attain additional cost benefits described below.

In support of a full-scope implementation of the AST methodology, STPNOC, supported by Polestar Applied Technology, Inc., performed radiological consequence analyses for the following DBAs that result in control room (CR) and offsite exposure as specified in Reference 3.

" Loss of Coolant Accident (LOCA)

• Fuel Handling Accident (FHA)

" Main Steam Line Break (MSLB)

" Steam Generator Tube Rupture (SGTR)

• Control Rod Ejection Accident (CREA)

" Locked Rotor Accident (LRA)

NOC-AE-07002127 Attachment 1 Page 12 of 219 Proposed changes to the current licensing basis for the South Texas Project (STP) justified by the AST analyses include the following items.

" The use of updated. meteorological data to calculate onsite and offsite atmospheric dispersion

" Relies on less filtration

. No credit taken for Fuel Handling Building Exhaust Air Ventilation filtration No credit taken for Control Room Ventilation makeup filtration

> No credit taken for either Control Room Ventilation makeup or recirculation cleanup filtration for the Fuel Handling Accident

" Containment isolation capability is no longer required to mitigate a FHA.

• Analysis of only a single limiting FHA rather than one analysis for an FHA inside containment and a second analysis for an FHA in the fuel handling building (FHB)

" Revised control room unfiltered in-leakage assumption The proposed changes related to the applicability requirements during movement of irradiated fuel use insights from Technical Specification Task Force Traveler TSTF-5 1, Revision 2 (Reference 5). TSTF-5 1, Revision 2, was approved by the NRC on July 31, 2003. TSTF-51 changed the TS operability requirements for certain engineered safety features such that they are not required to be operable after sufficient radioactive decay has occurred to ensure that offsite doses remain within limits. STPNOC's change differs from TSTF-51 in that the definition of "recently irradiated fuel" is not used. Instead, the STPNOC change is based upon a specification in the STP Technical Requirements Manual that precludes the movement of irradiated fuel until a "42 hour4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> decay time" has occurred following the achievement of subcriticality. This is consistent with the AST accident analysis assumption. Defining "recently irradiated fuel" has no practical application for the STPNOC proposed change. The "decay time" definition is consistent with the accident analyses assumption so that STPNOC would not be in a condition where recently irradiated fuel could be moved. The accident analyses were performed with a decay time such that the impacted engineered safety features are not required at any time that fuel is being moved.

Approval of this change will provide a source term for STP that will result in a more accurate assessment of the DBA radiological doses. The improved dose assessment allows relaxation of some current licensing basis requirements as described in Section 2.0. Upon implementation of AST, containment closure capability will no longer be required to mitigate an FHA. This proposed change provides flexibility when performing refueling activities by allowing movement of equipment through the containment boundary in support of outage activities while meeting accident radiological acceptance criteria. Outages can be optimized to achieve an overall risk reduction while also reducing outage time and cost. Outage resources can be directed to other activities, which ultimately should result in improvements in plant maintenance, operations and overall safety.

The revised radiological dose to the control room operator allows for a revised air unfiltered in-leakage assumption that provides a conservative margin over that determined by air in-leakage testing. In the Spring of 2004, STPNOC tested the Unit 1 control room envelope for unfiltered in-leakage using the tracer gas method. The highest measured unfiltered in-leakage was 9.4 cfm compared to the current licensing basis assumption of 0 cfm. Although the resultant dose

NOC-AE-07002127 Attachment 1 Page 13 of 219 increase was more than minimal as defined by 10 CFR 50.59 regulatory guidance, the result was within the regulatory limits. The assumption of 100 cfm unfiltered in-leakage used in the revised analyses was based on the Unit 1 results. The Unit 2 Control Room Envelope was recently tested for unfiltered in-leakage using the tracer gas method. The highest unfiltered in-leakage was 64 cfm. Regulatory limits are met in Unit 2 with compensatory measures. STPNOC is currently operating under a non-conforming licensing basis condition. The conditions in Units 1 and 2 are documented in the corrective action program. A revision to the source term and the assumed unfiltered in-leakage will assure that the revised dose analyses proposed by this amendment request are met.

Westinghouse Electric Company Nuclear Safety Advisory Letter NSAL-06-15, dated December 13, 2006, advised operators of Westinghouse plants that the single failure scenario for the steam generator tube rupture (SGTR) analysis may not be limiting. The methodology included evaluations of various single failures for a reference plant. Recent industry operating experience identified a condition where a failed-open main steam line isolation valve (MSIV) on the steamline from the ruptured steam generator (SG) may result in a steam flow that is higher than that previously assumed in the accident analysis and thus higher offsite doses rates.

The STP current SGTR analysis and the SGTR analysis presented in the safety evaluation for this licensing amendment request assumes a failed open SG power operated relief valve as the limiting single failure as far as assumed total steam release. An evaluation of NSAL-06-15 has resulted in a revised conclusion that the failed open MSIV results in the greater steam release at STP. This is because the steam valves to the moisture-separator reheater fail open on a loss of instrument air resulting from a loss of offsite power. The steam valves to the moisture-separator reheater fail closed for the reference plant thus significantly reducing the steam release from a failed open MSIV.

STP is currently operating under an administrative limit for reactor coolant system dose equivalent iodine that is lower than the Technical Specification limit. STP plans to correct this non conforming condition at the earliest opportunity. The most likely path to resolution will be a plant modification. Therefore, the assumptions, including the limiting single failure regarding total steam release, for the SGTR accident analysis performed for this amendment request will be consistent with the plant response to this event after the modification is completed. Until the modification is completed, STP will continue to maintain an administrative limit for reactor coolant system dose equivalent iodine so that the radiological dose limits for the SGTR analysis remain bounding. A commitment to maintain these controls until the plant is modified is described in Attachment 5. This condition is documented under Condition Report 07-2887.

STPNOC requests approval of the proposed amendment by March 30, 2008. Once approved, the amendment shall be implemented within 120 days due to the significant implementation scope of the subject changes. This implementation period will provide adequate time for the affected station documents to be revised using the appropriate change control mechanisms.

NOC-AE-07002127 Attachment 1 Page 14 of 219

2.0 PROPOSED CHANGE

The proposed changes related to the applicability requirements during movement.of irradiated fuel assemblies use insights from Technical Specification Task Force Traveler (TSTF)-5 1, "Revise Containment Requirements During Handling of Irradiated Fuel and Core Alterations,"

Revision 2. The NRC approved TSTF-51 on July 31, 2003. TSTF-51 changes the TS operability requirements for engineered safety features such that they are not required to be operable after sufficient radioactive decay has occurred to ensure that offsite doses remain within limits.

STPNOC's change differs from TSTF-51 in that the definition of "recently irradiated fuel" is not used. Instead, the STPNOC change is based upon a specification in the STP Technical Requirements Manual (TRM) that precludes the movement of irradiated fuel until a "42 hour4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> decay time" has occurred following the achievement of sub-criticality. This is consistent with the AST accident analysis assumption. The STP TS changes are based upon the TRM decay time requirement. The decay time specification was relocated from TS to the TRM with approval of STPNOC licensing amendments 145 and 133 to Unit 1 and Unit 2 respectively (Reference 7).

In the safety evaluation for this TS change, the NRC staff found that the "decay time requirement" specification does not need to be in the TS because it is not needed to ensure the decay time limit is met. This is because certain operational steps, such as containment entry, pressure vessel head removal, and cavity flood-up must be completed before fuel movement in the vessel is possible following critical operation. These preliminary activities require more than 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> to complete. The NRC staff found that these physical limitations are adequate to ensure compliance with the 42-hour limit (relocated to the TRM). Thus including the decay time limit in TSs is not needed to ensure this limit is met. Using insight from TSTF-51, the proposed change also deletes CORE ALTERATIONS from applicability requirements for some Limiting Conditions for Operation.

Using insights from TSTF-5 1, STPNOC is committing to the applicable provisions of Nuclear Utilities Management and Resources Council (NUMARC) 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Revision 3, (Reference

6) as described in TSTF-51 and documented in Attachment 5 to this submittal. Additional discussion regarding these provisions is described in Section 4.4.8 of this evaluation.

Proposed changes to the TS resulting from this submittal are summarized below.

Section 1.0, Definitions The dose conversion factors used to calculate the dose from DOSE EQUIVALENT 1-131 concentration are revised to those listed in Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion," 1988; (Table 2.1, Exposure-to-Dose Conversion Factors for Inhalation) (Reference 20) instead of TableE-7 of NRC Regulatory Guide 1.109, Revision 1, October 1977 (Reference 32). "Thyroid" is dropped from the definition of dose to reflect that the dose is now the total effective dose equivalent based on Alterative Source Term methodology.

NOC-AE-07002127 Attachment 1 Page 15 of 219 Table 3.3-3, Engineered Safety Features Actuation System Instrumentation Modes 5 and 6 for Functional Unit 3.b.4), "Containment Ventilation Isolation RCB Purge Radioactivity - High," are deleted as APPLICABLE MODES because automatic isolation is no longer required during core alterations or movement of irradiated fuel within containment to meet the AST design basis accident analysis.

ACTION 18 is modified appropriately.

Modes 5 and 6 for Functional Unit 10, "Control Room Ventilation," are deleted as APPLICABLE MODES because the accident mitigation capabilities of this system are no longer credited in AST design basis accident analysis for activities performed during these MODES.

The ACTION 28 requirement for Functional Unit 1O.d, "Control Room Intake Air Radioactivity - High," is modified to delete suspension of core alterations, movement of irradiated fuel, and crane operation with loads over the spent fuel pool because the.

accident mitigation capabilities of this system are no longer credited in AST design basis accident analysis during these activities.

The requirement for an operable Functional Unit 11, Fuel Handling Building (FHB)

Heating, Ventilation and Air Conditioning (HVAC), actuation instrumentation is deleted since the accident mitigation capabilities of the FHB HVAC system are no longer credited in AST design basis accident analysis.

• An administrative change is made to remove a Note from ACTION 20 because the provisions of the Note have expired.

Table 3.3-4, Engineered Safety Features Actuation System Instrumentation Trip Setpoints The trip setpoints and allowable values for Functional Unit 11 .a., "FHB HVAC," are deleted since the accident mitigation capabilities of this system are no longer credited in AST design basis accident analyses.

Table 4.3-2, Engineered Safety Features Actuation System Instrumentation Surveillance Requirements

• Modes 5 and 6 for Functional Unit 3.b.4), "Containment Ventilation Isolation RCB Purge Radioactivity - High," are deleted as APPLICABLE MODES because automatic isolation is no longer required during core alterations or movement of irradiated fuel within containment to meet the AST design basis accident analysis.

• Modes 5 and 6 for Functional Unit 10, "Control Room Ventilation," are deleted as APPLICABLE MODES because the accident mitigation capabilities of this system are no longer credited in AST design basis accident analysis for activities performed during these MODES.

NOC-AE-07002127 Attachment 1 Page 16 of 219 The requirement for performing surveillances for Functional Unit 11, Fuel Handling Building (FHB) Heating, Ventilation and Air Conditioning (HVAC), actuation instrumentation is deleted because the accident mitigation capabilities of the FHB HVAC system are no longer credited in AST design basis accident analysis.

TS 3/4.7.7, Control Room Makeup and Cleanup Filtration System

" The APPLICABILITY for Modes 5 and 6 is deleted. Requirements to suspend all operations during core alterations, movement of irradiated fuel, and crane operation with loads over the spent fuel pool are deleted because the accident mitigation capabilities of this system are no longer credited in AST design basis accident analysis during these activities. Requirements to suspend operations involving positive reactivity additions that could result in loss of required SHUTDOWN MARGIN or required boron concentration are deleted. Adequate SHUTDOWN MARGIN is controlled by TS 3/4.1.1, "Shutdown Margin Boration Control" and adequate boron concentration is controlled by TS 3/4.9.1, "Boration Concentration."

The safety analysis concludes that administrative controls and operator response time are adequate measures to preclude a loss of required SHUTDOWN MARGIN or required boron concentration. In addition, requirements to suspend operations involving positive reactivity additions are not found in Standard Technical Specifications for control room ventilation systems. Thus there are no radiological consequences.

" ACTION C for MODES 1, 2, 3,and 4 is modified to delete the requirements to suspend all operations involving movement of spent fuel and crane operations with loads over the spent fuel pool because the accident mitigation capabilities of this system are no longer credited in AST design basis accident analysis.

• An administrative change is made to remove a footnote from surveillance requirement 4.7.7.e 3) because the provisions of the footnote have expired.

TS 3/4.7.8, Fuel Handling Building (FHB) Exhaust Air System This specification is deleted because the accident mitigation capabilities of the FHB Exhaust Air HVAC system are no longer credited in AST design basis accident analysis.

TS 3.8.1.2, A.C. Sources Shutdown The actions to suspend movement of irradiated fuel or crane operation with loads over the spent fuel pool if the Limiting Condition for Operation is not met are deleted because the fuel handling accident analysis no longer credits the mitigation systems that are dependent upon this source of electrical power.

NOC-AE-07002127 Attachment 1 Page 17 of 219 TS 3.8.1.3, A.C. Sources Shutdown The actions to suspend movement of irradiated fuel or crane operation with loads over the spent fuel pool if the Limiting Condition for Operation is not met are deleted because the fuel handling accident analysis no longer credits the mitigation systems that are dependent upon these sources of electrical power.

TS 3.8.2.2, D.C. Sources Shutdown The action to suspend movement of irradiated fuel if the Limiting Condition for Operation is not met are deleted because the fuel handling accident analysis no longer credits the mitigation systems that are dependent upon these sources of electrical power.

TS 3.8.2.3, Onsite Power Distribution Shutdown The action to suspend movement of irradiated fuel if the Limiting Condition for Operation is not met are deleted because the fuel handling accident analysis no longer

• 'credits the mitigation systems that are dependent upon these sources of electrical power.

TS 3/4.9.4, Containment Building Penetrations during Refueling Operations This specification is deleted because containment isolation is no longer credited in AST design basis accident analysis during core alterations or movement of irradiated fuel.

TS 3/4.9.9, Containment Ventilation Isolation during Refueling Operations This specification is deleted because containment isolation is no longer credited in AST design basis accident analysis during core alterations or movement of irradiated fuel.

TS 3/4.9.12, Fuel Handling Building Exhaust Air System during Refueling Operations This specification is deleted because the accident mitigation capabilities of the FHB HVAC system are no longer credited in AST design basis accident analysis.

In summary, this proposal (1) removes the Fuel Handling Building Exhaust Air System LCO for all MODES, (2) removes the Control Room Makeup and Cleanup Filtration System LCO during MODES 5 and 6, and (3) removes the Containment Building Penetrations LCO including the Containment Ventilation Isolation System LCO during Modes 5 and 6 including Refueling Operations from the Technical Specifications. The actions to suspend movement of irradiated fuel or crane operation with loads over the spent fuel pool if the Limiting Condition for Operation is not met are deleted from the Electrical Sources Technical Specifications while Shutdown because the fuel handling accident analysis no longer credits the mitigation systems that are dependent upon this source of electrical power. Finally, the definition for DOSE EQUIVALENT 1-131 is revised.

NOC-AE-07002127 Attachment 1 Page 18 of 219 The proposed changes do not impact Technical Specification requirements for systems needed to prevent or mitigate CORE ALTERATION events other than the FHA. The proposed changes also do not change the requirements of systems needed to mitigate potential vessel drain down events, systems needed for decay heat removal, or the requirements to maintain high water levels over irradiated fuel.

Corresponding changes to the TS Bases will be made following approval of the proposed amendment in accordance with the TS Bases Control Program and 10 CFR 50.59. The planned changes to the affected TS Bases pages are provided in Attachment 3 for information.

STPNOC plans to delete TS 3/4.7.8, 3/4.9.4 and 3/4.9.12 and add requirements to the Technical Requirements Manual (TRM) to facilitate restoration of Containment closure or the Fuel Handling Building Exhaust Air System, as applicable, and to provide a filtered and monitored release path as a defense-in-depth measure to mitigate the consequences of a postulated FHA.

Further, STPNOC plans to insert a new TRM requirement to facilitate the restoration of one train of Control Room Makeup and Cleanup Filtration System as a defense-in-depth measure to mitigate the consequences of a postulated FHA. Attachment 4 provides the planned TRM pages for information. See Section 4.4.8 for further discussion.

The planned TRM specification for the FHB Exhaust Air System only requires one train to be OPERABLE or capable of being restored to an OPERABLE status to meet the Limiting Condition for Operation (LCO) whenever irradiated fuel is in the spent fuel pool. Movement of fuel within the spent fuel pool or crane operation with loads over the spent fuel pool will be required to be suspended if at least one FHB Exhaust Air Train can not be restored to OPERABLE status within the time required by the LCO. Surveillance Requirements will remain the same as TSs with the exception that the surveillance to verify that the system automatically starts upon initiation of a high radiation or safety injection test signal will not be included as a surveillance. This change does not propose to downgrade the safety classification of this system.

Amendments 125/113 (Reference 8) to Units 1 and 2, respectively, provide an allowed outage time of up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to restore at least one train of control room makeup and cleanup filtration system or one train of fuel handling building exhaust air filtration to an operable status when multiple trains of either system are inoperable in MODES 1, 2, 3 or 4. As a compensatory measure to ensure that applicable regulatory limits continue to be met, STPNOC committed to not intentionally enter the action for multiple trains of the Control Room Makeup and Cleanup Filtration System and the Fuel Handling Building Exhaust Air System simultaneously .in MODES 1, 2, 3 or 4. Although this proposed amendment request will relocate TS 3.7.8 to the TRM, the compensatory measure described for Amendments 125/113 will remain in place with procedural requirements revised as appropriate. TS Bases page B 3/4 7-5 of Attachment 3 reflects this change.

The planned TRM specification for Containment Building Penetrations requires containment building penetrations to be closed or capable of being closed within two hours following a fuel handling accident. Surveillance Requirements will remain the same as those previously in the TS -withthe exception that the surveillance to verify that the containment purge and exhaust

NOC-AE-07002127 Attachment 1 Page 19 of 219 isolation valves automatically close upon initiation of an isolation test signal will not be included.

The planned TRM specification for the Control Room Makeup and Cleanup Filtration System requires only one system to be OPERABLE or capable of being restored to an OPERABLE status to meet the Limiting Condition for Operation (LCO) whenever irradiated fuel is in the spent fuel pool or during the movement of irradiated fuel, which includes refueling operations in the reactor containment building. Movement of irradiated fuel, or crane operation with loads over the spent fuel pool will be required to be suspended if at least one Control Room Makeup and Cleanup Filtration train can not be restored within the time required by the LCO.

Surveillance Requirements will remain the same as TSs with the exception that the surveillance to verify that the system automatically starts upon initiation of a high radiation or safety injection test signal will not be included as a surveillance. This amendment change request does not propose to downgrade the safety classification of this system.

The two hours to restore the FHB Exhaust Air System and the Control Room Makeup and Cleanup Filtration System to OPERABLE status and to close containment closures in the event of a fuel handling accident is reasonable because these systems are not required to mitigate the accident. These systems are not credited in the accident analyses. Dose limits are within

• requirements assuming an instantaneous release from the FHA. These additional administrative actions are taken to further filter and monitor the release as a defense-in-depth measure.

3.0 BACKGROUND

3.1 Systems Affected by the Proposed Change The following systems are affected by this proposed amendment:

1. The Containment Ventilation Isolation System closes the containment isolation valves in the Normal Containment Purge System and the Supplementary Containment Purge System. This action isolates the containment atmosphere from the environment to minimize releases of radioactivity in the event of an accident. These systems are described in Sections 9.4.5.2.6 and 9.4.5.2.7 of the Updated Final Safety Analyses Report (UFSAR).

2. The Control Room Makeup and Cleanup Filtration System provides a protected environment from which the operators can control the unit following an uncontrolled release of radioactivity by maintaining the control room envelope at a positive pressure.

Outside air is filtered and mixed with the air being re-circulated within the control room.

Pressurization of the control room minimizes the infiltration of unfiltered air from the surrounding areas of the building. The Control Room Makeup and Cleanup Filtration System satisfies the design requirement of limiting dose to the control room operators following the design basis accident in accordance with General Design Criterion 19 of 10 CFR 50, Appendix A. This system consists of three 50-percent-capacity redundant

NOC-AE-07002127 Attachment 1 Page 20 of 219 trains. The system isolates normal supply ventilation and initiates filtered makeup and cleanup ventilation of the control room envelope following receipt of an accident initiation signal. Each train of filtered makeup and cleanup ventilation consists of a prefilter, a high efficiency particulate air (HEPA) filter, an activated charcoal adsorber section for removal of gaseous activity (principally iodines), and a fan. Ductwork, valves or dampers, and instrumentation also form part of the system. A second bank of HEPA filters, not credited in the accident analysis, follows the adsorber section to collect carbon fines and provides backup in case of failure of the main HEPA filter bank. This system is described in Sections 6.4, 6.5.1 and 9.4.1.1.1 of the UFSAR.

3. The Fuel Handling Building Exhaust Air System filters airborne radioactive particulates from the area of the fuel pool following a fuel handling accident or loss of coolant accident. This system consists of two independent and redundant filter trains.

Each train consists of a heater, a prefilter, a high efficiency particulate air (HEPA) filter, and an activated charcoal adsorber section for removal of gaseous activity (principally iodines). Three 50-percent-capacity main and booster fans serve the redundant exhaust trains. Heaters, ductwork, valves or dampers, and instrumentation also form part of the system functioning to reduce the relative humidity of the airstream. A second bank of HEPA filters, not credited in the accident analysisjfollows the adsorber section to collect carbon fines and provide backup in case the main HEPA filter bank fails. The system establishes a negative pressure in the Fuel Handling Building and initiates filtered exhaust ventilation from the building following receipt of a high radiation signal or a safety injection signal. This system is described in Sections 6.5.1 and 9.4.2 of the UFSAR.

4. The Engineered Safety Features (ESF) AC and DC Electrical Power Systems are designed with redundancy and independence of onsite power sources, distribution systems, and controls in order to provide a reliable supply of electrical power to the ESF electrical loads necessary to achieve safe plant shutdown, or to mitigate the consequences of postulated accidents. The ESF AC and DC Electrical Power Systems are described in Section 8.3 of the UFSAR.

5. The Reactor Containment Building serves to contain fission product radioactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained well within the requirements of 10 CFR 100.11. Additionally, the Reactor Containment Building provides radiation shielding from the fission products that may be present in the containment atmosphere following accident conditions.

6. The Fuel Handling Building is designed as a controlled-leakage structure. The Fuel Handling Building in conjunction with the Fuel Handling Building Exhaust Air System creates an enclosure to direct radioactivity releases to the environment through a filter

- bank such that offsite radiation exposures are maintained well within the requirements of 10 CFR 100.11.

No changes to these systems are required to implement the proposed change. The change will allow the automatic start feature of systems no longer credited in the accident analyses for

NOC-AE-07002127 Attachment 1 Page 21 of 219 mitigation to be disabled through the STPNOC modification process. The modification process provides checks to assure that the modification does not invalidate assumptions made in the PRA or adversely impact the severe accident management program.

3.2 The AST Rule On December 23, 1999, the NRC issued the Final Rule on "Use of Alternate Source Terms at Operating Reactors." The Final Rule, issued under 10 CFR 50.67, "Accident source term,"

allows holders of operating licenses issued prior to January 10, 1997, to voluntarily replace the traditional source term used in design basis accident analyses with alternative source terms. This action allows interested licensees to pursue cost beneficial licensing actions to reduce unnecessary regulatory burden without compromising the margin of safety of the facility.

The fission product release from the reactor core into containment is referred to as the "source term," and is characterized by the composition and magnitude of the radioactive material, the chemical and physical properties of the material, and the timing of the release from the reactor core. Since the publication of U.S. Atomic Energy Commission Technical Information Document, TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites,"

which is the currently used design basis document for calculation of offsite dose for loss of coolant accidents, significant advances have been made in understanding the composition and magnitude, chemical form, and timing of fission product releases from severe nuclear power plant accidents. Many of these insights developed out of the major research efforts started by the NRC and the nuclear industry after the accident at Three Mile Island. NUREG-1465 (Reference

9) was published in 1995 with revised ASTs for use in the licensing of future Light Water Reactors (LWRs). The NRC, in 10 CFR 50.67, later allowed the use of the ASTs described in NUREG-1465 at operating'plants. This NUREG represents the result of decades of research on fission product release and transport in Light-Water Reactors under accident conditions. One of the major insights summarized in NUREG-1465 involves the timing and duration of fission product releases.

The five release phases describing the progression of a severe accident in a LWR are listed in NUREG-1465 and are given below.

1. Coolant Activity Release

2. Gap Activity Release

3. Early In-Vessel Release

4. Ex-Vessel Release

5. Late In-Vessel Release Phases 1, 2, and 3 are considered in current (i.e., pre-AST) DBA evaluations; however, they are all assumed to occur instantaneously. Phases 4 and 5 are related to severe accident. evaluations.

Under the AST methodology, only the coolant activity release (i.e., Phase 1) is assumed to occur instantaneously and end with the onset of the gap activity release (i.e., Phase 2). This approach represents a more realistic time sequence for activity release. The insights from NUREG-1465 were subsequently incorporated into RG 1.183.

NOC-AE-07002127 Attachment 1 Page 22 of 219 3.3 STP Full Application of AST In order to utilize this more realistic approach, this license amendment request proposes to implement a full-scope application of the AST methodology addressing the composition and magnitude of the radioactive material, its chemical and physical form, and the timing of its release. STPNOC has performed radiological consequence analyses of the below DBAs that result in the most significant offsite exposures. The AST analyses have been performed in accordance with the guidance in RG 1.183 and NRC Standard Review Plan 15.0.1.

• Loss of Coolant Accident (LOCA)

• Fuel Handling Accidents (FHA) in the Fuel Handling Building (FHB) and in Containment

" Main Steam Line Break (MSLB)

• Steam Generator Tube Rupture (SGTR)

• Control Rod Ejection Accident (CREA)

" Locked Rotor Accident (LRA)

Implementation of an AST methodology changes only the regulatory assumptions regarding the analytical treatment of the DBAs. Implementation of the AST methodology will provide a source term for STP that will result in a more accurate assessment of the DBA radiological doses.

3.4 Precedent The NRC has previously approved implementation of the AST methodology at a number of nuclear power plants. The STPNOC analyses conform to NRC RG 1.183 as demonstrated in Attachment 6.

An aspect of the STPNOC analyses where specific precedence exists is the methodology and code for calculation of transient containment sump pH. Application of this computer code is found in the Vermont Yankee AST application (Reference 10). See Section 4.3.3.1.2 for further discussion.

The relocation of the Fuel Handling Building Air Exhaust System and the associated actuation instrumentation requirements and the relocation of the Containment Ventilation Isolation System and the associated actuation instrumentation requirements out of Technical Specifications based on using the alternate source term methodology were previously approved in the Surry Plant Amendment 230. (Reference 11) The deletion of the spent fuel pool ventilation filtration system from the Technical Specifications based on alternate source term methodology was previously approved in the Salem Plant Amendments 263 and 245. (Reference 12)

NOC-AE-07002127 Attachment 1 Page 23 of 219

4.0 TECHNICAL ANALYSIS

The AST analyses for STP were performed following the guidance in Standard Review Plan 15.0.1 and Regulatory Guide 1.183 except where alternate methods for complying with the specified portions of the NRC's regulations were used as allowed by Regulatory Guide 1.183.

Documentation of conformance to Regulatory Guide 1.183 and the allowed alternate methods are presented in Attachment 6 of this submittal.

The full-scope implementation consists of the following:

1. Identification of the core source term based on plant specific analysis of core fission product inventory.

2. Determination of the release fractions for the six Pressurized Water Reactor (PWR)

Design Basis Accidents (DBAs) identified in Appendices A, B, E, F, G, and H of Regulatory Guide 1.183 that could potentially result in control room and offsite doses.

These are the Loss Of Coolant Accident (LOCA), the Fuel Handling Accident (FHA), the Main Steam Line Break accident (MSLB), the Steam Generator Tube Rupture accident (SGTR), the Control Rod Ejection Accident (CREA), and the Locked Rotor Accident (LRA).

3. Calculation of new control room (CR), exclusion area boundary(EAB), and low population zone (LPZ) atmospheric dispersion factors (X/Q) for the containment leakage, plant vent, and steam generator (SG) secondary side power-operated relief valve (PORV) release paths.

4. Calculation of offsite and control room personnel Total Effective Dose Equivalent (TEDE).

5. Evaluation of containment sump pH to ensure that the particulate iodine deposited into the containment water during the DBA LOCA does not re-evolve beyond the amount recognized in the DBA LOCA analysis.

6. Evaluation of other related design and licensing bases such as NUREG-0737 (Reference 13).

Implementation of AST includes changes to the methodology presently used at STP for dose consequence analysis. These include:

1. Use of updated meteorological data (five years from 2000 to 2004) to calculate onsite and offsite atmospheric dispersion.

2. No credit for any filtration other than for the recirculation filters for the control room (CR) and the recirculation filters for the Technical Support Center (TSC). No filtration whatsoever has been relied upon in the analysis of the Fuel Handling Accident (FHA).

NOC-AE-07002127 Attachment 1 Page 24 of 219

3. Limits on DBA LOCA iodine removal from the containment atmosphere based on the containment sump pH going slightly below 7.0 after one day. (The lowest pH reached over 30 days post-accident is 6.8.)

4. Analysis of only a single limiting FHA rather than one analysis for an FHA inside containment and a second analysis for an FHA in the FHB.

The revised dose consequence analyses were prepared, reviewed, and are 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. The analyses were performed in a manner to ensure that dose analyses have not been "tuned" to a specific set of accident progression assumptions. Conservative parameters are used when calculating components in the dose analyses. For example, the determination of iodine re-evolution in the LOCA analysis assumes the maximum sump temperature, which occurs in the first minutes of the accident, in conjunction with the lowest pH, which occurs at the very end of the 30-day dose assessment period, to achieve the limiting results.

The accident analyses assumptions are not based upon risk insights.

The Technical Analysis will demonstrate that all post-accident radiological doses for DBAs required by Regulatory Guide 1.183 remain within regulatory limits.

Overview and Organization of Accident Descriptions This section provides the information on the analyses performed in support of the change of the design basis source terms from the current TID-14844 source terms to the NUREG-1465 Alternate Source Term. The section begins with a presentation of the updated meteorological data that was used to determine revised atmospheric dispersion factors for both offsite locations and the control room (Section 4.1). These revised y/Q values are used in all new analyses.

Presentations of the generic analytical models used in the revised analyses, are then presented (Section 4.2). This includes the formulation for the offsite dose model and detailed HVAC models for both the control room and Technical Support Center (TSC). Development of the radiological source terms that are used as a basis for the revised analyses is discussed. This discussion includes physical nuclide parameters and dose conversion factors. The core nuclide inventory is presented, along with isotopic concentrations for the reactor coolant system at the Technical Specification normal maximum iodine concentration of 1 pCi/gm and at the Technical Specification iodine spike limit of 60 pCi/gm. Similarly, secondary system nuclide concentrations are developed for 1% failed fuel and the Technical Specification normal maximum iodine concentration of 0.1 ptCi/gm. These discussions are followed by detailed descriptions of the following accidents:

Section 4.3 LOCA Section 4.4 Fuel Handling Accident

NOC-AE-07002127 Attachment 1 Page 25 of 219 Section 4.5 Main Steam Line Break Section 4.6 Steam Generator Tube Rupture Section 4.7 Control Rod Ejection Section 4.8 Locked Rotor The description of each accident follows the following format:

• An overview of the methodology of the analysis

• The analytical model(s) used to perform the analysis

• Development/discussion of the radiological source term

• Discussion of the radiological releases (usually steam flows)

• Assumptions and inputs

• A table of important parameters specific to that analysis

• Summary, including the dose results The contents of each subsection may change depending on the needs for the specific accident under discussion; A list of commonly used acronyms is presented in Table 4.0-1.

NOC-AE-07002127 Attachment 1 Page 26 of 219 Table 4.0-1 Frequently Used Acronyms AHU Air handling unit AST The Alternative Source Term, as defined in NUREG-1465 and Regulatory Guide 1.183 CLB The current licensing basis, including analyses, for the South Texas Project CR Control Room CRE Control Room (HVAC) Envelope CREA Control Rod Ejection Accident CSS Containment Spray System DCF Dose Conversion Factor DEI Dose Equivalent Iodine- 131 EAB Exclusion Area Boundary (site boundary)

ESF Engineered Safety Feature FHA Fuel Handling Accident FHB Fuel Handling Building IVC Isolation Valve Cubicle (location of PORVs and MSIVs)

LPZ Low Population Zone LOCA Loss of Coolant Accident LOOP Loss of Offsite Power LRA Locked Rotor Accident MSIV Main Steam Isolation Valve MSLB Main Steam Line Break PORV Power Operated Relief Valve RCB Reactor Containment Building RCP Reactor Coolant Pump RCS Reactor Coolant System SG Steam Generator SGTR Steam Generator Tube Rupture TSC Technical Support Center TSP Trisodium phosphate I)

NOC-AE-07002127 Attachment 1 Page 27 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION 4.1 Meteorology and Atmospheric Dispersion The revised X/Q values used for the AST application have been developed using more recent meteorological data than that used for the current licensing basis (CLB). These more recent data were obtained for the years 2000 to 2004 (five years worth of data). Polestar subcontracted ABS Consulting to perform the meteorological data analysis and the PAVAN and ARCON96 analyses.

The x/Q values resulting at the Control Room intake are calculated using the NRC-sponsored computer code ARCON96 consistent with the procedures in Regulatory Guide 1.194, "Atmospheric Relative Concentrations for Control.Room Radiological Habitability Assessments at Nuclear Power Plants," (Reference 14).

The X/Q values resulting at the EAB and LPZ are calculated using the NRC-sponsored computer code PAVAN (Reference 15), consistent with the procedures in Regulatory Guide 1.145 (Reference 16).

Onsite Meteorological Monitoring Program The.meteorological measurement program at STP consists of a 60-meter primary tower and a free standing 10-meter tower which serves as a backup to the primary system. The two methods used to determine atmospheric stability are:

a. Delta Temperature (i.e., vertical temperature difference), which is the principal method, and

b. Sigma theta (i.e., standard deviation of the horizontal wind direction), as a backup method.

Data, gathered per Safety Guide 23, "Onsite Meteorological Programs" (Reference 17), are used to determine the meteorological conditions specific to the plant site. The meteorological program includes information on site specific instrumentation and calibration procedures.

The meteorological tower is equipped with instrumentation that conforms to the system accuracy recommendations in Safety Guide 23. The dew point instrument is less accurate than the Regulatory Guide requires. The meteorological instrumentation is placed on horizontal booms oriented into the generally prevailing wind direction at the site. Equipment signals are transmitted to an instrument building with controlled environmental conditions. The instrument building, at or near the base of the tower, houses the recording equipment, processor cards, and digital recording equipment, etc. This instrumentation is used to process and retransmit the data to the end-point users.

NOC-AE-07002127 Attachment 1 Page 28 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION Recorded meteorological data are used to create joint frequency tables by quarter and to provide input to dispersion estimates of airborne concentrations of gaseous effluents in support of offsite radiation dose assessments. Instrument calibrations and data consistency evaluations are performed according to Safety Guide 23 to ensure maximum data accuracy. Capability is maintained to evaluate atmospheric dispersion better than 90% of the time even though individual instruments may not record data with 90% reliability. Delta-temperature is the primary method used for determining atmospheric stability. When the delta-temperature instrument is inoperable, wind direction sigma theta data are used to maintain 90% availability.

Likewise, when wind direction or speed is unavailable from the primary tower, data from the corresponding instruments on the backup tower are substituted.

Site Description The minimum EAB and LPZ boundaries are located at 1430 m and 4800 m. Note that Plant North is also True North. A simplified diagram of the units and the release points and receptors is provided in Figure 4.1-13.

A description of the climate at STP is provided in Section 2.3.1.1 of the UFSAR (Revision 14):

"... The climate of the region is subtropical maritime and is characterized by short mild winters and long hot summers. In the vicinity of the site, the humidity is generally high and rainfall is abundant throughout the year.

"Summer type climate extends from about May through September, with the highest temperature occurring during July and August. The summer weather is normally dominated by tropical maritime air masses associated with the Bermuda High. Days are typically hot and humid, and convective showers and thunderstorms are relatively frequent.

"Winter type climate extends from December through February, with the coldest weather occurring in January. The Gulf of Mexico modifies outbreaks of polar air masses to such an extent that temperatures below 327F occur on an average of less than four times per year

([UFSAR] Ref. 2.3-1).

"The fall type climate months are October and November, and the spring type climate months are March and April. Both transitional seasons are short and are characterized by mild, pleasant weather. The locations and a brief topographical description of the meteorological stations used to determine the general climate and local meteorology ([UFSAR] Section 2.3.2) in the STPEGS site region are presented in [UFSAR] Table 2.3-1.

"The STPEGS site is located approximately 8 miles north-northwest of Matagorda, Texas, just west of the Colorado River. The site elevation is approximately 25 ft. above mean sea level (MSL). More detail on local topography and its influences on local meteorology is presented in Section [UFSAR] 2.3.2."

NOC-AE-07002127 Attachment 1 Page 29 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION A description of the topographic influences on its meteorology and diffusion estimates is provided in UFSAR Section 2.3.2.2.2 (Revision 14):

"The terrain in the region of the STPEGS site is generally flat. Elevations above MSL average approximately 25 ft. [UFSAR] Figure 2.3-10 is a topographic map of the site area within a 5-mile radius. [UFSAR] Figure 2.3-11 is a topographic map of the site area within a 50-mile radius.

"The major local effect on site meteorology is the presence of the Gulf of Mexico and the resultant sea and land breeze circulations. The sea breeze generally forms during the spring and summer when the Gulf of Mexico's water temperature is colder than the air temperature and results in an onshore wind-flow. During periods of light geostrophic winds, surface winds may develop which blow onshore (sea to land) during the day and offshore (land to sea) at night. The formation of the sea breeze is the result of the temperature variation between water and land.

Turbulent mixing within the water effects a continuous downward transport of surface heat through large masses of water during spring and summer, thus lowering the surface water temperature (and also lowering the -temperature of the overlying air), in contrast with the strong surface heating of the air over the shoreline region. This contrast is also intensified because the water has a higher thermal capacity than that of the soil. As a result of this situation, temperatures over land are greater than those over water during spring and summer; this difference diminishes toward sunset and may reverse during the night. As the warmed air over the land begins to rise, a horizontal density gradient is formed which causes the heavier, colder air over the water to flow underneath the warm air during these seasons. To ensure continuity of the circulation cell, there is a return motion of the warmer air from land to the Gulf at higher levels. Although formation of the sea breeze circulation is usually perpendicular to the shoreline, Coriolis forces become significant as the system matures. During the late afternoon, the sea breeze can be expected to have a major component parallel to the shore to the right of the onground trajectory. Land breezes are the converse of sea breezes and may develop when sea temperatures are warmer than the land, such as during the fall and early winter or during the night in the summer. However, land breezes are generally weaker and less frequent than sea breezes ([UFSAR] Ref. 2.3-35).

"Therefore, considering the basis and characteristics of sea breeze circulations, these local wind systems are a definite factor relevant to the STPEGS site. The sea breeze circulation in southeast Texas extends approximately 25 mile inland ([UFSAR] Ref. 2.3-1). A study was undertaken to determine instances of sea breeze penetration to the STPEGS site. Synoptic weather conditions and hourly station data for selected stations in the STPEGS site area, obtained from the NWS for the period July 21, 1973 to July 20,1977, along with historical data on sea surface temperatures of the Gulf of Mexico in the site regions, were used to identify periods of potential sea breeze penetration into the STPEGS site. The site data were then examined for these periods to confirm sea breeze occurrences. Temperature drops and relative humidity changes were noted for three selected NWS stations in the vicinity of the STPEGS site as well as for the site, and hydrographs were plotted and analyzed for the NWS stations to confirm the occurrence of sea breezes in the site area. Based on these analyses, 51 days (primarily in the spring and summer) were identified

NOC-AE-07002127 Attachment 1 Page 30 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION during which the sea breeze penetration might reach the STPEGS site area. Of these, there were 35 days (10 percent on an annual basis) on which sea breeze occurrence was confirmed at the site. There were four cases of sea breeze at the STPEGS site on 2 consecutive days and one case of 4 consecutive days.

"Considering both the low frequency of consecutive daily occurrences of sea breeze penetration to the site and the low overall frequency of occurrence of sea breeze penetration to the site, the impact of sea breeze upon dose estimates for the STPEGS is expected to be insignificant

([UFSAR] Ref. 2.3-53). Since the Gulf of Mexico is a relatively warm body of water, air flowing over the Gulf is heated from below (especially prevalent in the fall and winter when the Gulf water temperatures average 3°F higher than the temperature of the overlying air near the water). This heating from below tends to increase the instability of the overlying air, enhancing diffusion. During the night, onshore flows of air warmed over the Gulf have a tendency to inhibit inversion formation over the land, further increasing the dilution potential of the

.atmosphere ([UFSAR] Ref. 2.3-1)."

Meteorological Data The CLB is based on meteorological data obtained from July 21, 1973, through September 30, 1977. This amendment uses meteorological data from the five-year period, (i.e., 2000 - 2004).

This data set was used in the used in the PAVAN and ARCON96 analyses discussed in this amendment.

4.1.1 Analysis of the 2000-2004 Meteorology Data The 2000-2004 hourly data was analyzed by ABS Consulting to ensure reasonableness and consistency. Only the lower level wind speed and direction from the 10m level and the 60-10m delta temperature were examined. The data were plotted and printed out for questionable periods. The questionable periods were examined more closely to verify the validity of the data.

Periods where data from the primary tower were missing were replaced with data from the backup tower. Replacement of the primary tower data with the backup tower data was verified by an ABS meteorologist.

The data contains 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of calm, 5 in 2001 (1 E stability from the N, 1 F stability from the N, 3 G stability from the SSE, WNW, and NW) and one in 2002 (G stability from the N). The decrease in the number of calms from the earlier data set is because of more sensitive instrumentation, so that a cutoff of 0.5 mph instead of 0.75 mph could be used.

When the final database was established, joint frequency distributions, wind roses and delta temperature plots were run to compare each year of data. The total wind frequency distribution for all stability classes is presented in Table 4.1-1. The joint frequency distributions by stability class are presented in Tables 4.1-2 through 4.1-8. A comparison of the distributions of stability classes with similar tables from the STP UFSAR1 are shown in Table 4.1-9.

1 JFSAR Revision 13, Tables 2.3-29 through -36.

NOC-AE-07002127 Attachment 1 Page 31 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION The results of these comparisons showed the five years of data were very consistent. There is a rather high percentage of unstable conditions over the five years, but it is consistent over the five year period. The stability class comparison in Table 4.1-9 shows about a five percent increase in A stability (from 7.6% to 12.4% of occurrences) in the recent data and about a five percent decrease in F/G stability (from 24.2% to 19.0% of occurrences). Section 2.3.2.1.3, Revision 14, of the STP UFSAR states that

"[UFSAR] Table 2.3-13 presents annual AT stability classifications for STPEGS data for the July 21, 1973 through September 30, 1977 period. These data indicate a predominance of neutral (D) to slightly stable (E) conditions. On an annual basis, stable conditions (E, F, and G) occur approximately 47 percent of the time while unstable conditions (A, B, and C) occur approximately 20 percent of the time."

From Table 4.1-9, for the average of the five years of data obtained from 2000 to 2004, stable conditions (E, F, and G) occurred approximately 46% of the time, similar to the 1973-1977 data.

Unstable conditions (A, B, and C) occurred approximately 25% of the time in the 2000-2.004 time period.

Table 4.1-10 provides a summary of the average wind speeds and peak wind directions during 2000-2004. A comparison of the wind speed distributions between the recent meteorological data and the CLB 2 data sets of data is shown on Figure 4.1-1. The 1 0 th and 20th percentiles compare well (approximately 4.2 mph for the 10th percentile for both data sets and 5.3 MPH for the 20th percentile). Even though the 8 0 th percentile shows some difference (approximately 12.6 mph for the more recent data as compared to approximately 15.2 mph for the CLB), it is the low wind speeds that would be controlling for the y/Q analysis.

The average wind speed for each year and the sector of the peak wind is presented in Table 4.1-

10. Wind roses for each year are provided in Figures 4.1-2 through 4.1-7. The distribution of wind directions are presented in Table 4.1-11 and are comparable to the CLB.

The delta temperature plots (Figures 4.1-8 through -12) show that the average delta temperature reading for the five year period was -0.3°F.

The overall five year data base is a consistent set of data in stability classification, wind direction, and wind speed and represents a good data set for use in atmospheric diffusion calculations.

2 From UFSAR, Rev 13, Table 2.3-29

NOC-AE-07002 127 Attachment 1 Page 32 of 219 4.1 METEOROLOGY AND A TMOSPHERIC DISPERSION Table 4. 1-1 Joint Frequency Distribution for 2000-2004 All Stability Classes Wind Speed (mph) (& l1in Wind Calm- 3.6- 7.6- 12.6- 18.6- 24.6- Average Direction Calm 3.5 7.5 12.5 18.5 .24.5 32.5 >32.5 Total Percent Speed N 3 298 916 1082 788 101 .3 0 3191 7.4 . 9.6 NNE 0 408 1300 1142 306 12 1 2 3171 7.4 7.5 NE 0 545 1502 1043 208 1 0 1 .3300 7.7 6.9 ENE 0 457 1146 712 147 2 0 0 2464 5.7 .6.7 E 0 403 1208 751 324 22 3 1 2712 6.3 7.5 ESE 0 377 1232 1073 608 54 3 1 3348 7.8 8.5 SE 0 248 1954 2177 1164 123 2 0 5668 13.2 9.4 SSE 1. 103 1757 2760 1414 .106 2 0 6143 14.3 10.0 S 0 65 1030 3034 1054 37 0 0 5220 12.2 10.1 SSW 0 35 675 872 139 3 0 0 1724 4.0 8.5 SW 0 36 282 416 52 2 0 0 788 1.8 8.3 WSW 0 37 131 133 26 1 0 0 328 0.8 7.6 W 0 71 275 66 20 0 0 0 432 1.0 5.9 WNW 1 197 395 93 29 1 0 0 716 1.7 5.4 NW 1 222 533 262 176. 25 0 0 1219 2.8 7.5 NNW 0 245 785 771 520 115 9 0 2445 5.7 9.4 Total 6 3747 15121 16387 6975 605 23 5 42869 100.0 Percent 0.0 8.8 35.3 38.2 16.3 1.4 0.1 0.0 100.0 Average Speed for This Table 8.7 Number of Invalid Hours 978 Number of Calm Hours for this Table 6 Number of Valid Hours for this Table 42869 Number of Variable Direction Hours for this Table 0 Total Hours for this Period 43847 Table 4.1-2 Joint Frequency Distribution for 2000-2004 Stability Class: A Extremely Unstable Wind Speed (mph) @ IOin Wind Calm- 3.6- 7.6- 12.6- 18.6- 24.6- Average Direction Calm 3.5 7.5 12.5 18.5 24.5 32.5 > 32.5 Total Percent Speed N 0 7 64 76 54 6 0 0 207 3.9 10.1 NNE 0 10 61 62 19 1 0 0 153 2.9 8.3 NE 0 8 50 72 12 0 0 0 142 2.7 8.3 ENE 0 3 34 59 12 1 0 0 109 2.0 8.9 E 0 3 26 34 30 0 0 0 93 1.7 10.0 ESE 0 8 29 90 116 7* 0 0 250 4.7 12.0 SE 0 2 34 307 241 45 0 0 629 11.8 12.6 SSE 0 4 52 319 337 54 0 0 766 14.4 12.9 5 0 6 109 1039 544 25 0 0 1723 32.3 11.5 SSW 0 6 105 297 86 3 0 0 -497 9.3 9.9 SW 0 6 53 101 24 .2 0 0 186 3.5 9.1 WSW 0 4 15 33 3 0 0 0 55 1.0 8.2 W 0 5 30 11 1 0 0 0 47 0.9 6.6 WNW 0 3 52 28 .7 0 0 0 90 1.7 7.4 NW 0 8 47 27 26 7 0 0 115 2.2 9.6 NNW 0 10 55 112 67 23 2 0 269 5.0 11.1 Total 0 93 816 2667 1579 174 2 0 15331 100.0 Percent 0.0 1.7 15.3 .50.0 29.6 3.3 0.0 0.0 100.0 Average Speed for This Table 11.1 Number of Invalid Hours 978 Number of Calm Hours for this Table 0 Number of Valid Hours for this Table 5331 Number of Variable Direction Hours for this Table 0 Total Hours for this Period 43847

NOC-AE-07002127 Attachment 1 Page 33 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION Table 4.1-3 Joint Frequency Distribution for 2000-2004 Stability Class: B Moderately Unstable Wind Speed (mph) @ 1Om Wind Calm- 3.6- 7.6- 12.6- 18.6- 24.6- Average Direction Calm 3.5 7.5 12.5 18.5 24.5 32.5 > 32.5 Total Percent Speed N 0 0 27 63 48 4 0 0 142 5.7 11.0 NNE 0 2 .40 70 28 0 0 0 140 5.6 9.4 NE 0 2 36 85 20 0 0 0 143 5.7 9.3 ENE 0 5 27 57 17 0 0 0 106 4.3 9.3 E 0 2 21 41 20 3 0 0 87 3.5 10.1 ESE 0 4 18 96 100 7 0 0 225 9.0 12.2 SE 0 0 27 203 132 23 0 0 385 15.4 12.2 SSE 0 1 33 186 119 12 0 0 351 14.1 11.8 S 0 1 75 244 87 1 0 0 408 16.4 10.3 SSW 0 0 55 52 11 0 0 0 118 4.7 8.3 SW 0 0 27 27 3 0 0 0 57 2.3 7.8 WSW .0 1 14 10 2 0 0 0 27 1.1 7.5 W 0 0 17 8 0 0 0 0 25 1.0 7.1 WNW 0 3 30 5 3 0 0 0 41 1.6 6.3 NW 0 1 26 27 21 5 0 0 80 3.2 10.4 NNW 0 2 32 57 55 12 1 0 159 6.4 11.5 Total 0 24 505 1231 666 67 1 0 2494 100.0 Percent 0.0 1.0 20.2 49.4 26.7 2.7 0.0 0.0 100.0 Average Speed for This Table 10.7 Number of Invalid Hours 978 Number of Calm Hours for this Table 0 Number of Valid Hours for this Table 2494 Number of Variable Direction Hours for this Table 0 Total Hours for this Period 43847 Table 4.1-4 Joint Frequency Distribution for 2000-2004 Stability Class: C Slightly Unstable Wind Speed (mph) @ I Om Wind Calm- 3.6- 7.6- 12.6- 18.6- 24.6- Average Direction Calm 3.5 7.5 12.5 18.5 24.5 32.5 > 32.5 Total Percent Speed M 7 f5 0lA IA 1. A 1 Q7 f 1 1,A NNE 0 1 52 74 36 2 0 0 165 6.0 9.7 NE 0 6 60 112 31 0 0 1 210 7.6 9.3 ENE 0 5 39 72 21 0 0 0 137 4.9 9.3 E 0 5 43 53 28 4 3 0 136 4.9 10.1 ESE 0 0 40 94 92 13 0 0 239 8.6 11.9 SE 0 5 41 236 173 22 0 0 477 17.2 11.9 SSE 0 1 47 212 137 12 0 0 409 14.8 11.5 S 0 3 61 185 67 0 0 0 316 11.4 10.0 SSW 0 1 45 41 6 0 0 0 93 3.4 8.2 SW 0 1 33 28 6 0 0 0 68 2.5 7.7 WSW 0 0 10 10 1 0 0 0 21 0.8 8.1 W 0 3 16 2 3 0 0 0 24 0.9 6.6 WNW 0 5 42 5 3 0 0 0 55 2.0 6.0 NW 0 3 32 24 21 3 0 0 83 3.0 9.6 NNW 0 1 29 56 43 17 5 0 151 5.4 12.4 Total 0 47 628 1270 733 83 9 1 2771 100.0 Percent 0.0 1.7 22.7 45.8 26.5 3.0 0.3 0.0 Average Speed for This Table 10.6 Number of Invalid Hours 978 Number of Calm Hours for this Table 0 Number of Valid Hours for this Table 2771 Number of Variable Direction Hours for this Table 0 Total Hours for this Period 43847

NOC-AE-07002127 Attachment 1 Page 34 of 219 4.1 METEOROLOGYAND ATMOSPHERICDISPERSION Table 4.1-5 Joint Frequency Distribution for 2000-2004 Stability Class: D Neutral Wind Speed (mph) @ 1Om Wind Calm- 3.6- 7.6- 12.6- 18.6- 24.6- Average Direction Calm 3.5 7.5 12.5 18.5 24.5 32.5 > 32.5 Total Percent Speed N 0 36 224 583 538 76 2 0 1459 11.5 11.7 NNE 0 32 282 535 203 7 0 0 1059 8.4 9.5 NE 0 55 283 505 125 1 0 0 969 7.6 8.9 ENE 0 37 204 334 80 1 0 0 656 5.2 8.8 E 0 37 190 316 200 14 0 1 758 6.0 10.1 ESE 0 35 207 444 246 22 2 1 957 7.6 10.3 SE 0 18 304 830 532 32 2 0 1718 13.6 10.9 SSE 0 16 267 1006 632 25 2 0 1948 15.4 11.2 S 0 15 194 721 276 8 0 0 1214 9.6 10.4 SSW 0 8 111 155 18 0 0 0 292 2.3 8.4 SW 0 7 41 87 12 0 0 0 147 1.2 8.6 WSW 0 5 32 24 10 0 .0 0 71 0.6 8.3 W 0 9 35 18 11 0 0 0 73 0.6 7.6 WNW 0 26 60 34 12 1 0 0 133 1.0 6.9 NW 0 32 117 90 89 9 0 0 337 2.7 9.4 NNW 0 29 170 316 306 55 1 0 877 6.9 11.4 Total 0 397 2721 5998 3290 251 9 2 12668 100.0 Percent 0.0 3.1 21.5 47.3 26.0 2.0 0.1 0.0 100.0 Average Speed for This Table . 10.3 Number of Invalid Hours 978 Number of Calm Hours for this Table 0 Number of Valid Hours for this Table 12668 Number of Variable Direction Hours for this Table 0 Total Hours for this Period 43847 Table 4.1-6 Joint Frequency Distribution for 2000-2004 Stability Class: E Slightly Stable -

Wind Speed (mph) @ 10m Wind Calm- 3.6- 7.6- 12.6- . 18.6- 24.6- Average Direction Calm 3.5 7.5 12.5 18.5 24.5 32.5 > 32.5 Total Percent Speed N 1 54 284 269 82 5 0' 0 695 6.1 8.2 NNE 0 50 313 302 20 2 1 2 690 6.0 7.5 NE 0 55 397 218 20 0 0 0 690 6.0 6.9 ENE 0 58 351 175 17 0 0 0 601 5.2 6.6 E 0 52 415 277 46 1 0 0 791 6.9 7.4 ESE 0 51 485 340 53 5 1 0 935 8.1 7.5 SE 0 36 847 580 86 1 0 0 1550 13.5 7.6 SSE 0 22 873 989 188 3 0 0 2075 18.1 8.5 S 0 17 456 823 79 3 0 0 1378 12.0 8.6 SSW 0 7 314 320 18 0 0 0 659 5.7 7.8 SW 0 10 111 165 7 0 0 0 293 2.6 8.1 WSW 0 11 40 44 8 1. 0 0 104 0.9 7.7 W 0 16 61 16 5 0 0 0 98 0.9 6.0 WNW 0 28 58 21 4 0 0 0 111 1.0 5.7 NW 0 38 121 68 19 1 0 0 247 2.2 6.9 NNW 0 52 243 209 49 8 0 0 561 4.9 8.0 Total 1 557 5369 4816 701 30 2 2 11478 100.0 Percent 0.0 4.9 46.8 42.0 6.1 0.3 0.0 0.0 100.0 Average Speed for This Table 7.8 Number of Invalid Hours 978 Number of Calm Hours for this Table 1 Number of Valid Hours for this Table 11478 Number of Variable Direction Hours for this Table 0 Total Hours for this Period 43847

NOC-AE-07002127 Attachment 1 Page 35 of 219 4.1 METEOROLOGY AND A TMOSPHERICDISPERSION Table 4. 1-7 Joint Frequency Distribution for 2000-2004 Stability Class: F Moderately Stable Wind Speed (mph) @ 1Om Wind Calm- 3.6- 7.6 12.6- 18.6- 24.6- Average Direction Calm 3.5 7.5 12.5 18.5 24.5 32.5 > 32.5 Total Percent Speed N 1 66 161 19 1 0 0 0 248 5.6 5.1 NNE 0 99 239 75 0 0 0 0 413 9.4 5.4 NE 0 103 268 32 0 0 0 0 403 9.2 4.7 ENE *0 115 247 13 0 0 0 0 375 8.5 4.4 E 0 141 311 26 0 0 0 0 478 10.9 4.6 ESE 0 159 304 8 1 0 0 0 472 10.7 4.3 SE 0 114 555 20 0 0 0 0 689 15.7 4.7 SSE 0 35 413 42 1 0 0 0 491 11.2 5.7 S 0 11 124 22 1 0 0 0 158 3.6 6.0 SSW 0 8 42 7 0 0 0 0 57 1.3 5.9 SW 0 5 16 8 0 0 0 0 29 0.7 6.1 WSW 0 11 17 12 2 0 0 0 42 1.0 6.1 W 0 22 62 6 0 0 0 0 90 2.0 4.6 WNW 0 38 57 0 0 0 0 0 95 2.2 4.0 NW 0 47 92 20 0 0 0 0 159 3.6 4.9

-NNW 0 48 138 17 0 0 0 0 203 4.6 5.0 Total I 1022 3046 327 6 0 0 0 4402 100.0

-Percent 0.0 23.2 69.2 7.4 0.1 0.0 0.0 0.0 100.0 Average Speed for This Table 4.9 Number of Invalid Hours 978 Number of Calm Hours for this Table I Number of Valid Hours for this Table 4402 Number of Variable Direction Hours for this Table 0 Total Hours for this Period 43847 Table 4.1-8 Joint Frequency Distribution for 2000-2004 Stability Class: G Extremely Stable Wind Speed (mph) @ 10m Wind Calm- 3.6- 7.6- 12.6- 18.6- 24.6- Average Direction Calm 3.5 7.5 12.5 18.5 24.5 32.5 > 32.5 Total Percent Speed N I 128 118 6 0 0 0 0 253 6.8 3.8 NNE 0 214 313 24 0 0 0 0 551 14.8 4.2 NE 0 316 408 19 0 0 0 0 743 19.9 4.1 ENE 0 234 244 2 0 0 0 0 480 12.9 3.7 E 0 163 202 4 0 0 0 0 369 9.9 3.9 ESE 0 120 149 1 0 0 0 0 270 7.2 3.8 SE 0 73 146 1 0 0 0 0 220 5.9 4.1 SSE I 24 72 6 0 0 0 0 103 2.8 4.9 S 0 12 11' 0 0 0 0 0 23 0.6 3.4 SSW 0 5 3 0 0 0 0 0 8 0.2 3.1 SW 0 7 1 0 0 0 0 0 8 0.2 2.1 WSW 0 5 3 0 0 0 0 0 8 0.2 3.6 W 0 16 54 5 0 0 0 0 75 2.0 4.7 WNW I 94 96 0 0 0 0 0 191 5.1 3.6 NW I 93 98 6 0 0 0 0 198 5.3 3.8 NNW 0 103 118 4 0 0 0 0 225 6.0 3.9 Total 4 1607 2036 78 0 0 0 0 3725' 100.0 Percent 0.1 43.1 54.7 2.1 0.0 0.0 0.0 0.0 100.0 Average Speed for This Table 4.0 Number of Invalid Hours 978 Number of Calm Hours for this Table 4 Number of Valid Hours for this Table 3725 Number of Variable Direction Hours for this Table 0 Total Hours for this Period 43847

NOC-AE-07002127 Attachment 1 Page 36 of 219 4.1 METEOROLOGY AND A TMOSPHERICDISPERSION Table 4.1-9 Stability Class Distribution

(%)

Updated Meteorological Data 3 Time Period Averages Stability CLB4 Class 2000 2001 2002 2003 2004 (1973-1977) 2000-2004 Difference A 14.6 (15) 15.1 (15) 11.3 (13) 12.4 (14) 8.4 (9) 7.6 12.4 4.8 B 5.8 (6) 5.4 (5) 7.3 (7) 5.3 (5) 5.3 (5) 6.0 5.8 -0.2 C 6.3 (6) 6.6 (7) 6.4 (6) 6.5 (6) 6.5 (6) 6.9 6.5 -0.4 D 27.5 (27) 27.2 (27) , 30.7 (31) 31.0 (31) 31.2 (31) 32.2 29.5 -2.7 E 31.0 (31) 23.2 (23) 24.9 (25) 25.4 (25) 29.5 (30) 23.1 26.8 3.7 F 8.7 (8) .11.5 (11) 10.3 (10) 10.2 (10) 10.7(11) 14.1 10.3 -3.8 G 6.1 (6) 11.0(11) 9.1 (9) 8.9 (9) 8.4 (9) 10.1 8.7 -1.4

% Data 97.2 95.7 99.7 99.9 96.3 - 97.8 -

0 STP percentage Table 4.1-10 Average Wind Speed and Peak Wind Direction: 2000-2004 Year Average Wind Speed (mph) Peak Wind Direction Sector 2000 9.4 SSE 2001 8.4 SE 2002 8.8 SSE 2003 8.1 S 2004 8.7 SSE 2000-2005 8.7 SSE 3 Based on joint frequency data for lOm wind speed, wind direction, and delta T 60-1-m.

4 From UFSAR, Rev 13, Table 2.3-13

NOC-AE-07002127 Attachment 1 Page 37 of 219 4.1 METEOROLOGYAND ATMOSPHERIC DISPERSION Table 4.1 -11 Wind Direction Distribution Wind Direction Percent of CLB 5 Percent of Recent Data NNE 7.1 .7.4 NE 7.3 7.7 ENE 5.4 5.8 E 5.7 6.3 ESE 6.4 7.8 SE 13.5 13.2 SSE 15.2 14.3 S 12.6 12.2 SSW 4.8 4.0 SW 2.3 1.8 WSW 1.1 0.8 W 1.3 1.0 WNW 1.3 1.7 NW 2.3 2.8 NNW 5.6 5.7 N 7.7 7.4 Calm 0.3 0.014 5 From UFSAR, Rev 13, Table 2.3-29, [Wind Frequency Distribution for All Observations]

NOC-AE-07002127 Attachment 1 Page 38 of 219 4.1 METEOROLOGY AND A TMOSPHERICDISPERSION Figure 4.1-1 Comparison of Wind Speed Distribution for STPEGS 100 90 80 .

70 -

60 50 40 -

30 20 " 10 -U-0O 0 t 0 5 10 15 20 25 MPH Figure 4.1-2 Wind Rose for 2000 WIND RIOSE PMdNDS FROM]

N ELegenVAWND V

SPEED LESS TIA.N3 5MPH WND SPEED LESS TKAN 7.5 MPH o WND SPEED LESS TWIZ5A15MPH WIND SPEED GREATER~ TH{AN 12.5MP

NOC-AE-07002127 Attachment 1 Page 39 of 219 4.1 METEOROLOGY AND A TMOSPHERIC DISPERSION Figure 4.1-3 Wind Rose for 2001 WIND ROSE (VANDS FROM)

N

$1

[Legwd CL a

WIND SPEED LESS THAN 35 MPH WIND SPEED LESS THAN 7.5 MPH WIND SPEED LESS T*HAN .5MPH WIND SPEED GRWEA'ER THAN 1Z5MPH Figure 4.1-4 Wind Rose for 2002 WND ROSE NAIINDS FROM)

N 16 Z e WIND SPEED LESS THAN 3.5PH 14IND SPEED LESS T-AN 3.5 MPH WND SPEED LESS THAN 2.5 MPH WIND SPEED 8RWATER THAN in5 MPH

NOC-AE-07002127 Attachment 1 Page 40 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION Figure 4.1-5 Wind Rose for 2003 WINZ ROSE WVNDS FROsI1 N

1I6 f A 0

WIND SPEED LESS THAN 3.5 MW4 WEED LESS THAI!7.5 MPH SPND WIND SPEED LESS THN U5 MPH

'W#ND SPEED GREATER ThAN 12-$ MPH

-I Figure 4.1-6 Wind Rose for 2004 WVND ROSE LWINDS FROM)

N

[Legerd I,

A 13 WIND VIND WIND WIND SPEED LESS THAN 3.5MPH SPEED LESS THAN 7,5 MPH SPEED LESS THAN IZSMPH SPEED GREATER THAN IZ5 HPH

NOC-AE-07002127 Attachment I Page 41 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION Figure 4.1-7 Wind Rose for 2000-2004 ViND ROSE WJNDS FROM)

N A r ND SPEED LESS THAN a5 mpH

• SPEED LESS THAN 7.5 MPH WIND 3 WIND SPEED LESS THAN 12.5 MPH WIND SPEED GREATER THAN 1Z5 MPH

NOC-AE-07002127 Attachment 1 Page 42 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION Figure 4.1-8 Delta-T Frequency for 2000 6P-Percent I*

DELTATEMP deg f Figure 4.1-9 Delta-T Frequency for 2001 I:

Percent 3: U Ed.1 8D 96 11-1 MB I8 1W 50 DELTA TEMP deg f

NOC-AE-07002127 Attachment 1 Page 43 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION Figure 4.1-10 Delta-T Frequency for 2002 63"ren Percent DELTATEMP deg f Figure 4.1-11 Delta-T Frequency for 2003 I

Pr2-Pereeni DELTA TEMP deg f

NOC-AE-07002127 Attachment 1 Page 44 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION Figure 4.1-12 Delta-T Frequency for 2004 13D 13.5-+-

I.!~

I.

I a Percent C0-I-Ii 14.

~1 ~

1 I I ifi Ii i - -

- I- I I I I

4,0 1 4.L I

-,a I

.42 I

-[js OD IIf I 32 I LB i 15.9 i am 1 SA 1 112- I 121 i I 11.

Ci I W

DELTATEMP deg f

NOC-AE-07002127 Attachment 1 Page 45 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION 4.1.2 Determination of the EAB and LPZ X/Qs Source/Receptor Scenarios and Configurations The minimum EAB and LPZ boundaries are located at 1430 m and 4800 m. Postulated releases do not qualify as elevated releases in accordance with Regulatory Guide 1.145; therefore, they were executed by PAVAN as "ground" type releases requiring an assumption of a 10 m release height. The minimum cross sectional area of the containment building used for the building wake calculation is 2734 m 2 . A containment height of 61.9 meters was used for the building wake factor in the annual average calculation.

Meteorological Data (PAVAN)

Meteorological data from the five-year period, (i.e., 2000 - 2004), were used in the PAVAN analysis. Independent of the consistency analysis performed on the meteorology data described in Section 4.1.1, the hourly data was processed into joint wind-stability occurrence frequency distribution for input into PAVAN. The only differences from the processing and checks described in Section 4.1.1 were:

" the inclusion of the calms;

• the F stability observations that became G stability because of the different classification scheme; and,

" the addition of the 36 mph upper bound so that a representative wind speed could be used for the highest speed bin.

The data were independently processed t6 develop PAVAN input with the following accounting:

• Total number of observations: 43847

• Total number of observations found to be invalid: 978

• Total number of F Stability observations from met data evaluation: 4402 (including 1 calm)

• Total number of G Stability observations from met data evaluation: 3725 (including 4 calms)

• Total number ofF Stability observations reclassified as G Stability for PAVAN: 157 (3.6% of F Stability observations from met data evaluation)

Total number ofF Stability observations for PAVAN: 4245

• Total number of G Stability observations for PAVAN: 3882 (including 5 calms)

Apparent Discrepancies in Stability Categories from Section 4.1.1 to the PAVAN Input When comparing the distributions from the consistency checks in Section 4. 1.1 to the PAVAN input, it appears that 153 observations (1 calm, 42 1-4 mph, 100 4-8 mph, and 10 8-13 mph observations) have been transferred from F Stability to G Stability. Safety Guide 23, Table 2, has the cutoffs for lapse rate, but does not clearly define what is to be done on the boundaries between the ranges. For instance, D is "-1.5 to -0.5"and E is "-0.5 to 1.5" °C/100m. F is defined as "1.5 to 4.0", and G is ">4.0". The ABS conversion program to convert the hourly data to PAVAN always puts cases on the boundary in the more stable category, making G ">=4.0". The

NOC-AE-07002127 Attachment 1 Page 46 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION temperatures supplied were 'F per 50m, so 3.6 in raw data is exactly 4.0 °C/1 00m. In the consistency checks performed for Section 4.1.1, these cases are classified as F, while the conversion program for PAVAN classified them as G. This is the source of the discrepancy, and leads to more conservative PAVAN results.

Wind Speed Categories Seven wind speed categories were defined according to Safety Guide 23 with the first category identified as "calm." The minimum wind speed (i.e., wind threshold) was set to 0.5 mph and "calm" wind speeds were distributed into the first wind speed group. The Safety Guide 23 wind speed categories and the categories used in this PAVAN analysis are presented in Table 4.1-12.

PAVAN requires an upper limit for the highest wind speed bin so as to have an average wind speed to use for computation. In the ABS conversion program, this is determined to be value for which the arithmetic mean wind speed in the bin matched the actual arithmetic mean wind speed of all of the hours in the bin.

NRC Regulatory Issues Summary 2006-4, Experience With Implementation OfAlternative Source Terms (Reference 43), states that:

The jointfrequency distributionsof wind speed, wind direction, and-atmosphericstability data used as input to PA VAN should have a large number of wind speed categories at the lower wind speeds in order to produce the best results (e.g., Section 4.6 of NUREG/CR-2858, "PA VAN: An Atmospheric Dispersion Program for Evaluating Design Basis Accidental Releases of Radioactive Materials from Nuclear Power Stations" (Ref 9),

suggests wind speed categories of calm, 0.5, 0.75, 1.0, 1.25, 1.5, 2.0, 3.0, 4.0 5.0, 6.0, 8.0 and 10.0 metkrs per second).

Section 4.6 of NUREG/CR-2858, "PAVAN: An Atmospheric Dispersion Program for Evaluating Design Basis Accidental Releases of Radioactive Materials from Nuclear Power Stations," provides suggestions for a set of wind speeds as follows:

"It has been found that ENVELOP produces the best results near the 0.5 percentile if the wind-speed data are classified into a large number of categories at the lower wind speeds, e.g., calm speed, 0.5, 0.75, 1.0, 1.25, 1.5, 2.0, 3.0, 4.0, 5.0, 6.0, 8.0, and 10.0 meters/second (see Card Type 11). The important aspect of having a large number of lower wind speed categories is to generate more X/Q values at the lower values of the cumulativefrequency since the 0.55% value is required."

NOC-AE-07002127 Attachment 1 Page 47 of 219 4.1 METEOROLOGYAND ATMOSPHERIC DISPERSION The guidance given with the input description (Table 3.1, Card Type 10; Card type 11 is actually the downwind boundary distances) is as follows:

"Maximum wind speed in each wind-speed category, in either miles/hour or meters/second. (If in miles/hour, set UCOR greater than 100). So that calms can be properly apportioneda direction, it is preferable that the first wind speed category have a maximum wind speed less than 1.5 meters/second."

ABS Consulting used the same seven wind speed groups for STP that are used in most of their work which are as follows: 0.5 mph, 3.5 mph, 7.5 mph, 12.5 mph, 18.5 mph, 24.5 mph, and 36.0 mph (i.e., 0.22 m/s, 1.56 m/s, 3.35 m/s, 5.59 m/s, 8.27 m/s, 10.95 m/s, and 16.09 m/s). These are appropriate for determining the offsite X/Q values using PAVAN.

Table 4.1-12 Defined Wind Speed Category Ranges For PAVAN Modeling Safety Guide 23 PAVAN-Assumed Category No. Speed Interval (mph) Maximum Speed (mph) 1 (Calm) 0 to< 1 0.56 2 1 to 3 3.5 3 4 to 7 7.5 4 8 to 12 12.5 5 13to 18 18.5 6 19 to 24 24.5 7 > 24 36.07 Joint Frequency data for the PAVAN input is presented in Tables 4.1-13 through 4.1-19. A summary is presented on Table 4.1-20.

Calculations PAVAN output summaries are provided on Tables 4.1-21, -22, and -23. The X/Q values for offsite locations were evaluated using the methods of Regulatory Guide 1.145. The offsite x/Q values recalculated for AST are presented in Table 4.1-24 along with the CLB values for comparison. The 0-2 hr z/Q is used for the worst 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses in the AST dose analyses. Note that in all cases, the revised X/Q values are more conservative than the CLB values.

6Calms are distributed into the first wind speed group.

7Determined from the data in the last wind speed group such that the actual mean wind speed in that group matched the average of the upper and lower wind speed limits of that group.

NOC-AE-07002127 Attachment 1 Page 48 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION Table 4.1-13 PAVAN Input:

Joint Frequency Distribution for 2000-2004 Stability Class: A Extremely Unstable Elevation: 1om Maximum Wind Speed (mph)

Wind Direction 0.5 3.5 7.5 12.5 18.5 24.5 36.0 Total N 7 64 76 54 6 0 0 207 NNE 10 61 62 19 1 0 0 153 NE 8 50 72 12 0 0 0 142 ENE 3 34 59 12 *1 0 0 109 E 3 26 34 30 0 0 0 93 ESE 8 29 90 116 7 0 0 250 SE 2 34 307 241 45 0 0 629 SSE 4 52 319 337 54 0 0 766 S 6 109 1039 544 25 0 0 1723 SSW 6 105 297 86 3 0 0 497 SW 6 53 101 24 2 0 0 186 WSW 4 15 33 3 0 0 0 55 W 5 30 11 1 0 0 0 47 WNW 3 52 28 7 0 0 0 90 NW 8 47 27 26 7 0 0 115 NNW 10 55 112 67 23 2 0 269 I

Total 93 816 2667 1579 174 2 0 5331

NOC-AE-07002127 Attachment 1 Page 49 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION Table 4.1-14 PAVAN Input:

Joint Frequency Distribution for 2000-2004 Stability Class: B Moderately Unstable Elevation: 1om Maximum Wind Speed (mph)

Wind Direction 0.5 3.5 7.5 12.5 18.5 24.5 36.0 Total N 0 27 63 48 4 0 0 142 NNE 2 40 70 28 0 0 0 140 NE 2 36 85 20 0 0 0 143 ENE 5 27 57 17 0 0 0 106 E 2 21 41 20 3 0 0 87 ESE 4 18 96 100 7 0 0 225 SE 0 27 203 132 23 0 0 385 SSE 1 33 186 119 12 0 0 351 S 1 75 244 87 1 0 0 408 SSW 0 55 52 11 0 0 0 118 SW 0 27 27 3 0 0 0 57 WSW 1 14 10 2 0 0 0 27 W 0 17 8 0 0 0 .0 25 WNW 3 30 5 3 0 0 0 41 NW 1 26 27 21 5 0 0 80 NNW 2 32 57 55 12 1 0 159 Total 24 505 1231 666 67 0 2494

NOC-AE-07002127 Attachment I Page 50 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION Table 4.1-15 PAVAN Input:

Joint Frequency Distribution for 2000-2004 Stability Class: C Slightly Unstable Elevation: 10m Maximum Wind Speed (mph)

Wind Direction 0.5 3.5 7.5 12.5 18.5 24.5 36.0 Total N 7 38 66 65 10 1 0 187 NNE 1 52 74 36 2 0 0 165 NE 6 60 112 31 0. 1 0 210 ENE 5 39 72 21 0 0 0 137 E 5 43 53 28 4 3 0 136 ESE 0 40 94 92 13 0 0 239 SE 5 41 236 173 22 0 0 477 SSE 1 47 212 137 12 0 0 409 S 3 61 185 67 0 0 0 316 SSW 1 45 41 6 0 0 0 93 SW 1 33 28 6 0 0 0 68 WSW 0 <10 10 1 0 0 0 21 W 3, 16 2 3 0 0 0 24 WNW 5 42 5 3 0 0 0 55 NW 3 32 24 21 3 0 0 83 NNW 1 29 56 43 17 5 0 151 Total 47 628 1270 733 83 10 0 2771

NOC-AE-07002127 Attachment 1 Page 51 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION Table 4.1-16 PAVAN Input:

Joint Frequency Distribution for 2000-2004 Stability Class: D Neutral Elevation: 1om Maximum Wind Speed (mph)

Wind Direction 0.5 3.5 7.5 12.5 18.5 24.5 36.0 Total N 36 224 583 538 76 2 1 1460 NNE 32 '282 535 203 7 0 0 1059 NE 55 283 505 125 1 0 0 969 ENE 37 204 334 80 1 0 0 656 E 37 190 316 200 14 1 0 758 ESE 35 207 444 246 22 3 0 957 SE 18 304 830 532 *32 2 0 1718 SSE 16 267 1006 632 25 2 0 1948 S 15 194 721 276 8 0 0 1214 SSW 8 111 155 18 0 0 0 292 SW 7 41 87 12 0 0 0 147 WSW 5 32 24 10 0 0 0 71 W 9 35 18 11 0 0 0 73 WNW 26 60 34 12 1 0 0 133 NW 32 117 90 89 9 0 0 337 NNW 29 170 316 306 55 1 0 877 Total 397 2721 5998 3290 251 11 1 12669

NOC-AE-07002127 Attachment 1 Page 52 of 219 4.1 METEOROLOGYAND ATMOSPHERIC DISPERSION Table 4.1-17 PAVAN Input:

Joint Frequency Distribution for 2000-2004 Stability Class: E Slightly Stable Elevation: lom Maximum Wind Speed (mph)

Wind Direction 0.5 3.5 7.5 12.5 18.5 24.5 36.0 Total N 54 284 269 82 5 0 0 695 NNE 50 313 302 20 2 3 0 690 NE 55 397 218 20 0 0 0 690 ENE 58, 351 175 17 0 0 0 601 E 52 415 277 46 1 0 0 791 ESE 51 485 340 53 5 1 0 935 SE 36 847 580 86 1 0 0 1550 SSE 22 873 989 188 3 0 0 2075 S 17 456 823 79 3 0 0 1378 SSW 7 314 320 18 0 0 0 659 SW 10 111 165 7 0 0 0 293 WSW 11 40 44 8 1 0 0 104 W 16 61 16 5 0 0 0 98 WNW 28 58 21 4 0 0 0 111 NW 38 121 68 19 1 0 0 247 NNW 52 243 209 49 8 0 0 561 Total 557 5369 4816 701 30 4 0 11478

NOC-AE-07002127 Attachment 1 Page 53 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION Table 4.1-18 PAVAN Input:

Joint Frequency Distribution for 2000-2004 Stability Class: F Moderately Stable Elevation: 1Gm Maximum Wind Speed (mph)

Wind Direction 0.5 3.5 7.5 12.5 18.5 24.5 36.0 Total N 63 158 19 1 0 0 2 243 NNE 92 228 72 0 0 0 0 392 NE 100 253 29 0 0 0 0 382 ENE 107 235 13 0 0 0 0 355 E 135 295 25 0 0 0 0 455 ESE 154 295 8 1 0 0 0 458 SE 110 543 20 0 0 0 0 673 SSE 33 406 42 1 0 0 1 483 S 11 121 22 1 0 0 0 155 SSW 7. 42 7 0 0 0 0 56 SW 5 16 8 0 0 0 0 29 WSW 10 16 12 2 0 0 0 40 W 20 62 5 0 0 0 0 87 WNW 37 56 0 0 0 0 1 94 NW 45 85 18 0 0 0 1 149 NNW 47 135 17 0 0 0 0 199 Total 976 2946 317 6 - 0 0 5 4250

NOC-AE-07002127 Attachment 1 Page 54 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION Table 4.1-19 PAVAN Input:

Joint Frequency Distribution for 2000-2004 Stability Class: G Extremely Stable Elevation: 1Om Wind Speed (mph)

Wind Direction 0.5 3.5 7.5 12.5 18.5 24.5 36.0 Total N 131 121 6 0 0 0 0 258 NNE 221 324 27 0 0 0 0 572 NE 319 423 22 0 0 0 0 764 ENE 242 256 2 0 0 0 0 500 E 169 218 5 0 0 0 0 392 ESE 125 158 1 0 0. 0 0 284 SE 77 158 1 0 0 0 0 236 SSE 26 79 6 0 0 0 0 111 S 12 14 0 0 '0 0 0 26 SSW 6 3 0 0 0 0 0 9 SW 7 1 0 0 0 0 0 8 WSW 6 4 0 0 0 0 0 10 W 18 54 6 0 0 0 0 78 WNW 95 97 0 0 0 0 0 192 NW 95 105 8 0 0 0 0 208 NNW 104 121 4 0 0 0 0 229 Total 1653 2136 88 0 0 0 0 3877

NOC-AE-07002127 Attachment 1 Page 55 of 219 4.1 METEOROLOGY AND A TMOSPHERICDISPERSION Table 4.1-20 Summary of PAVAN Input:

Joint Frequency Distribution for 2000-2004 Summary of All Stability Classes Elevation: 1om Maximum Wind Speed (mph)

Wind Direction 0.5 3.5 7.5 12.5 18.5 24.5 36.0 Total N 298 916 1082 788 101 3 3 3191 NNE 408 1300 1142 306 12 3 0 3171 NE 545 1502 1043 208 1 1 0 3300 ENE 457 1146 712 147 2 .0 0 2464 E 403 1208 751 324 22 4 0 2712 ESE 377 1232 1073 608 54 4 0 3348 SE 248 1954 2177 1164 123 2 0 5668 SSE 103 1757 2760 1414 106 2 1 6143 S 65 1030 3034 1054 37 0 0 5220 SSW 35 675 872 139 3 0 0 1724 SW 36 282 416 52 2 0 0 788 WSW 37 131 133 26 1 0 .0 328 W 71 275 66 20 0 0 0. 432 WNW 197 395- 93 29 1 0 1 716 NW 222 533 262 176 25 0 1 1219 NNW 245 785 771 520 115 9 0 2445 42869 15121 16387 6975 605 28 Total 3747 16387 6975 605 28 42869

NOC-AE-07002127 Attachment 1 Page 56 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION Table 4.1-21 Relative Concentration (X/Q) Values (sec/m 3) Versus Averaging Time

@EAB Hrs/yr 0-2hr X/Q Downwind 0-2 2-8 8-24 1-4 4-30 Annual is Exceeded in Sector hours hours hours days days Average Sector S 1.22E-4 5.49E-5 3.69E-5 1.56E-5 4 .53E-6 9.96E-7 39.4 SSW 1.23E-4 5.81E-5 3.99E-5 1.77E-5 5..49E-6 1.31E-6 29.5 SW 1.28E-4 6.20E-5 4.3 1E-5 1.96E-5 6 .34E-6 1.59E-6 43.7 WSW 1.25E-4 5.81E-5 3.97E-5 1.73E-5 5;.28E-6 1.23E-6 35.3 W 1.19E-4 5.57E-5 3.80E-5 1.66E-5 5;.07E-6 1.19E-6 23.5 WNW 1.15E-4 5.3 8E-5 3.68E-5 1.62E-5 4 .95E-6 1.16E-6 17.5 NW 1.12E-4 5.45E-5 3.80E-5 1.74E-5 5;.66E-6 1.43E-6 9.3 NNW 1.02E-4 4.94E-5 3.44E-5 1.57E-5 5;.06E-6 1.27E-6 27.3 N 5.53E-5 2.7 1E-5 1.89E-5 8.71E-6 2 .86E-6 7.30E-7 1.7 NNE 3.96E-5 1.78E-5 1.20E-5 5.02E-6 1 .45E-6 3.15E-7 1.0

, NE 2.85E-5 1.21E-5 7.86E-6 3.1OE-6 8 .12E-7 1.58E-7 1.3 ENE 1.21E-5 5.43E-6 3.64E-6 1.52E-6 4r.35E-7 9.38E-8 1.0 E 3.92E-5 1.63E-5 1.05E-5 4.04E-6 1 .03E-6 1.92E-7 1.9 ESE 9.55E-5 3.91E-5 2.50E-5 9.49E-6 2 .36E-6 4.30E-7 29.2 SE 1.04E-4 4.36E-5 2.83E-5 1.1OE-5 22.86E 5.48E-7 30.3 SSE 1.12E-4 4.90E-5 3.24E-5 1.32E-5 3 .65E-6 7.55E-7 14.9 Max X/Q 1.28E-4 Total hours around Site: 306.8 SRP 2.3.4 2.02E-4 9.07E-5 6.08E-5 2.55E-5 7.32E-6 1.59E-6 Site Limit 1.44E-4 6.84E-5 4.71E-5 2. 1OE-5 6.58E-6 1.59E-6 DISTANCE: 1430 m WIND SENSORS HEIGHT: 1Om TYPE OF RELEASE: Ground-level Release DELTA-T HEIGHTS: 10.0 - 60.0 m

NOC-AE-07002127 Attachment 1 Page 57 of 219 4.1 METEOROLOGY AND A TMOSPHERICDISPERSION Table 4.1-22 Relative Concentration (X/Q) Values (sec/rn 3) Versus Averaging Time

@ LPZ Hrs/yr 0-2hr X/Q Downwind 0-2 2-8 8-24 1-4 4-30 Annual is Exceeded in Sector hours hours hours days days 9 .52E-7 Average Sector S 3.54E-5 1.48E-5 9.57E-6 3.71E-6 1.80E-7 21.2 SSW 4.59E-5 1.93E-5 1.25E-5 4.88E-6 1.27E-6 2.43E-7 29.5 SW 5.27E-5 2.24E-5 1.46E-5 5.75E-6 1.51E-6 2.95E-7 43.7 WSW 4.79E-5 1.98E-5 1.28E-5 4.90E-6 1.24E-6 2.30E-7 35.3 W 4.12E-5 1.74E-5 1.13E-5 4.43E-6 1.16E-6 2.23E-7 23.5 WNW 3.62E-5 1.55E-5 1.02E-5 4.07E-6 1.09E-6 2.17E-7 17.5 NW 3.3 1E-5 1.49E-5 9.98E-6 4.19E-6 1.21E-6 2.64E-7 9.3

--NNW 2.64E-5 1.21E-5 8.14E-6 3.47E-6 1.02E-6 2.28E-7 12.1 N 1.32E-5 6.13E-6 4.18E-6 1.82E-6 5.50E-7 1.27E-7 1.6 NNE 9.23E-6 3.97E-6 2.60E-6 1.04E-6 2.79E-7 5.57E-8 0.8 NE 6.09E-6 2.50E-6 1.60E-6 6.08E-7 1.52E-7 2.77E-8 1.1 ENE 2.1OE-6 9.47E-7 6.36E-7 2.69E-7 7.80E-8 1.72E-8 0.9 E 8.55E-6 3.46E-6 2.20E-6 8.27E-7 2.02E-7 3.62E-8 1.9 ESE 2.72E-5 1.04E-5 6.41E-6 2.25E-6 5.02E-7 7.99E-8 14.9 SE 2.94E-5 1.15E-5 7.21E-6 2.61E-6 6.05E-7 1.OIE-7 15.4 SSE 3.28E-5 1.33E-5 8.44E-6 3.16E-6 7.72E-7 1.38E-7 14.9 Max X/Q 5.27E-5 Total hours around Site: 243.7 SRP 2.3.4 6.39E-5 2.63E-5 1.68E-5 6.42E-6 1.61E-6 2.95E-7 Site Limit 4.29E-5 1.88E-5 1.25E-5 5.10E-6 1.42E-6 2.95E-7 DISTANCE: 4800 m WIND SENSORS HEIGHT: lom TYPE OF RELEASE: Ground-level Release DELTA-T HEIGHTS: 10.0 - 60.0 m

NOC-AE-07002127 Attachment 1 Page 58 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION Table 4.1-23 X/Q Values Based on DT(60M-10M) Stability Data and 10 Meter Winds January 1, 2000 - December 31, 2004 Averaging Maximum 9 5 Percent 50 Percent 1

Time0 Distance Sector Overall Overall Worst Case ° 2.02 x1.

1-hour Minimum Exclusion Area 1.28 x 10-4 3.19x 10. 8.59 x 104 2-hour Boundary (EAB)- (SW) (ESE,SE,S,NNW) 1430 meters 3.67 x 106' 1-hour Low Population Zone (LPZ)- 5.27 x 10-' 6.40 x 10.' 3.18 x 104 2-hour 4800 meters (SW) 5 (ESE,SE,S,NNW) 8-hour Low Population Zone (LPZ)- 2.24 x 10. 2.63 x 10-5 4800 meters (SW) 16-hour Low Population Zone (LPZ)- 1.46 x.10 5 1.68 x 10-5 4800 meters (SW) 72-hour Low Population Zone (LPZ)- 5.75 x 10.6 6.42 x 10.6 4800 meters (SW) 624-hour Low Population Zone (LPZ)- 1.51 x 10-6 1.61 x 10-6 4800 meters (SW)

The directions for the sectors given above are the directions of the "Affected Sectors" (i.e., wind from the east will affect a west sector).

Table 4.1-24 X/Q Values for Radiological Dose Calculations - EAB and LPZ (sec/rn 3)

EAB (1430m) LPZ (4800m)

Updated Met Data Updated Met Data Time Interval CLB (PAVAN) CLB (PAVAN) 0-2 hrs 1.3E-4 1.44E-4 3.8E-5 5.27E-5 2-8 hrs N/A N/A 1.6E-5 2.24E-5 8-24 hrs N/A N/A l.lE-5 1.46E-5 1-4 days N/A N/A 4.3E-6 5.75E-6 4-30 days N/A N/A 1.2E-6 1.51E-6 8The 1-hour value was calculated; the 2-hour value is assumed to be equal to the 1-hour value.

9 Maximum sector values are the highest 0.5 percent Sector X/Q values.

10 Worst case values are the highest calculated 1-hour values, all sectors considered.

NOC-AE-07002127 Attachment 1 Page 59 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION 4.1.3 Control Room and Technical Support Center X/Q Analyses For each unit at STP, there are three release points:

" the containment building outer wall surface;

• the Plant Vent; and,

" the SG power operated relief valve (PORV) nearest the Control Room intake (this is also the area of the steam release for a postulated MSLB)

These points are illustrated in Figure 4.1-13.

Figure 4.1-13 Simplified Plot Plan with Release Points and Receptors Unit 2 1 Unit 1 North C I H D

NOC-AE-07002127 Attachment 1 Page 60 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION Table 4.1-25 Key to Figure 4.1-13: Release and Receptor Locations Release Source Figure 4.1-13 Label Applicable Accidents Unit 1 Containment A LOCA, CREA Unit 1 Plant Vent B LOCA (ESF leakage)

LOCA (supplemental purge)

FHA in FHB or RCB Unit 1 East PORV/IVC C MSLB, SGTR, LRA Unit 1 Control Room/TSC D All Intake Unit 2 Containment E LOCA, CREA Unit 2 Plant Vent F LOCA (ESF leakage)

LOCA (supplemental purge)

FHA in FHB or RCB Unit 2 East PORV/IVC G MSLB, SGTR, LRA Unit 2 Control Room/TSC H All Intake"l For all postulated accidents, steam releases from the secondary system (including the MSLB) are all assumed to occur in the Isolation Valve Cubicle (IVC), located between the containment building and the turbine building. This structure houses the main steam lines, the safety relief valves and the SG PORVs. The distance from the closest SG PORV to the control room HVAC emergency intake was used as the basis for the PORV-to-CRE x/Q. Since this maximizes the X/Q, a X/Q for each PORV, steam line, or safety relief valve was not generated. The PORV-to-CRE X/Q is used for all secondary system steam releases.

Releases from the Fuel Handling Building (for the Fuel handling Accident and the LOCA ESF leakage) are vented to the atmosphere via the Plant Vent. The RCB normal and supplemental purge is also via the same Plant Vent. Therefore, for the FHA releases and the LOCA supplemental purge release, the Plant-Vent-to-Control Room X/Q is used. Releases from the RCB Personnel Airlock are also exhausted via this Plant Vent. The Plant-Vent-to- Control Room X/Q also bounds a release from the RCB Equipment Hatch opening since the Plant Vent is much closer to the Control room air intake than the Equipment Hatch (which is located on the southwest quadrant of the RCB).

Each unit at STP has two associated receptors, the Control Room Emergency Makeup Air Intake and the Electrical Auxiliary Building Air Intake. The Control Room Emergency Makeup Air Intake is the air intake for both the Control Room and the Technical Support Center (TSC)

HVAC systems.

" On each unit, the EAB HVAC Intake is immediately South of the Control Room/TSC intake.

NOC-AE-07002127 Attachment 1 Page 61 of 219 4.1 METEOROLOGYAND ATMOSPHERIC DISPERSION The control room and TSC are both wholly contained within the Electrical Auxiliary Building.

Therefore, unfiltered in-leakage entering either the Control Room or the TSC would come from the Electrical Auxiliary Building atmosphere. Since the Electrical Auxiliary Building Air Intake is adjacent to the Control Room Emergency Makeup Air Intake, the X/Q values calculated for the Control Room/TSC are also used for the Electrical Auxiliary Building and the unfiltered in-leakage entering either the Control Room or the TSC.

The postulation of a loss of offsite power does not change the location of release points or receptor locations. Steam releases from the secondary side are conservatively assumed to be released through the PORVs in the IVC. This is closer to the CR/TSC HVAC intake than any release points in the Turbine Generator Building (TGB).

Updated Control Room X/Q values for releases from the containment, from the plant vent, and from the PORV area were calculated using the computer code ARCON96 (Reference 18) using the methods of Regulatory Guide 1.194. The STP meteorological databases for the five-year period (2000 - 2004) were used in the ARCON96 modeling analysis. Wind measurements were taken at 10 m and the vertical temperature difference was measured between 60 m and 10 m. The minimum wind speed (i.e., wind threshold) was set to the ARCON96 default value of 0.5 m/sec in accordance with Regulatory Guide 1.194, Table A-2.

ARCON96 requires the direction from the receptor to the source. Because plant north and true north are aligned, there is no need to correct directions. Using guidance from Section 3.2.4.5 of Regulatory Guide 1.194, the containment surface releases are taken to be on the surface of the containment at the horizontal location closest to the receptor. The release elevation for containment surface releases, using the Section 3.2.4.5 of Regulatory Guide 1.194, is the vertical center of the above-grade portion of the containment projected on a plane tangent to the containment surface and perpendicular to the line of sight from the containment center to the intake. Accordingly, the elevation of the containment leakage is determined to be 129.5 feet.

Using Sections 3.2.4.4 and 3.2.4.5 of Regulatory Guide 1.194, the initial sigmas for the containment surface source are:

Yyo = 26' 4" = 8.03 m, and ay¥ 0 = 33' 10"'= 10.31 m.

To determine the building area, the guidance in Regulatory Guide 1.194, Table A-2, was followed. The area to be used for each release.point was chosen to be the vertical cross-section of the building that has the largest impact on the building wake for the release point. For all of the release points considered, the largest impact on the building wake is the containment building. The containment is treated as a right cylinder surmounted by half of a spheroid with horizontal radius equal to the cylinder, radius and vertical radius equal to the height difference between the containment spring line and the top of the containment. The grade elevation is

NOC-AE-07002127 Attachment 1 Page 62 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION 28'0", the containment spring line is 153'0", and the top of the containment is 231'0". The containment radius is 79'0". The resulting area of the containment is 29,429 ft2.

The plant grid system is used to place release and receptor locations on a Cartesian coordinate grid. With the grid data, distances are computed in two dimensions (x,y) only. Distances between release locations and receptor points are presented in Table 4.1-26. All releases are treated as point sources, with the exception of containment leakage, which is treated as a diffuse source. The height of these release points are all less than 2.5 times the height of their adjacent buildings and therefore, in accordance with Regulatory Guide 1.194, are modeled as "ground level" releases. Buoyancy or mechanical jets of high energy releases are not credited in the X /Q analyses.

A X/Q value was determined from each release point in both units to each receptor in that unit.

The maximum X/Q value for a release point/receptor point pair in one unit was used in the analyses (for example, the X/Q value from the PORV in Unit 2 to the Control Room in Unit 2, and the x/Q value from the RCB in Unit 1 to the Control Room in Unit 1). The 0-2 hr X/Q is used for the worst 2-hour doses.

Tables 4.1-26 through 4.1-29 present data that was used to develop the ARCON96 analyses.

Table 4.1-30 through 4.1-35 present ARCON96 results. A summary of the resulting X/Qs for each source/receptor pair is presented in Table 4.1-36. Table 4.1-37 presents a summary of the ARCON96 results used in the radiological analyses.

Table 4.1-26 Geometric Relationships Between Release Locations and Receptors Control Direction to Release Receptor Room for Distance Source Height Height Release Location Unit (in) (°) (in) (M)

Ul RCB Leakage 1 62.44 274.45 30.94 16.46 Ul Plant Vent 1 62.50 240.81 21.03 16.46 Ul East PORV 1 84.39 292.11 20.73 16.46 U2 RCB Leakage 2 62.14 274.46 30.94 16.46 U2 Plant Vent 2 62.50 240.81 21.03 16.46 U2 East PORV 2 84.11 292.19 20.73 16.46

NOC-AE-07002127 Attachment 1 Page 63 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION Table 4.1-27 Data Used to Generate ARCON96 Inputs Parameter Value Containment diameter 158 feet Containment height 203 feet Area of containment 29,429 ft2 Surface roughness length 0.2m Minimum wind speed 0.5 m/s Wind direction window 900 Averaging sector width 4.3 Distances between Release points and See Table 4.1-26 Receptors Number of hours in the averages and the minimum number of hours:

ARCON96 defaults:

Hours Minimum 1 1 2 2 4 4 8 8 12 11 24 22 96 87 168 152 360 324 720 648

NOC-AE-07002127 Attachment 1 Page 64 of 219 4.1 METEOROLOGYAND ATMOSPHERIC DISPERSION Table 4.1-28 ARCON96 Input:

Unit 1 Releases to Unit 1 Control Room/TSC Release Source Parameter RCB Plant Vent East PORV Height of lower wind speed instrument (in) 10 Height of upper wind speed instrument (in) 60 Wind speed units Miles per hour Release type Ground level Release Height (m) 30.9 21.0 20.7 2

Building Area (m ) 2734.0 Effluent vertical velocity (m/s) 0.0 Vent or Stack Flow (m3/s) 0.0 Vent or Stack radius (in) 0.0 Direction: Intake to Source (deg) 274 241 292 Wind Direction Sector Width (deg) 90 Wind Direction Window (deg) .229 - 319 196-286 247-337 Distance to Intake (in) 62.4 62.5 84.4 Intake Height (in) 16.5 Terrain Elevation Difference (in) 0.0 Minimum Wind Speed (m/s) 0.5 Surface Roughrness Length (in) 0.20 Sector Averaging Constant 4.3 Initial value of Sigma Y 8.03 _ 0.0 0.0 Initial value of Sigma Z 10.31 0.0 0.0

NOC-AE-07002127 Attachment 1 Page 65 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION Table 4.1-29 ARCON96 Input:

Unit 2 Releases to Unit 2 Control Room/TSC Release Source Parameter RCB Plant Vent East PORV Height of lower wind speed instrument (in) 10 Height of upper wind speed instrument (in) 60 Wind speed units Miles per hour Release type Ground level Release Height (m) 30.9 21.0 20.7 Building Area (m 2 ) 2734.0 Effluent vertical velocity (m/s) 0.0 Vent or Stack Flow (m3 /s) 0.0 Vent or Stack radius (m) 0.0 Direction: Intake to Source (deg) 274 241 292 Wind Direction Sector Width (deg) 90 Wind Direction Window (deg) 229-319 196-286 247-337 Distance to Intake (in) 62.1 62.5 84.1 Intake Height (in) 16.5 Terrain Elevation Difference (in) 0.0 Minimum Wind Speed (m/s) 0.5 Surface Roughness Length (in) 0.20 Sector Averaging Constant 4.3 Initial value of Sigma Y 8.03 0.0 0.0 Initial value of Sigma Z 10.31 0.0 0.0

NOC-AE-07002127 Attachment 1 Page 66 of 219 Table 4.1-30 ARCON96 Results Unit 1 Containment to Unit 1 Control Room/TSC Total number of hours of data processed 43848 Hours elevated plume w/dir. in window 0 Hours of missing data 885 Hours of calm wind 123 Hours direction in window 2391 Hours direction not in window or calm 40449 Averaging Period (hours) 1 2 4 8 12 24 96 168 360 720 Upper Limit 1.OOE-3 1.OOE-3 1.OOE-3 1.OOE-3 L.OOE-3 1.OOE-3 1.OOE-3 1.OOE-3 1.OOE-3 I.OOE-3 Lower Limit 1.OOE-7 1.OOE-7 1.OOE-7 1.OOE-7 I.00E-7 1.OOE-7 1.OOE-7 1.OOE-7 1.OOE-7 1.OOE-7 Above range 0. 0. 0. 0. 0. 0. 0. 0. 0. 0.

In Range 2514. 3295. 4573. 6680. 8556. 13048. 27768. 34351. 38630. 38820.

Below Range 0. 0. 0. 0. 0. 0. 0. 0. 0. 0.

Zero 40449. 39543. 38071. 35668. 33915. 29186. 13589. 6481. 1890. 52.

Total X/Qs 42963. 42838. 42644. 42348. 42471. 42234. 41357. 40832. 40520. 38872.

% non-Zero 5.85 7.69 10.72 15.77 20.15 30.89 67.14 84.1'3 95.34 99.87 95%-ilex/Q 2.16E-4 2.12E-4 1.76E-4 1.57E-4 1.26E-4 9.30E-5 5.41E-5 4.39E-5 3.46E-5 2.71E-5 95% x/Q for standard averaging intervals Hourly Value Range 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2.16E-4 Max x/Q Min x/Q 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.37E-4 Centerline 5.67E-4 7.60E-5 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 6.11E-5 Sector-Average- 3.30E-4 4.43E-5 I to 4 days 4.11E-5 4 to 30 days 2.30E-5 Table 4.1-31 ARCON96 Results Unit 1 Plant Vent to Unit 1 Control Room/TSC Total number of hours of data processed 43848 Hours elevated plume w/dir. in window 0 Hours of missing data 885 Hours of calm wind 123 Hours direction in window 2987 Hours direction not in window or calm 39853 Averaging Period (hours) 1 2 4 8 12 24 96 168 360 720 Upper Limit 1.OOE-2 1.OOE-2 1.OOE-2 1.OOE-2 1.OOE-2 1.OOE-2 1.00E-2 I.OOE-2 1.OOE-2 1.00E-2 Lower Limit 1.OOE-6 1.OOE-6 I.OOE-6 1.OOE-6 1.OOE-6 1.OOE-6 1.OOE-6 1.OOE-6 1.OOE-6 1.OOE-6 Above range 0. 0. 0. 0. 0. 0. 0. 0.

In Range 3310. 3988. 5435. 7816. 9936. 14604. 28578. 34340. 38553. 38871.

Below Range 0. 0. 0. 0. 0. 0. 495, 1.

Zero 39853. 38850. 37209. 34532. 32535. 27630. 12779. 6492. 1472. 0.

Total X/Qs 42963. 42838. 42644. 42348. 42471. 42234. 41357. 40832. 40520. 38872.

% non-Zero 7.24 9.31 12.75 18.46 23.39 34.58 69.10 84.10 96.37 100.00 95%-ileX/Q 6.76E-4 7.12E-4 6.20E-4 5.74E-4 4.58E-4 3.27E-4 2.03E-4 1.78E-4 1.45E-4 1.12E-4 95% X/Q for standard averaging intervals Hourly Value Range 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 7.12E-4 Max X/Q Minx/Q 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 5.28E-4 Centerline 2.05E-3 1.62E-4 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 2.04E-4 Sector-Average 1.19E-3 9.47E-5 I to 4 days 1.61E-4 4 to 30 days 9.76E-5

NOC-AE-07002127 Attachment 1 Page 67 of 219 Table 4.1-32 ARCON96 Results Unit 1 East PORV to Unit 1 Control Room/TSC Total number of hours of data processed 43848 Hours elevated plume w/dir. in window 0 Hours of missing data 885 Hours of calm wind 123 Hours direction in window 3481 Hours direction not in window or calm 39359 Averaging Period (hours) 1 2 4 8 12 24 96 168 360 720 Upper Limit 1.00E-2 I.OOE-2 I.OOE-2 1.OOE-2 I.OOE-3 1.OOE-3 1.OOE-3 1.OOE-3 1.00E-3 1.OOE-3 Lower Limit 1.00E-6 l.OOE-6 1.OOE-6 1.OOE-6 1.OOE-7 L.OOE-7 1.OOE-7 l.00E-7 1.OOE-7 L.00E-7 Above range 0. 0. 0. 0. - 0. 0. 0. 0.

In Range 3604. 4543. 6063. 8422. 10472. 15255. 30100. 35798. 39341. 38821.

Below Range 0. 0. 0. 0. 0. 0. . 0. 0.

Zero 39359. 38295. 36581. 33926. 31999. 26979. 11257. 5034. 1179. 51.

Total X/Qs 42963. 42838. 42644. 42348. 42471. 42234. 41357. 40832. 40520. 38872.

% non-Zero 8.39 10.61 14.22 19.89 24.66 36.12 72.78 87.67 97.09 99.87 95%-ilex/Q 6.08E-4 4.71E-4 4.40E-4 3.96E-4 3.22E-4 2.35E-4 1.34E-4 1.09E-4 9.07E-5 7.97E-5 95% X/Q for standard averaging intervals Hourly Value Range 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 6.08E-4 Max X/Q Min X/Q 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3.26E-4 Centerline 1.15E-3 7.69E-5 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.54E-4 Sector-Average 6.7 1E-4 4.48E-5 1 to 4 days 1.OOE-4 4 to 30 days 7.13E-5 Table 4.1-33 ARCON96 Results Unit 2 Containment to Unit 2 Control RoorrTSC Total number of hours of data processed 43848 Hours elevated plume w/dir. in window 0 Hours of missing data 885 Hours of calm wind 123 Hours direction in window 2391 Hours direction not in window or calm 40449 Averaging Period (hours) 1 2 4 8 12 24 96 168 360 720 Upper Limit 1.OOE-3 1.OOE-3 1.OOE-3 1.OOE-3 1.OOE-3 1.OOE-3 1.OOE-3 1.OOE-3 1.00E-3 1.OOE-3 Lower Limit 1.OOE-7 1.00E-7 l.OOE-7 1.OOE-7 l.OOE-7 l.OOE-7 L.OOE-7 l.OOE-7 1.OOE-7 1.OOE-7 Above range 0. 0. 0. 0. 0. 0. 0. 0.

In Range 2514. 3295. 4573. 6680. 8556. 13048. 27768. 34351. 38630. 38820.

Below Range 0. 0. 0. 0. 0. 0. 0. 0.

Zero 40449. 39543. 38071. 35668. 33915. 29186. 13589. 6481. 1890. 52.

Total X/Qs 42963. 42838. 42644. 42348. 42471. 42234. 41357. 40832. 40520. 38872.

% non-Zero 5.85 7.69 10.72 15.77 20.15 30.89 67.14 84.13 95.34 99.87 95%-ile X/Q 2.17E-4 2.13E-4 1.77E-4 1.57E-4' 1.27E-4 9.34E-5 5.44E-5 4.40E-5 3.49E-5 2.72E-5 95% x/Q for standard averaging intervals Hourly Value Range 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2.17E-4 Max X/Q Min X/Q 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.37E-4 Centerline 5.70E-4 7.64E-5 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 6.15E-5 Sector-Average 3.32E-4 4.45E-5 1 to 4 days 4.14E-5 4 to 30days' 2.30E-5

NOC-AE-07002127 Attachment 1 Page 68 of 219 Table 4.1-34 ARCON96 Results Unit 2 Plant Vent to Unit 2 Control Room/TSC Total number of hours of data processed 43848 Hours elevated plume w/dir. in window 0 Hours of missing data 885 Hours of calm wind 123 Hours direction in window 2987 Hours direction not in window or calm 39853 Averaging Period (hours) 1 2 4 8 12 24 96 168 360 720 Upper Limit 1.OOE-2 1.00E-2 1.OOE-2 1.QOE-2 1.OOE-2 1.OOE-2 1.OOE-2 1.OOE-2 1.OOE-2 1.OOE-2 Lower Limit 1.00E-6 1.OOE-6 1.OOE-6 1.OOE-6 1.OOE-6 1.OOE-6 1.OOE-6 1.00E-6 1.OOE-6 1.OOE-6 Above range 0. 0. 0. 0. 0. 0. 0. 0.

In Range 3110. *3988. 5435. 7816. 9936. 14604. 28578. 34340. 38553. 38871.

Below Range 0. 0. 0. 0. 0. 0. 495. 1.

Zero 39853. 38850. 37209. 34532. 32535. 27630. 12779. 6492. 1472. 0.

Total X/Qs 42963. 42838. 42644. 42348. 42471. 42234. 41357. 40832. 40520. 38872.

%non-Zero 7.24 9.31 12.75 18.46 23.39 34.58 69.10 84.10 96.37 100.00 95%-ile X/Q 6.76E-4 7.12E-4 6.20E-4 5.74E-4 4.58E-4 3.27E-4 2:03E-4 1.78E-4 1.45E-4 1.12E-4 95% x/Q for standard averaging intervals Hourly Value Range 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 7.12E-4 Max X/Q Min X/Q 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 5.28E-4 Centerline 2.05E-3 1.62E-4 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 2.04E-4 Sector-Average 1.19E-3 9.47E-5 I to 4 days 1.61E-4 4 to 30 days 9.76E-5 Table 4.1-35 ARCON96 Results Unit 2 East PORV to Unit 2 Control Room/TSC Total number of hours of data processed 43848 Hours elevated plume w/dir. in window 0 Hours of missing data 885 Hours of calm wind 123 Hours direction in window 3481 Hours direction not in window or calm 39359 Averaging Period (hours) 1 2 4 8 12 24 96 168 360 720 Upper Limit I.OOE-2 I.OOE-2 I.OOE-2 I.OOE-2 I.OOE-3 I.OOE-3 I.OOE-3 I.OOE-3 I.OOE-3 1.00E-3 Lower Limit 1.OOE-6 I.OOE-6 1.OOE-6 1.OOE-6 1.OOE-7 I.OOE-7 1.OOE-7 1.OOE-7 1.OOE-7 1.OOE-7 Above range 0. 0. 0. 0. 0. 0. 0. 0.

In Range 3604. 4543. 6063. 8422. 10472. 15255. 30100. 35798. 39341. 38821.

Below Range 0. 0. 0. 0. 0. 0. 0. 0.

• Zero 39359. 38295. 36581. 33926. 31999. .26979. 11257. 5034. 1179. 51L Total X/Qs 42963. 42838. 42644. 42348. 42471. 42234. 41357. 40832. 40520. 38872.

%non-Zero 8.39 10.61 14.22 19.89 24.66 36.12 72.78 87.67 97.09 99.87 95%-ileX/Q 6.13E-4 4.73E-4 4.43E-4 3.98E-4 3.23E-4 2.36E-4 1.34E-4 1.1OE-4 9.09E-5 8.02E-5 95% X/Q for standard averaging intervals Hourly Value Range 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 6.13E-4 'Max x/Q Min X/Q 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3*27E-4 Centerline 1.16E-3 7.74E-5 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.55E-4 Sector-Average 6.76E-4 4.51E-5 I to 4 days 1.01E-4 4 to 30 days 7.18E-5

NOC-AE-07002127 Attachment 1 Page 69 of 219 4.1 METEOROLOGY AND ATMOSPHERIC DISPERSION The 95%-ile X/Qs from ARCON96 are summarized in Table 4.1-36. The maximum X/Q for a release location was chosen for use in the dose analyses.

Table 4.1-36 Control Room/TSC 95 Percentile X/Qs (sec/m 3)

Time Interval Release Location 0-2 hours 2-8 hours 8-24 hours 1-4 days 4-30 days Ul RCB Leakage 2.16E-04 1.37E-04 6.11E-05 4.11E-05 2.30E-05 Ul Plant Vent 7.12E-04 5.28E-04 2.04E-04 1.61E-04 9.76E-05 U1 East PORV 6.08E-04 3.26E-04 1.54E-04 1.OOE-04 7.13E-05 U2 RCB Leakage 2.17E-04 1.37E-04 6.15E-05 4.14E-05 2.30E-05 U2 Plant Vent 7.12E-04 5.28E-04 2.04E-04 1.61E-04 9.76E-05 U2 East PORV 6.13E-04 3.27E-04 1.55E-04 1.01E-04 7.18E-05 Table 4.1-37 Summary of Control Room and TSC x/Q Values (sec/n 3)

Containment Plant Vent PORV Time Interval CLB ARCON96 CLB ARCON96 CLB 12 ARCON96 0-2 hrs 1.06E-3 2.17E-4 1.29E-2 7.12E-4 N/A 6.13E-4 2-8 hrs 1.06E-3 1.37E-4 1.29E-2 5.28E-4 N/A 3.27E-4 8-24 hrs 7.03E-4 6.15E-5 8.55E-3 2.04E-4 N/A 1.55E-4 1-4 days 4.45E-4 4.14E-5 5.42E-3 1.61E-4 N/A 1.01E-4 4-30 days 1.91E-4 2.30E-5 2.32E-3 9.76E-5 N/A 7.18E-5 12 The CLB does not have Control Room or TSC doses for the MSLB, SGTR, CREA, or LRA analyses.

NOC-AE-07002127 Attachment 1 Page 70 of 219 4.2 ANALYTICAL MODELS 4.2 Analytical Models The RADTRAD code (Version 3.0.3, Reference 19) was used to determine offsite doses and doses to Control Room and Technical Support Center personnel. However, the Fuel Handling Accidents used a simplified spreadsheet technique as discussed in. Section 4.4.

No credit for personal protective equipment or prophylactic drugs is taken in the analyses.

The 0-2 hr X/Q is used for the worst 2-hour doses for offsite, control room, and TSC dose analyses.,

4.2.1 Offsite Dose Model The analytical equation for determining the offsite doses is described in Section 2.3.1 of the RADTRAD documentation. The following is a summary of that discussion.

The dose to the hypothetical individual is calculated using the specified X/Qs and the amount of each nuclide released during the exposure period. The air immersion dose from each nuclide, n, in an environmental compartment is calculated as:

D~n ev= An QD CF,,,

where De = air immersion (cloudshine) dose due to nuclide n in the environment compartment (Sv)

DCFc,n = FGR 11 and 12 (References 20 and 21) air immersion (cloudshine) dose conversion factor for nuclide n as discussed in Section 1.4.3.3 of the RADTRAD documentation. (Sv m 3/Bq s)

X/Q = atmospheric relative concentration (s/mi3)

An = released activity of nuclide n (Bq).

The inhalation dose from each nuclide, n, in an environmental compartment is calculated as:

Di, =An A BR*DCFi,n where Deiv = inhalation dose commitment due to nuclide n in the environment compartment (Sv)

BR = breathing rate (in 3 / s)

DCFi,n = inhalation dose conversion factor for nuclide n as discussed in Section 1.4.3.3 of the RADTRAD documentation (Sv/Bq)

X/Q = atmospheric relative concentration (s/m 3)

SA, = released activity of nuclide n (Bq).

NOC-AE-07002127 Attachment 1 Page 71 of 219 4.2 ANALYTICAL MODELS The breathing rates used in the offsite analyses are presented in Table 4.2-1.

Table 4.2-1 Offsite Breathing Rates (m3/sec)

Time LPZ and EAB"3 0 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3.5 x 10.4 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.8 x 10-4 1 - 4 days 2 .3 x 10-4 44- 30days 2 .3 x 10-4 The TEDE is determined at the EAB for the limiting 2-hour period and at the outer boundary of the LPZ. No correction is made for depletion of the effluent plume by deposition on the ground.

4.2.2 Control Room Analytical Model To determine the dose to personnel in the control room, the RADTRAD code and the built-in control room model are used. The analytical equation for determining the control room and TSC doses is described in Section 2.3.2 of the RADTRAD documentation. The following is a summary of that discussion.

The dose to a hypothetical individual in the control room is calculated based on the time-integrated concentration in the control room compartment. The air immersion dose in the control room is:

cn = C, (t) dt (DCFCn/ GF)

Where CI(t) is the instantaneous concentration of nuclide n in the compartment. The Murphy-Campe (Reference 22) geometric factor, GF, relates the dose from an infinite cloud to the dose from a cloud of volume Vas:

1173 GF =V0.338 13 Reference 3, page 16

NOC-AE-07002127 Attachment 1 Page 72 of 219 4.2 ANALYTICAL MODELS The inhalation dose in the control room is Di R= C, (t)dt BR OF DCFi,,

GF where OF = occupancy factor.

No credit is taken for the use of personal protective equipment or prophylactic drugs in the accident analyses when calculating dose consequences to the control room operator.

The control room envelope is located at elevation 35 ft and in two heating, ventilating, and air conditioning (HVAC) rooms at elevations 10 ft and 60 ft. in the Electrical Auxiliary Building as shown in Figure 4.2-1 (from Figure 6.4-1 of the Updated Final Safety Analysis Report (UFSAR)).

The Control Room HVAC system is designed to maintain the control room envelope at a minimum of 0.125-inch water gauge (wg) positive pressure relative to the surrounding area, following postulated accidents (other than hazardous chemical/smoke releases) and/or Loss-of-Offsite Power (LOOP), by introducing makeup air equivalent to the expected exfiltration air during plant emergency conditions (Engineered Safety Features [ESF] signal and/or high radiation in outside air). The design outside makeup air is 2,000 ft3/min and drawn from a single intake on the east side of the Electrical Auxiliary Building at elevation 80 ft-0 in. Additionally, during postulated accident conditions, on detection of high radiation in the outside air or safety injection (SI) signal, outside makeup air for the control room envelope is automatically routed through makeup air units and cleanup units containing charcoal filters. The control room air is also automatically recirculated partially (i.e., 10,000 ft3/min) through control room air cleanup units containing charcoal filters. This arrangement provides cleanup of the control room air.

The control room envelope HVAC system is not connected to other areas or HVAC systems where the potential for radioactivity exists, except for sharing common air intake and exhaust with the remaining Electrical Auxiliary Building.

The Control Room HVAC model schematic is given by Figure 4.2-2. The mathematical model used to represent the system uses a single outside air intake and a filtered make-up inflow which mixes with part of the recirculating air in the Control Room Envelope. The combined recirculating air and make-up air stream is then filtered before being supplied to the air-handling unit along with the remaining recirculating air. The air handling unit supplies the conditioned air to the control room envelope. A summary of these parameters is presented in Table 4.2-2. The assumed unfiltered in-leakage into the control room envelope has been revised to 100 cfln for the AST analyses as the result of control room in-leakage testing as described below.

Unless otherwise noted, the analyses assume there is an emergency diesel failure and that only two trains of HVAC are in operation. The make-up flow rate for two trains of emergency HVAC

NOC-AE-07002127 Attachment 1 Page 73 of 219 4.2 ANALYTICAL MODELS operation is 2000 cfrn. The make-up flow is assumed to operate at +10% of design flow (2200 cfm). The flow rate for two trains of Control Room HVAC recirculation flow is 8600 cfn (430014 cfm per train). The Control Room HVAC exhaust flow rate (Label G) is 2300 cfmn. The Control Room HVAC exhaust flow rate is the sum of the make-up flow (Label A) and the unfiltered in-leakage (Label F). Also, each of the three trains of Control Room HVAC system contains 2 sets of 2-inch charcoal filters. The first 2-inch filter is the make-up filter. Filtered make-up air is then combined with recirculated air and then passes through the 2-inch recirculation filter before entering the Control Room.

In the CLB, if one train of control room HVAC is not functioning, for example due to diesel generator failure, not all of the makeup air would be filtered twice before it is introduced into the control room envelope. In the worst case, 235 cfrn of the makeup air is filtered by the makeup units, but not by the recirculation units, before it is introduced into the control room envelope. In the revised AST analyses, this assumption is not needed since credit is not taken for filtration of the make-up air.

In contrastto the CLB, the revised AST analyses assume that all makeup flow is unfiltered (e.g.,

removing the 4 inches of filtration per train, 2 inches for the makeup filters and 2 inches of the cleanup filters, for make-up air in the CLB). Only the recirculation filtration is credited. Hence, the assumed make-up air flow (Label A on Figure 4.2-2) on Table 4.2-2 is assumed to be 0 cfmn.

The 2200 cfmn make-up flow is added to the 100 cfm unfiltered in-leakage value (which includes the contribution from door pumping action from Control Room ingress and egress) and a total of 2300 cfin is assumed to directly enter the Control Room without filtration (Label F on Figure 4;2-2). No credit is taken for the use of non-ESF ventilation systems during the Design Basis Accident. In summary, Table 4.2-2 reflects the air flow with 2 trains operating while Table 4.2-3 reflects the flows used in the analyses.

The Control Room recirculation clean-up filter efficiencies are assumed to have 95% removal efficiency for elemental iodine and organic iodine and 99% removal efficiency for particulates.

The assumption of 100 cfmn unfiltered in-leakage is validated by in-leakage testing conducted in Unit 1 in March 2004 and in Unit 2 in March 2007. The testing was conducted using the tracer gas method described in ASTM E741-00 (Reference 23). The test results for Unit 1 were reported in Reference 24 in response to NRC Generic Letter 2003-01, "Control Room Habitability." The limiting train combination test results were 9.4 +/- 50 scfmn in Unit 1 and 64

+/- 8 scfm in Unit 2. Therefore, an unfiltered in-leakage assumption of 100 cfm is conservative.

The calculated control room volume is 304000 ft3 . Approximately 10% of this volume is 3

occupied by walls and equipmentl The volume used in dose analyses is 274080 ft .

14 Per plant procedures, the acceptance criteria for the surveillance testing of the make-up flow and make-up+clean-up flow is 1000 cfm +/- 10% and 6000 cfm +/- 10%, respectively. Therefore, it is acceptable to have a recirculation flow rate of4300 cfm ([6000 cfm x 0.9]- [1000 cfm x 1.1]) =5400 cfm - 1100 cfn =4300 cfm. The +/- 10% band on the flow rages is based on the acceptance criteria of TS Surveillance 4.7.7.c.3.

NOC-AE-07002127 Attachment I Page 74 of 219 4.2 ANALYTICAL MODELS Note that the Fuel Handling Accident analysis does not credit either the make-up or recirculation filters. The Control Room internal air is assumed to be in equilibrium with the air outside the Control Room HVAC intake. Therefore, the Control Room is not assumed to be pressurized during the accident, nor are any assumptions made as to the functioning of the Control Room HVAC systems.

NOC-AE-07002127 Attachment 1 Page 75 of 219 4.2 ANALYTICAL MODELS Figure 4.2-1 Control Room Envelope

NOC-AE-07002127 Attachment 1 Page 76 of 219 4.2 ANALYTICAL MODELS Figure 4.2-2: Control Room HVAC Analytical Model A

Table 4.2-2 Control Room HVAC Flow Rates (2 trains)

CLB AST Flow Path Label Flow Rate (cfrn) Flow Rate (cfm)

Make-up A 2200 2200's Clean-up (or recirc) B 11,700 10,800 Clean-up (or recirc) C 9500 8600 A/C Intake D 21,660 24,760 A/C + Clean-up Exhaust E 33,360 35,560 Unfiltered In-leakage F 10 10016 Exhaust Flow G 2210 2300 16 Set

.15 Settoto0 cfm in the analytical cfm 2200+100=2300 model.

in See the Table 4.2-3.

analytical model. See Table 4.2-3.

NOC-AE-07002127 Attachment 1 Page 77 of 219 4.2 ANALYTICAL MODELS Table 4.2-3 17 Parameters Used in Modeling the Control Room Parameter CLB AST Pressurization (makeup) flow (Label A) 2200 cfm 0 cfmn' Pressurization (makeup) 2" filter efficiencies 19 :

inorganic (elemental) 95% 0%

organic 95% 0%

particulate 99% 0%

Clean-up (recirculation) flow (Label C) 9500 cfm 8600 cfm Clean-up (recirculation) filter efficiencies ( 2, 2" filters):

inorganic (elemental) 95% 95%

organic 95% 95%

particulate 99% 99%

Free Volume 274,080 ft3 Unfiltered In-leakage (Label F) 10 cfrn 2300 cfm total (from door (including door pumping action) pumping action)

Portion of above make-up flow that bypasses Control 235 cfrni2 0 cfm Room recirculation clean-up filters (two trains)

All X/Q's Table 4.1-37 Control Room Occupancy Factors 0-24 hrs 100%

1-4 days 60%

4-30 days 40%

Breathing Rate 3.5E-4 m 3/sec 4.2.2.1 CRE Unfiltered In-leakage and Possible "Sneak" Paths The unfiltered in-leakage into the CRE is assumed to be 100 cfm for all accidents. The Control Room and the TSC are enclosed in the Electrical Auxiliary Building and the surrounding spaces are supplied by the Electrical Auxiliary Building HVAC system. The intake of the Electrical Auxiliary Building HVAC system is located just south of the Control Room/TSC HVAC intakes on the east wall of the Electrical Auxiliary Building (points D/H on Figure 4.1-13). Since the two intakes are very close, the Control Room/TSC X/Q's are used for the air entering the Electrical Auxiliary Building HVAC, and, therefore, for the unfiltered in-leakage.

17 This table is based upon the current UFSAR Table 6.4-2, Control Room Dose Analysis 18For AST, all makeup flow is assumed to be unfiltered, bypassing the 2" recirculation filters.

19For the CLB, 1765 cfm is filtered through makeup and recirculation filters; 235 cfm is filtered through makeup filters only. The effective filter efficiencies for 2000 cfm were used.

20 Only receives filtration from the 2" makeup filters

NOC-AE-07002127 Attachment I Page 78 of 219 4.2 ANALYTICAL MODELS Since the spaces surrounding the Control Room are in the Electrical Auxiliary Building, the chances for a more direct, unanalyzed, path (i.e., a "sneak" path) for airborne contaminants to enter the Control Room are minimized. The largest potential source of a "sneak" path is the Electrical Penetration area which is directly between the Control Room Envelope and the containment building (on the bottom of the Control Room Envelope, west of the Relay Room and Computer Room, as depicted in Figure 4.2-1) for the LOCA and Control Rod Ejection accidents. However, the possibility of leakage from the containment into the penetration area and finally into the Relay Room and Control Room Envelope is minimized by the presence of double doors between the Relay Room and the penetration area (partially shown on Figure 4.2-1). In addition, there is no equipment located in the penetration area that must be manipulated or observed in a post-accident scenario. Therefore, traffic through the doors would be minimal, if any. In consideration of the above, leakage from the penetration area into the Control Room Envelope is not considered credible.

4.2.3 Technical Support Center (TSC) Analytical Model To determine the dose to personnel in the TSC, RADTRAD is used and the control room node in the code is used as the TSC. The analytical model of the TSC is identical to the one discussed for the control room in Section 4.2.2, above. No credit is taken for the use of personal protective equipment 'or prophylactic drugs in the accident analyses when calculating dose consequences to TSC personnel. A description of the TSC HVAC model is given below.

It is assumed that walls and equipment occupy 25% of the TSC volume measured from exterior dimensions. The TSC volume used in radiological dose analysis is 48170 ft3. The TSC HVAC make-up flow passes through two 2-inch carbon filters in series. However, for conservatism, this analysis assumes that all makeup flow is unfiltered. A portion of the recirculation flow from the TSC passes through the carbon filters. The remainder of the recirculation flow combines with the make-up flow prior to entering the air-handling unit. The TSC HVAC model schematic is given by Figure 4.2-3.

NOC-AE-07002127 Attachment 1 Page 79 of 219 4.2 ANALYTICAL MODELS Figure 4.2-3 TSC HVAC Analytical Model A

Fan Shaft In-leakage 3

Unfiltered In-leakage Table 4.2-4 TSC HVAC Flow Rates Design Assumed Flow Rate (cfm)

Flow Flow Path Label (cfim) CLB AST Make-up A 1200 1210 12102 Clean-up (or recirc) + Make-iup B 6200 5960 5960 Clean-up (or recirc) C 5000 4750 4750 A/C Intake D 5225 5225 A/C + Clean-up E 9975 9975 Fan Shaft In-leakage F 0 5 5 Unfiltered In-leakage G 10 1022 Exhaust Flow H 1200 1225 1225 21 Set to 0 cfm in the analysis. See Table 4.2-5.

22 Set to 1210+10+5=1225 cfm in the analysis. See Table 4.2-5.

NOC-AE-07002127 Attachment 1 Page 80 of 219 4.2 ANALYTICAL MODELS TSC filter efficiencies are based on two 2-inch filters in series.

Table 4.2-5 Parameters Used in Modeling the TSC Parameter CLB AST Pressurization (makeup) flow (cfm) (Label A) 1210 603 Clean-up (recirculation) flow (cfln) (Label C) 4750 Filter efficiencies:

inorganic (elemental) 99%

organic 99%

particulate 99%

Free Volume 48,167 ft3 Unfiltered In-leakage (cfin) (Labels F & G) 15 1 1225 All X/Q's Table 4.1-37 Control Room/TSC Occupancy Factors 0-24 hrs 100%

1-4 days 60%

4-30 days 40%

Breathing Rate 3.5E-4 m3/sec The TSC HVAC make-up flow rate is 1100 cfm. The TSC HVAC make-up flow rate (Label A) operates at +10% off design (1210 cfrn). The recirculation flow rate is 5000 cfm. The recirculation flow rate (Label C of Figure 4.2-3) operates at -5% off design (4750). The TSC HVAC exhaust flow rate (Label H) is 1225 cfm. The fan shaft in-leakage (Label F) is 5 cfm.

The unfiltered in-leakage (Label G) is 10 cfm. The TSC HVAC exhaust flow rate is the sum of the make-up flow (Label A), the fan shaft in-leakage (Label F) and the unfiltered in-leakage (Label G).

The TSC HVAC system is non-safety; therefore, no single failures are assumed.

23 All AST makeup flow is assumed to be unfiltered, bypassing the 4" of filtration used for the makeup and recirculation pathways.

NOC-AE-07002127 Attachment 1 Page 81 of 219 4.2 ANALYTICAL MODELS 4.2.4 Radiological Source Terms 4.2.4.1 Dose Conversion Factors and Physical Parameters The dose conversion factors (DCF) used in the LOCA and FHA are the default RADTRAD values (Reference 19, Tables 1.4.3.3-1 and -2), with slight modifications for parents with short-lived daughters (I-135,Cs-137, Te-129m, Te-131m, Ru-103, Ru-106, Zr-97, Ce-144).

The DCFs used in the MSLB, SGTR, CREA, and LRA analyses are presented in Table 4.2-6.

The CLB DCFs are based on ICRP-30 (Reference 25). The AST DCFs for external exposure (EDE) and inhalation (CEDE) are from the Federal Guidance Report No. 12, "External Exposure to Radionuclides in Air, Water, and Soil" (Reference 21) and the Federal Guidance Report No.

11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," respectively.

Table 4.2-6 Dose Conversion Factors CLB AST EDE Thyroid Beta-Skin Whole Body (Sv-m 3/Bq- CEDE (rem/Ci) (rem-m 3/Ci-sec) (rem-m3/Ci-sec) (Sv/Bq)

Isotope sec) 1-131 1.080E+06 4.087E-02 6.734E-02 1.82E-14 8.89E-09 (3.29E+4 rem/Ci) 1-132 6.438E+03 1.617E-01 4.144E-01 1.12E-13 1.03E-10 (3.81 E+2 rem/Ci) 1-133 1.798E+05 1.032E-01 1.088E-01 2.94E-14 1.58E-09 (5.85E+3 rem/Ci) 1-134 1.066E+03 2.011E-01 4.81 0E-01 1.30E-13 3.55E-11 (1.3 1E+2 rem/Ci) 1-135 3.130E+04 1.153E-01 2.953E-01 7.98E-14 3.32E-10 (1.32E+3 rem/Ci)

Kr-83m N/A 1.547E-05 5.550E-06 1.50E-18 0 Kr-85m N/A 5.468E-02 2.768E-02 7.48E-15 0 Kr-85 N/A 4.843E-02 4.403E-04 1.19E-16 0 Kr-87 N/A 3.482E-01 1.524E-01 4.12E-14 0 Kr-88 ,N/A 1.221E-01 3.774E-01 1.02E-13 0 Kr-89 N/A 3.981E-01 3.232E-01 N/A N/A Sr-89 N/A N/A N/A 7.73E-17 1.12E-08 Xe-131m N/A 1.544E-02 1.439E-03 3.89E-16 0 Xe-133m N/A 3.227E-02 5.069E-03 1.37E-15 0 Xe-133 N/A 1.145E-02 5.772E-03 1.56E-15 0 Xe-135m N/A 3.144E-02 7.548E-02 2.04E-14 0 Xe-135 N/A 7.066E-02 4.403E-02 1.19E-14 0

NOC-AE-07002127 Attachment 1 Page 82 of 219 4.2 ANAL YTICAL MODELS Table 4.2-6 Dose Conversion Factors CLB AST EDE Thyroid . Beta-Skin Whole Body (Sv-m 3/Bq- CEDE Isotope (rem/Ci) (rem-m 3/Ci-sec) (rem-m 3/Ci-sec) sec) (Sv/Bq)

Xe-137 N/A 4.642E-01 3.026E-02 N/A N/A Xe-138 N/A 1.728E-01 2.135E,-01 5.77E-14 0 Rb-86 N/A N/A N/A 4.81E-15 1.79E-09 Rb-87 N/A N/A N/A 1.82E-18 8.74E- 10 Rb-88 N/A N/A N/A 3.36E-14 2.26E-1 1 Rb-89 N/A N/A N/A 1.06E-13 1.16E-11 Cs-134 N/A N/A N/A 7.57E-14 1.25E-08 Cs-135 N/A N/A N/A 5.65E-19 1.23E-09 Cs-136 N/A N/A N/A 1.06E-13 1.98E-09 Cs-137 N/A N/A N/A 7.74E-18 8.63E-09 Cs-138 N/A N/A N/A 1.21E-13 2.74E- 11 Ba-137m N/A N/A N/A 2.88E-14 0 The LOCA and FHA analyses use the default RADTRAD isotopic data and progeny data (Reference 19, Table 1.4.3.2-2). Table 4.2-7 presents physical data for the isotopes of interest for the MSLB, SGTR, CREA, and LRA analyses. The half life data is from Reference 26. The progeny and decay fractions are from RADTRAD (Reference 19, Table 1.4.3.2-2). Some meta-stable isotopes of xenon not contained in RADTRAD are assumed to always decay to the ground state of the same isotope.

Table 4.2-7 Isotopic Half Lifes, Parent-to-Daughter Decay Isotopes and Fractions T 1/2 Isotope (sec) Daugahter 1 Fraction 1 Daughter 2 Fraction 2 1-131 6.947E+05 Xe-131m 0.1100E-01 1-132 8.208E+03 1-133 7.488E+04 Xe-133m 0.2900E-01 Xe-133 0.9700E+/-00 1-134 3.156E+03 1-135 2.365E+04 Xe-135m 0.1 500E+00 Xe-135 0.8500E+/-00 Kr-83m 6.696E+03 Kr-85m 1.613E+04 Kr-85 0.2100E+00 Kr-85 3.386E+08 Kr-87 4.572E+03 Rb-87 0.1000E+01 Kr-88 1.022E+04 Rb-88 0.1000E+01 Kr-89 1.890E+02 Rb-89 0.1000E+01 Sr-89 4.365E+06

NOC-AE-07002127 Attachment 1 Page 83 of 219 4.2 ANALYTICAL MODELS Table 4.2-7 Isotopic Half Lifes, Parent-to-Daughter Decay Isotopes and Fractions T1/2 Isotope (see) Daughter 1 Fraction 1 Daughter 2 Fraction 2 Xe-131m 1.028E+06 Xe-133m 1.892E+05 Xe-133 0.1000E+01 Xe-133 4.530E+05 Xe-135m 9.180E+02 Xe-135 0.9940E+00 Cs-135 0.6000E-03 Xe-135 3.276E+04 Cs-135 0.1000E+01 Xe-137 2.292E+02 Xe-138 8.460E+02 Rb-86 1.610E+06 Rb-87 1.515E+18 Rb-88 1.062E+03

• .Rb-89 9.240E+02 Sr-90 0.1000E+01 Cs-134 6.51 OE+07 Cs-135 7.250E+13 Cs-136 1.140E+06 Cs-137 9.51 OE+08 Ba-137m 0.9500E+00 Cs-138 1.932E+03 Ba-137m 1.531E+02 4.2.4.2 Reactor Core Source Terms The basic source terms used in the Current Licensing Basis for the reactor core and the reactor coolant system were taken from Revision 4 of the Westinghouse RadiationAnalysis Design Manual (Reference 27). This document was based on the 1973 ORIGEN (Reference 28) computer code. Revision 5 of the RadiationAnalysis Design Manual (Reference 29), based upon ORIGEN 2.1 (Reference 30), has been used for all AST analyses.

Table 4.2-8 provides a comparison of the major parameters used to determine the source terms in the RadiationAnalysis Design Manual. The major difference in the two revisions of the analysis is the different versions of ORIGEN used. Also, the difference in the assumed reactor coolant system (RCS) cleanup flow rate (letdown rate) lowers the calculated isotopic inventory in the RCS (see Table 4.2-14). Since there is no purging of the volume control tank (VCT), the gases reach equilibrium in the VCT and RCS. Since the analyses are performed at 1% failed fuel, both sets of data bound actual plant operation. The iodine concentrations resulting from the I% failed fuel assumption are much greater than those at the Technical Specification maximum of 1 jiCi/gm.

NOC-AE-07002127 Attachment 1 Page 84 of 219 4.2 ANALYTICAL MODELS Table 4.2-8 Comparison of Revisions to the RadiationAnalysis Design Manual Parameter Rev 4/CLB Rev 5/AST ORIGEN version 0 2.1 Reactor Power 4100 MWt 4100 MWt Core Burnup 20K!40K/60K 20K/40K/60K (EOL, 3 region equilibrium core) MWD/MTU MWD/MTU Or Or 509, 1018, and 1527 EFPD 509, 1018, and 1527 EFPD Reactor Coolant Volume 13,521 ft3 13,521 ft3 -

% Failed Fuel 1% 1%

RCS Letdown Rate 100 gpm @130'F 140 gpm @130'F and 2250 psia and 2250 psia Volume Control Tank purge rate 0 cfm 0 cfm The AST values used in this analysis were derived using guidance outlined in Regulatory Guide 1.183. The ORIGEN 2.1 code was used to calculate plant-specific fission product inventories for use in the dose analyses. The assumed period of irradiation was sufficient (three-region equilibrium cycle core at end of life with the three regions having operated at 39.31 MW/MTU for 509, 1018, and 1527 EFPD, respectively) to allow the activity of dose-significant radionuclides to reach equilibrium or to reach maximum values. The reactor core inventory is presented in Table 4.2-9.

Table 4.2-9 Comparison of CLB and AST Reactor Core Sources (Ci) 24 25 Isotope CLB AST  % Difference Kr83m 1.40E+07 1.40E+07 0.0%

Kr85m 3.OOE+07 2.90E+07 -3.3%

Kr85 1 .20E+06 1.20E+06 0.0%

Kr87 5.50E+07 5.50E+07 0.0%

Kr88 7.90E+07 7.80E+07 -1.3%

Kr89 9.70E+07 9.50E+07 -2.1%

Xel31m 7.70E+05 1.10E+06 42.9%

Xe133m 3.30E+07 6.80E+06 -79.4%

Xe133 2.30E+08 2.20E+08 -4.3%

24 Reference 27 25 Reference 29

NOC-AE-07002127 Attachment I Page 85 of 219 4.2 ANALYTICAL MODELS Table 4.2-9 Comparison of CLB and AST Reactor Core Sources (Ci)

Isotope CLB 24 AST 21  % Difference Xel35m 4.60E+07 4.20E+07 -8.7%

Xe 135 6.50E+07 5.50E+07 -15.4%

Xel37 2.00E+08 1.90E+08 -5.0%

Xel38 1.90E+08 1.80E+08 -5.3%

1131 1.14E+08 1.06E+08 -7.0%

1132 1.64E+08 1.52E+08 -7.3%

1133 2.40E+08 2.20E+08 -8.3%

1134 2.60E+08 2.40E+08 -7.7%

1135 2.20E+08 2.OOE+08 -9.1%

Cs134 3.30E+07 2.20E+07 -33.3%

Cs136 9.30E+06 6.30E+06 -32.3%

Cs137 1.40E+07 1.30E+07 -7.1%

Sb129 3.70E+07 3.40E+07 -8.1%

Te129m 9.50E+06 5.OOE+06 -47.4%

Tel 29 3.50E+07 3.30E+07 -5.7%

Tel3lm 1.70E+07 1.50E+07 -11.8%

Ba137m 1.30E+07 1.20E+07 -7.7%

Ba140 2.OOE+08 1.90E+08 -5.0%

Rul03 1.80E+08 1.60E+08 -11.1%

Ru105 1.20E+08 1.1OE+08 -8.3%

Rul06 5.80E+07 5.50E+07 -5.2%

Y91 1.40E+08 1.40E+08 0.0%

Y92 1.50E+08 1.40E+08 -6.7%

Y93 1.70E+08 1.60E+08 -5.9%

Zr95 1.90E+08 1.80E+08 -5.3%

Zr97 1.90E+08 1.80E+08 -5.3%

Nb95 2.OOE+08 1.30E+08 -35.0%

La140 2.1OE+08 1.90E+08 -9.5%

La142 1.80E+08 1.70E+08 -5.6%

Pr143 1.70E+08 1.60E+08 -5.9%

Nd147 7.40E+07 7.1OE+07 -4.1%

Cel41 1.90E+08 1.80E+08 -5.3%

Ce143 1.80E+08 1.70E+08 -5.6%

Ce144 1.40E+08 1.40E+08 0.0%

NOC-AE-07002127 Attachment 1 Page 86 of 219 4.2 ANALYTICAL MODELS Table 4.2-9 Comparison of CLB and AST Reactor Core Sources (Ci)

Isotope CLB 24 AST 25  % Difference Sr89 1.10E+08 1.1OE+08 0.0%

Sr90 L.00E+07 9.70E+06 -3.0%

Sr9l 1.40E+08 1.30E+08 -7.1%

Sr92 1.50E+08 1.40E+08 -6.7%

The non-LOCA design bases analyses used the Departure from Nucleate Boiling Ratio (DNBR) as a fuel damage criterion.

4.2.4.2.1 Peak Pin Evaluation for non-LOCA Fuel Gap Inventory Footnote 11 for Regulatory Guide 1.183, Table 3, Non-LOCA Fractionof FissionProduct Inventory in Gap states that the release fractions for Table 3 are "acceptable for use with currently approved LWR fuel with a peak burnup of 62,000 MWD/MTU provided that the maximum linear heat generation rate does not exceed 6.3 kW/ft peak rod average power for burnups exceeding 54 GWD/MTU" (the "54/6.3" criteria).

Westinghouse's design code, ANC (Reference 31), was used to calculate the best estimate pin power and pin burnup for a fuel cycle. For the purpose of this evaluation, the code is used to calculate and edit the limiting relative power and the limiting pin burnup of the assembly for Unit 1 Cycles 13 and 14 Unit 2 Cycle 12. The 3 cycles evaluated are typical 18-month cycles that were designed for about 500 EFPD hot full power energy plus an additional 30 EFPD of coastdown operation.

The evaluation selected the limiting relative pin power and the limiting pin burnup of the assembly. This assessment approach is conservatively bounding as it assumes that the maximum power rod is also the maximum bumup rod of the assembly. At hot full power condition (3853 Mwth), the average linear power density of a fuel pin is 5.4 Kw/ft. Therefore, the 6.3 Kw/ft pin power limit corresponds to a "relative" pin power value of 1.167 (normalized to an average of 1.0).

The Unit 2 Cycle 12 maximum relative pin power for burnup exceeding 54 GWD/MTU is 1.055, well below the 1.167 limit. The maximum pin bumup for the cycle, including 30 EFPD of coastdown, is 58,433 MWD/MTU. The Unit 1 Cycle 13 limiting pin burnup remains below 54 GWD/MTU at the end of hot full power (cycle burnup about 19350 MWD/MTU). The maximum pin burnup slightly exceeds 54 GWD/MTU at extended coastdown (20700 MWD/MTU). However, since the limiting burnup assemblies are located on the core periphery, the relative pin power is only 0.901, well below the 1.167 limit.

NOC-AE-07002127 Attachment 1 Page 87 of 219 4.2 ANALYTICAL MODELS The Unit 1 Cycle 14 limiting pin burnup exceeds 54 GWD/MTU at the middle of the cycle (about 9,000 MWD/MTU). Assemblies having high pin burnup are located on the core periphery. The relative pin powers for these assemblies are less than 0.7. As the core depletes, eight of the in-board assemblies have maximum pin burnup exceeding 54 GWD/MTU near the end of the cycle. The maximum relative pin power for these assemblies reaches 1.042 at the end of hot full power (about 18380 MWD/MTU). The maximum pin bumup remains below 58 GWD/MTU. Continued operation with power coastdown does not show an increase in the relative pin power. The highest pin burnup for the cycle, including a power coastdown to 19500 MWD/MTU, is 60,588 MWD/MTU. The relative pin power for the assembly is 0.703. All parameters are well below the limits specified in the Regulatory Guide 1.183.

The above evaluation of STP's typical cycle designs shows that the "54/6.3" criteria is met with significant margin. The highest relative pin power for pin burnup greater than 54 GWD/MTU is 1.055, which corresponds to a linear heat rate of about 5.7 Kw/ft. STP uses a low-low core leakage design, placing high bumup fuel on the core periphery, to improve the fuel economy. As a result, the high burnup fuel assemblies typically have a low relative power. In some instances, a limited number of twice-burned fuel assemblies may be placed in inboard locations to optimize the core power peaking behavior. In this case, the assembly will be driven to have a higher power. Unit 2 Cycle 12, for instance, has a peak pin power of 1.055 for burnups greater than 54 GWD/MTU. Inboard placements of these assemblies are usually planned one cycle in advance and are evaluated during the cycle design to make sure the power peaking and pin burnup are not outside the norm. STP plans to continue to use this design approach for future core designs, therefore, it is expected that the "54/6.3" criteria will continue to be met with adequate margin.

Currently, the licensed limit for the maximum bumup of a fuel pin is 62,000 MWD/MTU.

However, the 6.3 Kw/ft pin power limit for burnup greater than 54 GWD/MTU is not currently a requirement for reload cycle design verification. To ensure this criterion is met in future cycles, the procedure used to check the adequacy of a core design has been revised to include an evaluation on the pin power/burnup of the design core.

In summary, an evaluation was performed to determine the best estimate fuel rod average burnup and power for STP's typical 18-month cycle designs. The evaluation shows that the Regulatory Guide "54/6.3" criteria for the application of Alternative Source Term are met with significant margin. The highest relative pin power for pin burnup greater than 54 GWD/MTU is 1.055, which corresponds to a linear heat rate of 5.7 Kw/ft. STP plans to continue to use a low-low leakage core design approach and it is expected that the "54/6.3" criteria will continue to be met.

NOC-AE-07002127 Attachment 1 Page 88 of 219 4.2 ANALYTICAL MODELS 4.2.4.3 Dose Equivalent 1-131 and Coolant Activity Coolant activity limits are specified in terms of dose equivalent (DE) 1-131. This is the 1-131 concentration that would provide the same dose response as the combined concentration of all iodine isotopes in the coolant. In the CLB, thyroid dose response is used as the measure of equivalency; and the actual isotopic concentrations representing an equivalent concentration of DE 1-131 for the CLB are as follows (60 pCi/gm DE 1-131 is presented as the example; other concentrations would scale proportionately):

Table 4.2-10 Isotopic Concentrations for 60 pCi/gm Representing An Equivalent Concentration of DE 1-131 in the Current Licensing Basis giCi/gm Thyroid DCF Product (thyroid rem/Ci) 1-131 46 1.08E+06 5.OE+07 1-132 52 6.44E+03 3.4E+05 1-133 72 1.80E+05 1.3E+07 1-134 10 1.07E+03 1.1E+04 1-135 40 3.13E+04 1.3E+06 Sum divided by 1-131 DCF 6.OE+01 In Table 4.2-10, the activity (second column) is multiplied by the dose conversion factor (DCF)

(third column) to obtain a product (fourth column). These are summed and the sum is divided by the 1-131 DCF to obtain the DE 1-131. The combined concentrations are equivalent in terms of dose response to 60 pCi/gm of 1-131 as intended.

It is readily evident that any number of concentration combinations of these five iodine isotopes can be equivalent to 60 gCi/gm of 1-131. However, once the ratio of each of the four isotopes I-132 through 1-135 is established relative to 1-131, then only a single set of concentrations will correspond to 60 pCi/gm of 1-131 (or any other given concentration) in terms of dose equivalency.

Independent of adoption of the AST, this submittal also adopts Reference 29 as the basis for iodine dose equivalency. Therefore, there are two parts to the change in iodine dose equivalency proposed in this amendment request: (1) the use of CEDE DCFs for consistency with the dose basis for AST and (2) the adoption of the relative iodine isotopic concentrations from Reference 29.

The 1-131 dose equivalency from Reference 29 based on thyroid dose (ICRP-30 DCFs) is shown in Table 4.2-1 IA.

NOC-AE-07002127 Attachment 1 Page 89 of 219 4.2 ANALYTICAL MODELS Table 4.2-1 IA Isotopic Concentrations Representing An Equivalent Concentration of DE I-131 Using Updated Iodine RCS Concentrations and Thyroid DCFs Thyroid DCF

[tCi/g mn (thyroid rem/Ci) Product 1-131 42.5 1.08E+06 4.59E+07 1-132 60 6.44E+03 3.86E+05 1-133 70 ,1.80E+05 1.26E+07 1-134 13 1.07E+03 1.39E+04 1-135 190 3.13E+04 5.95E+06 Sum divided by 1-131 DCF 6.OE+01 Repeating the Reference 29 calculation using CEDE DCFs based on the RADTRAD AST default file (Table 1.4.3.3-2, "Dose Conversion Factors for NUREG-1465 Nuclides" from Reference 9), the comparison in Table 4.2-1 1B is obtained.

Table 4.2-11 B Isotopic Concentrations Representing An Equivalent Concentration of DE 1-131 Using Updated Iodine RCS Concentrations and CEDE DCFs CEDE DCF gCi/ Ma (thyroid rem/Ci) Product 1-131 42.5 3.29E+04 1.40E+06 1-132 60 3.8 1E+02 2.29E+04 1-133 70 5.85E+03 4.1OE+05 1-134 13, 1.3 1E+02 1.70E+03 1-135 190 1.23E+03 2.34E+05 Sum divided by 1-131 DCF 6.3E+01 This means that defining dose equivalency based on CEDE DCFs and the individual radionuclide concentrations from Reference 29 would result in a CEDE dose response that would exceed that for 60 [tCi/gm of 1-131 by about 5%. Therefore, for analysis purposes, the isotopic concentrations in Table 4.2-12 is proposed as the dose equivalency to 1-131 considering both CEDE DCFs and the relative iodine isotopic concentrations from Reference 29.

NOC-AE-07002127 Attachment 1 Page 90 of 219 4.2 ANALYTICAL MODELS Table 4.2-12 Proposed Isotopic Concentrations Representing An Equivalent Concentration of DE 1-131 CEDE DCF jiCi/g mn (thyroid rem/Ci) Product 1-131 40.6 3.29E+04 1.34E+06 1-132 57 3.81E+02 2.17E+04 1-133 67 5.85E+03 3.92E+05 1-134 12 1.3 1E+02 1.57E+03 1-135 182 1.23E+03 2.24E+05 Sum divided by 1-131 DCF 6.OE+01 This result confirms that dose calculations performed using the specific concentrations shown above will produce the same dose result as 60 ptCi/gm of 1-131 when CEDE is the measure of dose consequence.

To make a relevant comparison for TEDE, one must assume a breathing rate. In the limit, for a case with a very high breathing rate or one in which substantial shielding reduces the external exposure dose to a very low level, the comparison would be similar to that for CEDE. Therefore, to make the TEDE comparison, the minimum breathing rate from RG 1.183 has been used; i.e.,

1.75E-4 m 3/sec. For such a case, a pseudo-DCF for TEDE can be established where the TEDE DCF is the sum of the CEDE DCF multiplied by the assumed breathing rate and the effective dose equivalent (EDE) external exposure DCF, both values being taken from the RADTRAD AST default file. Then, the calculation can be repeated once again, with the results in Table 4.2-13.

Table 4.2-13 Proposed Isotopic Concentrations Representing An Equivalent Concentration Of DE 1-131 Using a TEDE DCF 4Ci/gm TEDE DCF Product 1-131 40.6 5.82E+00 2.36E+02 1-132 57 4.81E-01 2.74E+01 1-133 67 1.13E+00 7.57E+01 1-134 12 5.04E-01 6.05E+00 1-135 182 5.11E-01 9.30E+01 Sum divided by 1-131 DCF 7.5E+01 This shows the conservatism of using the CEDE DCFs to define DE 1-131 for the purpose of making TEDE calculations. If the given coolant concentrations of 1-131 through 1-135 are used

NOC-AE-07002127 Attachment 1 Page 91 of 219 4.2 ANALYTICAL MODELS to make AST dose calculations, the calculated TEDE dose could be equivalent to as much as 75

ýtCi/gm of 1-131 but not less than 60 gCi/gm depending on shielding and breathing rate.

Therefore, the proposed definition of DE 1-131 can be used conservatively for AST dose analyses and for the AST licensing basis. If the original Reference 29 iodine isotopic concentrations are used, the results will be even more conservative (by about 5% as discussed above).

4.2.4.4 Reactor Coolant System Source Terms 4.2.4.4.1 RCS at 1% Failed Fuel The Reactor Coolant System source terms for 1% failed fuel are presented in Table 4.2-14.

Table 4.2-14 Comparison of CLB and AST Reactor Coolant Sources

@ 1% Failed Fuel (GCi/gm)

Isotope CLB 26 AST  % Difference Kr83m 3.8E-01 3.7E-01 -2.6%

Kr85m 1.6E+00 1.5E+00 -6.3%

Kr85 7.7E+00 7.6E+00 -1.3%

Kr87 L.OE+00 9.8E-01 -2.0%

Kr88 2.9E+00 2.8E+00 -3.4%

Kr89 8.4E-02 8.4E-02 0.0%

Xel3lm 1.9E+00 2.8E+00 47.4%

Xel33m 1.6E+01 4.2E+00 -73.8%

Xe133 2.4E+02 2.4E+02 0.0%

Xel35m 4.5E-01 4.OE-01 -11.1%

Xe135 8.5E+00 7.6E+00 -10.6%

Xe137 1.7E-01 1.6E-01 -5.9%

Xe138 5.9E-01 5.8E-01 -1.7%

1131 2.4E+00 1.7E+00 -29.2%

1132 2.7E+00 2.4E+00 -11.1%

1133 3.7E+00 2.8E+00 -24.3%

1134 5.5E-01 5.2E-01 -5.5%

1135 2.1E+00 7.6E+00 261.9%

Rb86 2.4E-02 1.7E-02 -29.2%

26 Reference'27 27 Reference 29

NOC-AE-07002127 Attachment I Page 92 of 219 4.2 ANALYTICAL MODELS Table 4.2-14 Comparison of CLB and AST Reactor Coolant Sources

@ 1% Failed Fuel (pCi/gm)

Isotope CLB 26 AST27  % Difference Rb88 3.6E+00 3.7E+00 2.8%

Rb89 2.4E-01 1.7E-01 -29.2%

Cs134 3.0E+00 1.4E+00 -53.3%

Cs136 3.6E+00 2.5E+00 -30.6%

Cs137 1.6E+00 1.1E+00 -31.3%

Tel29m 1.6E-02 6.3E-03 -60.6%

Te129 1.6E-02 1.0E-02 -37.5%

Tel3lm 2.3E-02 1.6E-02 -30.4%

Te132 2.6E-01 1.8E-01 -30.8%

Bal37m 1.5E+00 1.OE+00 -33.3%

Bal40 3.5E-03 2.5E-03 -28.6%

Mo99 6.6E-01 4.6E-01 -30.3%

Tc99m 6.OE-01 4.2E-01 -30.0%

Rul03 5.OE-04 3.3E-04 -34.0%

Ru106 1.6E-04 1.1E-04 -31.3%

Y91 4.7E-04 3.2E-04 -31.9%

Y92 9.2E-04 7.8E-04 -15.2%

Y93 3.1E-04 2.5E-04 -19.4%

Zr95 5.5E-04 3.8E-04 -30.9%

Nb95 5.5E-04 3.8E-04 -30.9%

Lal40 1.1E-03 7.OE-04 -36.4%

Pr143 5.1E-04 3.6E-04 -29.4%

Ce143 4.2E-04 3.1E-04 -26.2%

Ce144 4.3E-04 2.8E-04 -34.9%

Sr89 3.3E-03 2.4E-03 -27.3%

Sr90 1.8E-04 1.2E-04 -33.3%

Sr9l 6.9E-03 3.8E-03 -44.9%

Sr92 1.1E-03 1.OE-03 -9.1%

The CLB used only the iodine, krypton, and xenon isotopes. The AST analyses use these and the cesium and rubidium isotopes (unless otherwise noted).

NOC-AE-07002127 Attachment 1 Page 93 of 219 4.2 ANALYTICAL MODELS 4.2.4.4.2 RCS lodines at Normal Tech Spec Limit of 1 FtCi/gm Since the iodine concentrations at 1% failed fuel bound the concentrations for the normal Technical Specification limit of 1 pCi/gm DE 1-131, the concentrations corresponding to the 1%

failed fuel condition are used instead of the 1 ptCi/gm DE 1-131 concentrations in the revised analyses.

4.2.4.4.3 RCS lodines at Spiking Tech Spec Limit of 60 pCi/gm The initial iodine concentration in the reactor coolant is based on 60 [iCi/gm DE 1-131.

Equation 1 shows the formulation for calculating DE 1-131.

DCF132 DCF1 33 DCF1 34 DCF1 35 X131 + X132 x +t-133 X X134 X "33" - 1 +135 X= 60 (Eq 1)

DCF131 DCF131 DCF131 DCF131 where, X131 = concentration of 1-131 X132 = concentration of 1-132 X133 = concentration of 1-133 X134 = concentration of 1-134 X135 = concentration. of 1-135 DCF 131 = 1-131 dose conversion factor DCF 132 = 1-132 dose conversion factor DCF 133 = 1-133 dose conversion factor DCF 134 = 1-134 dose conversion factor DCF 135 = 1-135 dose conversion factor The relative abundance of each isotope in the RCS is used in conjunction with Equation 1 to solve for the five concentrations. The concentration of each isotope in the RCS, based on 1%

failed fuel, is presented in Table 4.2-14. The dose conversion factors are also included in Table 4.2-6. These dose conversion factors are the thyroid conversions from Reference 20.

Table 4.2-15 shows the calculation for the Reactor Coolant System (RCS) iodine concentration,

• basedon Thyroid DCFs, for 1% failed fuel.

NOC-AE-07002127 Attachment 1 Page 94 of 219 4.2 ANAL YTICAL MODELS Table 4.2-15 RCS Iodine Concentrations for 1% Failed Fuel 28 29 CLB AST Isotope (GCi/gm) (YtCi/gm)  % Difference I-131 2.4 1.7 -29.2%

1-132 2.7 2.4 -11.1%

1-133 3.7 2.8 -24.3%

1-134 0.55 0.52 -5.5%

1-135 2.1 7.6 261.9%

Table 4.2-16 RCS Iodine Concentrations and DCFs 31 Concentration30 Thyroid DCF Isotope (pC/gmn) (Sv/Bq) 1-131 1.7 2.92E-7 1-132 2.4 1.74E-9 1-133 2.8 4.86E-8 1-134 0.52 2.88E-10 1-135 7.6 8.46E-9 The following relationships are based on the concentrations in Table 4.2-16.

32 = (2"Y1.7)4> 433 = (2"81.7)8>31 = (0.52y7 Z4 35 (7"6/.7)63>

The relationships above are substituted in Equation 1 and this equation is solved for X131.

A summary of the RCS iodine concentrations is provided in Table 4.2-17.

28 Reference 27 29 Reference 29 30 Reference 29, page 5.34 31 Reference 20, page 136

NOC-AE-07002 127 Attachment 1 Page 95 of 219 4.2 ANALYTICAL MODELS Table 4.2-17 RCS Iodine Concentrations for a Pre-existing Iodine Spike to 60 pCi/grn CLB AST Isotope (GCi/gm) ([tCi/grn)  % Difference 1-131 46 42.5 -7.6%

1-132 52 60.0 15.4%

1-133 72 70.0 -2.8%

1-134 10 13.0 30.0%

1-135 40 190.0 375.0%

4.2.4.4.4 RCS Cs and Rubidium Concentrations The RCS initial Cs and.Rb concentrations are given in Table 4.2-14. However, a preexisting iodine spike is conservatively assumed to cause an increase in Cs and Rb activities, along with the increase in iodine concentrations. Table 4.2-18 shows the total activities from a pre-accident spike.

Table 4.2-18 Total RCS Cs and Rb Activity for a Pre-Accident Iodine Spike (Ci)

Isotope CLB AST Rb-86 N/A 1.36E+2 Rb-88 N/A 2.95E+4 Rb-89 N/A 1.34E+3 Cs-134 N/A 1.12E+4 Cs-136 N/A 1.99E+4 Cs-137 N/A 8.77E+3 For cases involving an accident-induced spike, the activities are shown accident-dependant and provided in the respective accident discussion.

NOC-AE-07002127 Attachment 1 Page 96 of 219 4.2 ANALYTICAL MODELS 4.2.4.5 Secondary System Source Terms 4.2.4.5.1 Secondary System Iodine Concentrations The initial iodine concentration in the secondary systems is based on the Technical Specification limit of 0.10 pCi/gm DE 1-131. Equation 2 shows the formulation for calculating DE 1-131.

DCF13 DCF 1 33 DCF3 4 DCF13 (

X 13 1 +- 132 X + X133 Y + z 13 4 3x + z 13 5 x - 0.10 (Eq 2)

DCF131C F131 DCY131 DCF1 31 where, X131 = concentration of 1-131 X132 = concentration of 1-132 X133 = concentration of 1-133 X134 = concentration of 1-134 X135 = concentration of 1-135 DCF 131 = 1-131 dose conversion factor DCF 132 = 1-132 dose conversion factor DCF 133 = 1-133 dose conversion factor DCF 134 = 1-134 dose conversion factor DCF 135 = 1-135 dose conversion factor The relative abundance of each isotope in the RCS is used in conjunction with Equation 2 to solve for the five concentrations. The concentration of each isotope in the RCS, based on 1%

failed fuel, is presented in Table 4.2-16. The dose conversion factors are also included in Table 4.2-16.

The following relationships are based on the concentrations in Table 4.8-19.

=(2"4/.7)Z131 (12 X133 = (2".7)8131 X4= (0521 .7 1 31 43 7.61.7)53 (35 The relationships above are substituted in Equation 2 and this equation is solved for X131.

A summary of the secondary iodine concentrations is provided in Table 4.2-19.

NOC-AE-07002127 Attachment 1 Page 97 of 219 4.2 ANALYTICAL MODELS Table 4.2-19 Secondary Iodine Concentrations at 0.1 pCi/grn (4Ci/grn)

Isotope CLB AST  % Difference 1-131 7.5E-02 7.08E-02 -5.6%

1-132 8.8E-02 1.OOE-01 13.6%

1-133 1.2E-01 1.17E-01 -2.5%

1-134 1.8E-02 2.17E-02 20.6%

1-135 6.6E-02 3.17E-01 380.3%

The large increase in 1-135 is attributable to the change in relative DCFs from the CLB to the AST/TEDE analysis.

4.2.4.5.2 Secondary System Noble Gas Concentrations The noble gas concentrations and the organic iodine concentration are determined as a function of the primary-to-secondary leak rate and the steam flow rate (1.574E+07 lbn/hr). The RCS concentrations are taken from Reference 29. The secondary concentrations are calculated using the equation below.

Secondanry Concentration RCS Concentrationx (Primary- to - Secondary Leakrate)

Steam Flow Rate The initial RCS and secondary activities are presented in Table 4.2-20. The RCS mass used for calculating the activities is 2.658E+8 gin. The secondary mass is 659,412 lbm (2.991E+8 gm).

This results in the secondary side concentration of a nuclide being a factor of 3.18E-5 that of the primary side concentration.

NOC-AE-07002127 Attachment 1 Page 98 of 219 4.2 ANALYTICAL MODELS Table 4.2-20 Initial RCS (@60 jtCi/gm) and Secondary Concentrations (@ 0.1 gCi/gm DEI)

(Noble Gases based on 1% Failed Fuel) 32 33 34 RCS Secondary RCS Secondary Isotope (GCi/gm) (pCi/gm) (Ci) (Ci) 1-131 4.25E+01 7.08E-02 1.1E+04 2.1E+01 1-132 6.OOE+01 1.OOE-01 1.6E+04 3.OE+01 1-133 7.OOE+01 1.17E-01 1.9E+04 3.5E+01 1-134 1.30E+01 2.17E-02 3.5E+03 6.5E+00 1-135 1.90E+02 3.17E-01 5.1 E+04 9.5E+01 Kr-83m 3.7E-01 1.2E-05 9.8E+01 3.6E-03 Kr-85m 1.5E +00 4.8E-05 4.OE+02 1.4E-02 Kr-85 7.6E+00 2.4E-04 2.OE+03 7.2E-02 Kr-87 9.8E-01 3.1E-05 2.6E+02 9.3E-03 Kr-88 2.8E+00 8.9E-05 7.4E+02 2.7E-02 Kr-89 8.4E-02 2.7E-06 2.2E+01 8.1E-04 Rb-86 1.7E-02 5.4E-07 4.5E+00 1.6E-04 Rb-88 3.7E+00 1.2E-04 9.8E+02 3.6E-02 Rb-89 1.7E-01 5.4E-06 4.5E+01 1.6E-03 Xe-131m 2.8E+00 8.9E-05 7.4E+02 2.7E-02 Xe-133m 4.2E+00 1.3E-04 1.1E+03 3.9E-02 Xe-133 2.4E+02 7.6E-03 6.4E+04 2.3E+00 Xe-135m 4.OE-01 1.3E-05 1.1E+02 3.9E-03 Xe-135 7.6E+00 2.4E-04 2.OE+03 7.2E-02 Xe-137 1.6E-01 5.1E-06 4.3E+01 1.5E-03 Xe-138 5.8E-01 1.8E-05 1.5E+02 5.4E-03 Cs-134 1.4E+00 4.5E-05 3.7E+02 1.3E-02 Cs-136 2.5E+00 8.0E-05 6.6E+02 2.4E-02 Cs-137 1.1E+00 3.5E-05 2.9E+02 1.OE-02 Cs-138 8.9E-01 2.8E-05 2.4E+02 8.4E-03 32 Table 4.2-17 for iodine data and Reference 29 for balance of data.

33 Table 4.2-19 for iodine data, balance of data determined from the previous equation.

34 Last digit is subject to round-off changes in individual analyses.

NOC-AE-07002127 Attachment 1 Page 99 of 219 4.2 ANALYTICAL MODELS 4.2.5 Iodine Species Released from Steam Generators For the applicable accidents, the release of iodines from the fuel (and RCS) is modeled in accordance with Regulatory Guide 1.183 (Appendix EA: 4.85% elemental, 0.15% organic, and 95% particulate. Appendix E also states that the iodine release from the SG should be 97%

elemental and 3% organic. This is a result of not releasing the particulates that comprise 95% of the RCS flow mixing into the bulk SG water. However, Section 5.5.4 allows for a partition factor of 100 for iodines and states that "[t]he retention of particulate radionuclides in the steam generators is limited by the moisture carryover for the steam generators." The contradiction is that in Appendix E, Section 4, the particulates are seemingly not released and in Section 5.5.4 there is some guidance on handling particulates.

The STP analysis, therefore, make two assumptions:

1. Organic iodines are released without the reduction of 100 afforded by the partition factor granted in Appendix E, Section 5.5.4; and

2. Release of iodine particulates will be modeled, in seeming contradiction to Appendix E, Section 4, but using the partition factor of 100.

Therefore, the 4.85/0.015/95 split from the RCS becomes 4.85 / 0.15 / 95 100 1 100 when the partition factors are applied. The resulting split is then 0.485/0.15/0.95 among the iodine species. Renormalizing, the fractions are 4.2% elemental, 13.1% organic, and 82.7%

particulate.

Based on the above, the STP analyses use an elemental/organic/particulate species split of 4.2%/13.1%/82.7% in lieu of the Regulatory Guide 1.183 split of 97%/3%/0%. Note that the number of curies of iodines released are greater than that required by Regulatory Guide 1.183 (particulates are released and no partition factor is used to reduce the amount of organics released).

NOC-AE-07002127 Attachment 1 Page 100 of 219 4.3 LOSS OF COOLANT ACCIDENT 4.3 Loss of Coolant Accident Radiological Assessment 4.3.1 Methodology Overview The LOCA is modeled as a release of nuclides from the reactor core into the containment building. Subsequent releases to the environment are as follows:

" Leakage through the containment walls, at the allowed Technical Specification leakage rate of 0.3% for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and one half that value after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

• The (pre-clad rupture) activity in the reactor coolant system through the containment supplemental purge system, terminating when the supplemental purge system isolation valves close (automatically upon receipt of the safety injection signal)

Leakage via Engineered Safety Features (ESF) components in the Fuel Handling Building, at an assumed rate of 8280 cc/hr (double the allowed leakage rate of 4140 cc/hr).

Credit for containment spray is taken to reduce the amount of radionuclides available for leakage from the containment.

The radiological source term characteristics and release timing are based on the Alternative Source Term (AST) methodology in Regulatory Guide 1.183 and from NUREG-1465.

Atmospheric dispersion factors from Section 4.1, above, are used in this analysis.

Doses to the public at the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ) and to operators in the Control Room and the Technical Support Center (TSC) are determined.

4.3.2 Radiological Source Term For conservatism, the LOCA core source terms are those associated with a DBA power level of 4100 MWth compared to the licensed power level of 3853 MWth with a 0.6% measurement uncertainty.

The AST values used in this analysis were derived using guidance outlined in Regulatory Guide 1.183. The ORIGEN 2.1 code was used to calculate plant-specific fission product inventories for use in the DBA LOCA dose analyses. The assumed period of irradiation was sufficient (three-region equilibrium cycle core at end of life with the three regions having operated at 39.31 MW/MTU for 509, 1018, and 1527 EFPD, respectively) to allow the activity of dose-significant radionuclides to reach equilibrium or to reach maximum values. Certain radionuclides appearing in the default list of radionuclides for the RADTRAD 3.03 computer code but not appearing in the summary of the ORIGEN analysis were taken from the PWR default .NIF file for RADTRAD. These include Ba139, Lal41, and Np239 (used as-is from the PWR default .NIF file) and Am241, Cm242, Cm244, Pu238, Pu239, Pu240, and Pu241 (used with activities

NOC-AE-07002127 Attachment 1 Page 101 of 219 4.3 LOSS OF COOLANT ACCIDENT increased by a factor of three for conservatism because of their half-lives being greater than 100 days).

In addition to the radionuclides appearing in the RADTRAD list, Kr83m, Xel 3 lm, Xel 33m, and Xe135m were added for dose analysis purposes based on their inclusion in TID-14844. Xe138 was also added. Co58 and Co60 were deleted from the list because only 63 radionuclides can be used. A study indicated that omitting Co58 and Co60 decreased the control room dose by about 0.01 percent while adding the noble gas isotopes increased the control room dose by about 0.1 percent.

Fission product activities were calculated for immediately after shutdown and decayed for the required times. The shutdown values are shown in Table 4.3-1 (values are from Table 4.2-9, except they are expressed in terms of Ci/MWt for RADTRAD). The CLB analyses assumed 100% of the noble gases, 50% of the iodines, and 1% of the core solids were released from the core, per TID-14844. For the CLB offsite, TSC, and Control Room doses, only the iodines and noble gases were considered.

Table 4.3-1 LOCA: Reactor Core Fission Product Inventory @ t=0 (Ci/MWt) 35 36 Isotope CLB (TID) AST  % Difference Kr83m 3.41E+03 3.41E+03 -0.1%

Kr85m 7.32E+03 7.07E+03 -3.4%

Kr85 2.93E+02 2.93E+02 0.1%

Kr87 1.34E+04 1.34E+04 -0.1%

Kr88 1.93E+04 1.90E+04 -1.4%

Kr89 2.37E+04 2.32E+04 -1.9%

Xel31m 1.88E+02 2.68E+02 42.7%

Xe133m 8.05E+03 1.66E+03 -79.4%

Xe133 5.61E+04 5.37E+04 -4.3%

Xel35m 1.12E+04 1.02E+04 -9.1%

Xe135 1.59E+04 1.34E+04 -15.5%

Xe137 4.88E+04 4. 63E+04 -5.1%

Xe138 4.63E+04 4.39E+04 -5.3%

1131 2.78E+04 2.59E+04 -6.9%

1132 4.00E+04 3.7 1E+04 -7.3%

1133 5.85E+04 5.37E+04 -8.3%

1134 6.34E+04 5.85E+04 -7.8%

35The three isotopes in bold italics were only used in the STARDOSE (Reference 34) confirmatory analyses.

36 Derived from Table 5-9, Reference 27

NOC-AE-07002127 Attachment 1 Page 102 of 219 4.3 LOSS OF COOLANT ACCIDENT Table 4.3-1 LOCA: Reactor Core Fission Product Inventory @ t=0 (Ci/MWt) 35 Isotope CLB (TID)3 6 AST  % Difference 1135 5.37E+04 4.88E+04 -9.1%

Rb86 9.92E+01 Cs134 8.05E+03 5.37E+03 -33.3%

Cs136 2.27E+03 1.54E+03 -32.1%

Cs137 3.41E+03 3.17E+03 -7.2%

Sb127 3.05E+03 Sb129 9.02E+03 8.29E+03 -8.1%

Tel27m 4.32E+02 Tel27 3.05E+03 Tel29m 2.32E+03 1.22E+03 -47.3%

Tel29 8.54E+03 8.05E+03 -5.7%

Tel3lm 4.15E+03 3.66E+03 -11.7%

Tel32 3.90E+04 3.82E+04 -2.1%

Ba137m 3.17E+03 2. 93E+03 -7.6%

Ba139 4.98E+04 Ba140 4.88E+04 4.63E+04 -5.1%

Mo99 5.12E+04 4.83E+04 -5.7%

Tc99m 4.39E+04 4.07E+04 -7.3%

Ru103 4.39E+04 3.90E+04 -11.2%

Ru105 2.93E+04 2.68E+04 -8.4%

Ru106 1.41E+04 1-.34E+04 -5.3%

RhI05 3.05E+04 Y90 3.56E+03 Y91 3.41E+04 3.41 E+04 -0.1%

Y92 3.66E+04 3.41E+04 -6.8%

Y93 4.15E+04 3.90E+04 -5.9%

Zr95 4.63E+04 4.39E+04 -5.3%

Zr97 4.63E+04 4.39E+04 -5.3%

Nb95 4.88E+04 4.32E+04 -11.4%

Lal40 5.12E+04 4.63E+04 -9.6%

Lal41 4.62E+04 La142 4.39E+04 4.15E+04 -5.5%

Pr143 4.15E+04 3.90E+04 -5.9%

Nd147 1.80E+04 1.73E+04 -4.1%

Am241 2.75E+00 Cm242 1.05E+03 Cm244 6.17E+01 Ce141 4.63E+04 4.39E+04 -5.3%

NOC-AE-07002127 Attachment 1 Page 103 of 219 4.3 LOSS OF COOLANT ACCIDENT Table 4.3-1 LOCA: Reactor Core Fission Product Inventory @ t=0 (Ci/MWt) 35 Isotope CLB (TID)36 AST  % Difference Ce143 4.39E+04 4.15E+04 -5.5%

Ce144 3.41 E+04 3.41 E+04 -0.1%

Np239 5.12E+05 Pu238 8.71E+01 Pu239 1.96E+01 Pu240 2.48E+01 Pu241 4.17E+03 Sr89 2.68E+04 2.68E+04 -0.1%

Sr90 2.44E+03 2.37E+03 -2.8%

Sr91 3.41E+04 3.17E+04 -7.2%

Sr92 3.66E+04 3.41E+04 -6.8%

4.3.3 Radiological Releases 4.3.3.1 Radiological Releases from the Containment Activity released to the containment is apportioned to the sprayed and unsprayed regions according to volume, 0.8 to the sprayed region and 0.2 to the unsprayed region based on the relative volumes.

Containment spray removal coefficients continue to be based on Standard Review Plan 6.5.2, "Containment Spray as a Fission Product Cleanup System," Revision 2, December 1998, with "particulate" removal coefficients applied to "aerosols." Spray timing is adjusted slightly to reflect AST-caused differences in time to reach decontamination factor (DF) credit limits.

The assumed containment leak rate directly to the environment is the same as the CLB. For the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the leak rate is assumed to be at the Containment Leakage Testing Program (Technical Specification 6.8.3.j) limit of 0.30% per day, while for the remainder of the 30-day period the leak rate is assumed to be 0.15% per day.

Primary containment leakage to the environment is modeled as a diffuse area source in conformance with Regulatory Guide 1.194.

Containment leakage through electrical penetrations into the electrical penetration area is very limited (0.025 cfm out of 7.02 cfm containment leakage). It is held up in the electrical penetration area as a source of gamma shine dose to the Control Room. A discussion of releases

NOC-AE-07002127 Attachment 1 Page 104 of 219 4.3 LOSS OF COOLANTACCIDENT into the electrical penetration area as a source for a "sneak path" of airborne contaminants into the Control Room is provided in Section 4.2.2.1.

4.3.3.1.1 Release from the Containment Supplemental Purge Subsystem Containment leakage via open supplemental purge lines occurs for the first 23 seconds of the onset of the accident (following the Containment Pressure-High I signal and including valve closing time, Standby Diesel Generator startup time and signal and sequencer delays). The assumed volumetric flow rate is found in Table 4.3-11. This leakage is released to the environment via the plant vent.

During this time period, fuel failure has not occurred (see Table 4.3-10). This release consists of reactor coolant blowing down into the containment. The flow rate out of the supplemental purge line is assumed to be at maximum choke flow. The flow is doubled to account for flow in both the intake and exhaust lines. The purge system exhausts via the Plant Vent.

The reactor coolant concentrations are based on 1% failed fuel, which is greater than the values corresponding to the 1 RCi/gm DE 1-131 Technical Specification limit. Accordingly, these values also bound the Regulatory Guide 1.183, Appendix A, Section 3.8, position that iodine concentrations corresponding to 1.0 [tCi/gm DE 1-131 should be used.

'The CLB uses the reactor coolant concentrations for xenons and kryptons for 1% failed fuel (Table 4.2-14). A pre-existing iodine spike to 60 [tCi/gm DE 1-131 was modeled by using a value of 60 pCi/gm for 1-13 1.

4.3.3.1.2 Containment Sump pH and Iodine Re-evolution An evaluation of containment sump pH was conducted to ensure that the particulate iodine deposited into the containment water during the DBA LOCA does not re-evolve beyond the amount recognized in the DBA LOCA analysis. The objective of the analysis was to determine the transient containment sump pH so that the removal of elemental and particulate iodine (cesium iodide - CsI) from the containment atmosphere in the course of the DBA LOCA would not be overstated. The analysis credits the pH buffering effect of trisodium phosphate (TSP) stored in the containment sump.

4.3.3.1.2.1 Determination of Sump pH The calculation methodology for containment sump pH control is based on the approach outlined in NUREG-1465 and NUREG/CR-5950, (Reference 35). Specifically, credit is taken for TSP dissolution in the containment water as a result of released reactor coolant and injected spray water coming in contact with the stored TSP in the lower elevation of containment.

NOC-AE-07002127 Attachment 1 Page 105 of 219 4.3 LOSS OF COOLANT ACCIDENT The pH of the containment sump water was then calculated using the STARpH code (Reference 36). The STARpH computer code is used for determining the pH of the containment sump (PWR) or.. suppression pool (BWR). It has been used in the following AST applications that have received a satisfactory NRC Safety Evaluation:

• Perry

• Hope Creek

• Browns Ferry

• Vermont Yankee

• Waterford-3 The amount of cable insulation in containment was determined by performing a survey of design documents and determining the total volume of cable insulation and jacket materials. As would be expected, STP has a larger mass of cable insulation than most plants. STP has three safety trains instead of the traditional two. Also, the Residual Heat Removal System (RHR) is located inside the containment. Additionally, the containment is relatively large at 158 feet in diameter.

Cable data is presented in Table 4.3-2.

The design inputs were conservatively established to maximize the post-LOCA production of acids and to minimize the post-LOCA production and/or addition of bases.

In calculating the sump pH, the three major contributors to strong acid production are considered: boric acid from the reactor coolant system, the accumulators, and the refueling water storage tank (RWST); nitric acid from radiolysis of water; and, hydrochloric acid from radiolysis of chloride-bearing cable jacket/insulation. Production of organic acid from coatings is also evaluated. For South Texas, this contribution was found to be negligible.

Major assumptions used in the sump pH analysis are:

1. Per the Technical Specifications, the containment contains a minimum of 11,500 lbm of trisodium phosphate (TSP). Trisodium phosphate is stored in baskets located on the containment floor where they would be submerged in the event of a LOCA. During each refueling outage, a surveillance is performed to verify that the six trisodium phosphate storage baskets are in place, have maintained their integrity, and are filled with trisodium phosphate such that the level is within the specified range.

2. For cables without specific dimension data, the fraction of the cable cross section that is insulation is assumed to be 0.6 and all the insulation and jacket material is assumed to be Hypalon.

3. The as-built thickness of cable insulation is assumed to be a maximum of 10% larger and the jacket material is assumed to be a maximum of 25% larger than the design specification value.

4. Cable insulation quantities are increased by 5% to bound future modifications that add cable to the containment building.

NOC-AE-07002127 Attachment 1 Page 106 of 219 4.3 LOSS OF COOLANT ACCIDENT

5. The fraction of the aerosol source term in the sump is 0.9.

Spray removal of activity will wash a large fraction of the activity from the atmosphere into the containment sump. Most of the aerosol activity will be released during the 1.3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> in-vessel release phase (Regulatory Guide 1.183), giving a maximum release rate of 1/1.3= 0.77 per hour. With a spray removal rate of 6.9 per hour in the sprayed region and a total containment volume about 1.25 times greater than the sprayed region, the effective lambda is about 6.9/1.25=5.5 per hour. The maximum equilibrium fraction of aerosol airborne during the release phase would be about 0.77/5.5=0.14. Beyond the end of the release phase, this fraction would rapidly decrease, most likely approaching zero. Even though some of the release would effectively remain airborne as transient spray droplets and wetted surfaces, greater than 0.9 would be expected to be waterborne, and it is, therefore, conservative to use this fraction.

6. Organic acid from radiolysis of organic materials dissolved from the containment surface coatings in contact with the pool can be neglected.

The [H+] from production of organic acid in the containment sump is expected to be a small fraction of the total [H+] from nitric and hydrochloric acid calculated to be produced from the radiolysis of water and cables. The bases for this assertion are:

a) The [H+] from organic acid produced in the RTF experiments with painted surfaces varies from 4.2E-07 moles/liter for Epoxy and Polyurethane paints to 1.7E-05 mol/L for vinyl paint with paints cured 3 months (Reference 37), whereas the total [H+] calculated from the production of nitric and hydrochloric acids is 1.04E-03 mol/L; b) The dissolution of organics from paints (the controlling mechanism for the radiolytic production of organic acid) decreases with the age of the paint (reduction factor of approximately 4 from 10 to 100 days and an additional factor of 2 from 100 days to 1000 days) (Reference 37), whereas the STP containment surfaces were originally coated with organic materials prior to reactor startup in 1988 (Unit 1) and 1989 (Unit 2) (over 6000 days of aging) and only limited touchup has been done during outages since that time; and, c) The painted surfaces in contact with the containment sump are coated with epoxy paints.

Thus, the [H+] from organic acids will be 4.2E-07/1.04E-03, or 0.039% of the [H+]

produced from nitric and hydrochloric acids.

7. No credit is taken for basic alkali metal compounds that result from fission products co-released with the iodine.

8. Cesium compounds are not credited in the long-term pH analyses and the determination of the final (i.e., 30 day) pH value.

9. The favorable impact of fission product chemistry on sump pH is largely ignored.

Although some HI may be formed, the amount of HI would be overwhelmed by the favorable impact of Cs compounds, in particular CsOH.

NOC-AE-07002127 Attachment 1 Page 107 of 219 4.3 LOSS OF COOLANT ACCIDENT

10. For conservatism, 10% of non-noble gas activity is assumed to remain airborne for the full 30 days, even in the presence of sprays. All of the noble gas activity is assumed to remain airborne. This increases the amount of radiation exposure to cables.

11. For HC1 formation, a factor of two reduction is credited for beta shielding of cable in trays. This is conservative since cables are usually layered in trays, providing a significant amount of self shielding. A factor of ten is credited for the 16% of cable that is estimated to be in conduit.

The inputs for the pH evaluation are presented in Table 4.3-2.

Table 4.3-2 Containment Sump pH Control Inputs Current Licensing Input/Assumption Basis AST Mass of water in post-accident sump 2.44E+09 grams 2.44E+09 grams Boron concentration (as boric acid) in sump 3060 ppm 3060 ppm Mass of TSP dodecahydrate 11,500 lbm 11,500 Ibm Initial sump pH after TSP dissolution 7.01 7.01 Containment volume N/A 3.38E+06 ft3 Volume of Hypalon in containment N/A 1.76E+07 cc Mass of Hypalon in containment N/A 2.73E+07 grams Density of Hypalon N/A 1.55 gm/cc Representative thickness of cable jacket N/A 64.5 mils Representative cable outside diameter N/A 0.65 inches Percent of cable in conduit N/A 16%

Conduit outside diameter N/A 1.94 inches Conduit thickness N/A 0.153 inches Fraction of exposed cables in trays N/A 100%

The STARpH code was used to determine the amount of [HNO 3] in the sump water generated by radiolysis of water. Organic acids from the containment surfaces coated with organic materials was neglected. A water density of 1.0 gm/ml was used to minimize the volume and maximize the HNO 3 concentration. The cumulative amount of HNO 3 in the containment sump as a function of time is provided in Table 4.3-3.

NOC-AE-07002127 Attachment 1 Page 108 of 219 4.3 LOSS OF COOLANTACCIDENT The STARpH code was used to determine the amount of [HCI] in the sump water generated by radiolysis of cable insulation. STARpH uses the methodology described in Appendix B of NUREG/CR-5950. Using the formulation in Appendix B of NUREG/CR-5950, the rate of HC1 formulation is given by R - R(H + RpH where R the total rate of HC1 generation rate (gm-mols/sec)

RyH the HC1 generation rate due to y radiolysis (gm-mols/sec)

RDH the HCl generation rate due to P3 radiolysis (gm-mols/sec)

RyH'is determined to be 3.OOE-15 gm-mols/sec and ROH is determined to be 1.75E-15 gm.-

mols/sec. After correcting for beta shielding, RPH is 1.52E-15 gm-mols/sec. The cumulative amount of HC1 in the containment sump as a function of time is provided in Table 4.3-3.

To determine the sump pH the mass of boron and TSP are also used. The sump temperature at 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> is 172°F (78'C). After which, it will decrease with long-term cooling of the sump.

Based on curve fitting and extrapolation of dissociation constant data as a function of temperature (0 to 50'C) from Dean (Reference 38), at 78°C the dissociation constant (KA) for boric acid is 9.04E-10 and that for P0 4 (KA2) is 5.01E-08. These parameters were used in STARpH to yield the pH values as a function of time presented in Table 4.3-3.

The initial effects on post-accident containment sump pH is from rapid fission product transport and the formation of cesium compounds, which results in increasing the containment sump pH.

The buffering effect of TSP within a few hours is sufficient to offset the effects of these acids that are transported to the sump and maintain containment sump pH at or above 7.0 for the first day.

The impact of HCl formation from cable radiolysis is about four times greater than the impact of nitric acid formation from water radiolysis.

As radiolytic production of nitric acid and hydrochloric acid proceeds and these acids are transported to the pool over the first days of the event, the pH becomes more acidic. After the first day, the containment sump pH will begin to decrease, reaching 6.8 by the end of the 30-day duration of the radiological consequence analysis for the DBA LOCA, and the impact of that decrease has been reflected in the Control Room and offsite doses.

Although the results of this analysis indicate that the sump pH drops slightly below 7.0, in reality there should be little impact on the actual iodine re-evolution due to the conservatisms in the analysis:

• Conservative estimates on cable dimensions and materials were made to increase the cable insulation mass and its effect on sump pH;

NOC-AE-07002127 Attachment 1 Page 109 of 219 4.3 LOSS OF COOLANTACCIDENT

" Cesium compounds are not credited in the long-term pH analyses and the determination of the final (i.e., 30 day) pH value;

" No credit is taken for basic alkali metal compounds that result from fission products co-released with the iodine;

• Conservative assumptions were made to retain 10% of non-noble gas activity as airborne activity for the full 30 days, even in the presence of sprays, and all of the noble gas activity is assumed to remain airborne (increasing the amount of radiation exposure to cables); and

• Conservative assumptions were made concerning the vulnerability of cables to beta radiation.

The above conservatisms are consistent with conservatisms in other Polestar sump pH analyses.

In addition, further conservatisms are incorporated into the determination of iodine re-evolution, discussed in Section 4.3.3.1.2.2.

Table 4.3-3 Sump Concentrations and pH as a Function of Time End of Time Interval [HNO 3] [HCI] [H+] pH 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 8.19E-06 2.70E-05 3.52E-05 7.0 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.13E-05 4.42E-05 5.55E-05 7.0 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 1.77E-05 8.OOE-05 9.77E-05 7.0 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 2.82E-05 1.29E-04 1.58E-04 7.0 1 day 4.21E-05 1.84E-04 2.27E-04 7.0 3 days 8.13E-05 3.34E-04 4.16E-04 6.9 10 days 1.53E-04 6.1OE-04 7.64E-04 6.9 20 days 1.99E-04 7.48E-04 9.47E-04 6.9 30 days 2.29E-04 8.12E-04 1.04E-03 6.8 4.3.3.1.2.2 Iodine Re-evolution The STP DBA LOCA analysis assumes iodine removal from the containment atmosphere by both containment sprays and natural diffusion to walls. This will lead to a large fraction of activity being deposited in the containment sump. The sump water will also retain soluble gaseous and soluble fission products such as iodides and cesium, but not noble gases. Once deposited, the iodine will remain in solution as long as the containment sump pH is maintained at or above 7.0. An analysis of the associated iodine DF for containment iodine removal and retention was also performed.

NOC-AE-07002127 Attachment 1 Page 110 of 219 4.3 LOSS OF COOLANT ACCIDENT When evaluating the impact of pH being below 7 in the long term (i.e., the elemental iodine DF for spray and natural removal), the maximum sump temperature is used in conjunction with the lowest pH, even though the first condition occurs in the first minutes of the accident and the latter condition occurs at the very end of the 30-day dose assessment period. This is a very conservative treatment of the impact of sump pH on iodine DF.

Major assumptions used to determine the elemental iodine DF in the containment are:

1. The containment sump has a pH of 6.8 at the time of the maximum sump temperature;

2. The containment sump reaches its maximum temperature of 266°F at 1600 seconds into the LOCA (based on STP analyses);

3. The mass of 1-127 is 6.2 kg (48.8 moles);

4. The mass of 1-129 is 3.2 kg (164.4 moles);

5. Assume a release fraction of 0.4 for 1-127 and 1-129; and,

6. The sump volume is 2.44E+06 liters (2.44E+09 grams at a density of 1 gmr/cc).

To determine the maximum DF for elemental iodine, both the fraction of total iodine in the sump water that is in elemental form and how much elemental iodine would have to be airborne to be in equilibrium with the remaining elemental iodine in the water must be determined.

To determine the mass fraction of dissolved iodine that is in elemental form, Equation 24 of Reference 39 is used, along with constants "a" and "b" from Table 5 of Reference 39.

Rearranging the equation yields 2[12] / [F] = 2([H+]2[I-]) / (a + (b[H+])

where 2 = the mass fraction is 2x the mole fraction, a = (6.05+/- 1.83)x 10"1, b = 1.47E-09.

Assuming the mass fraction is very small, [F] z [1]. Also, using the smaller value of "a" to maximize the elemental iodine mass fraction, and the masses for 1-127 and 1-129 and the sump volume assumed above, yields 2[12] / [I] = 2(3.5E-05[H+]2) / (4.22E-14 + 1.47E-09[H+]).

A plot of the total iodine that is in elemental form in the sump water as a function of pH (where

[H+] = 1 0 -pH) is presented in Figure 4.3-1.

NOC-AE-07002127 Attachment 1 Page 111 of 219 4.3 LOSS OF COOLANT ACCIDENT Figure 4.3-1: 12 Fraction vs pH 6.0E-05 5.0E-05 0 4.OE-05

,.0 3.0E-05 2.OE-05 1.OE-05 0.0E+00 6.70 6.75 6.80 6.85 6.90 6.95 7.00 7.05 pH The relative concentration of the elemental iodine in the sump water to that in the atmosphere is developed from Reference 39, page 55:

P = 10(629-0149T) for T in degrees Kelvin.

The same expressions for iodine speciation in the sump and partitioning between the sump and containment atmosphere appear in NUREG/CR-5950 as well as Reference 39.

Equilibrium is reached when

[I2]L / [I2ICG = P = 12L VG / I2GVL where V L = volume of liquid V G = volume of gas 12L = mass of 12 in the liquid 12G = mass of I 2 in the gas.

The mass of 12 in the gas is then 12G = 12LVG /P VL

NOC-AE-07002127 Attachment 1 Page 112 of 219 4.3 LOSS OF COOLANT ACCIDENT Using the volume of the containment and the containment sump, the relative iodine concentration is 12G / 12L = 39.2/P Therefore, the fraction of released iodine that may be airborne, as a function of temperature is 12G / 12L = (39.2 *1 0 (6.29-0.0149T) )(2)(3.5E-05[H+]2 ) / (4.22E-14 + 1.47E-09[H+]).

Since the fraction of released iodine that is airborne initially as elemental iodine is 0.0485 (Regulatory Guide 1.183), the DF will be the initial fraction divided by the fraction at any time, or DF = 0.0485 [(39.2

• 1 0 (6.29-0.0149T) )(2)(3.5E-05[H+]2) / (4.22E-14 + 1.47E-09[H+])]

Assuming the maximum containment temperature (265.87F), the DF as a function of pH is shown in Figure 4.3-2.

Figure 4.3-2: Iodine Decontamination Factor as a Function of pH 1.6E+02 1.4E+02 1.2E+02 1.0E+02 -Z 8.0E+01 6.OE+01 4.OE+01 2.OE+01 O.OE+O0 - . ..

6.7 6.75 6.8 6.85 6.9 6.95 7 7.05 pH

NOC-AE-07002127 Attachment 1 Page 113 of 219 4.3 LOSS OF COOLANT ACCIDENT A DF of 60, corresponding to a pH of 6.8, is to be used in the dose analysis, even though the calculated value of pH at 30 days is just below 6.85. Note that at a pH of 7.0, the DF approaches 150. The calculation is very conservative in that (1) the highest sump temperature is used and (2) the lowest pH is assumed throughout the duration of the accident. The DF of 60 will be exceeded at all times since early in the accident the sump pH is greater than 6.8 and later the sump temperature is much less than the maximum value.

Iodine Re-evolution from ESF Leakage The percentage of iodine in ESF system coolant leakage outside containment that becomes airborne may be assumed to be 10% as long as the pH is equal to or greater than 7.0. However, this is only true for the first day per Table 4.3-3. From Figure 4.3-1, at a pH of 6.8, the 12 fraction is about 2.5 times greater than at a pH of 7.0; and at a pH of 6.9, the 12 fraction is about 1.6 times greater than at a pH of 7.0.

Therefore, to account for the impact of a pH less than 7 on iodine re-evolution from ESF leakage, the 10% re-evolution fraction. (for a pH > 7, Regulatory Guide 1.183) is increased by the ratio of elemental iodine abundance at pH = f(t) to that at pH = 7. From t = 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to t = 20 days a factor of 1.6 (16%) (corresponding to pH = 6.9) is used, and from t = 20 days to t = 30 days a factor of 2.5 (corresponding to pH = 6.8) is used.

Impact of Using a Transient DF To judge the conservatism of using a DF value of 60 at the sump's minimum pH and maximum temperature, the equation above for DF as a function of temperature and pH was used, along with the sump temperature over time, to generate Figure 4.3-3. The sump pH was assumed to follow Table 4.3-3. (The step changes for the pH(t) plot are due to the step decreases in sump pH, per Table 4.3-3.) If iodine is allowed to re-evolve in the dose analysis according to what is implied above, the doses would certainly be lower than the current results. However, the dose reduction is probably not significant. Note the ratio of about 1.6 between the two plots from about one day to about 20 days (480 hours0.00556 days <br />0.133 hours <br />7.936508e-4 weeks <br />1.8264e-4 months <br />), and then the increase to 2.5 by 30 days (720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />).

These are the ratios that were used to increase the 10% iodine re-evolution fraction for ESF leakage. So the greater degree of precision resulting from applying the time-dependent DF concept would have no effect on the ESF leakage analysis.

Also, 0.15% of the iodine remains airborne as organic, so when the DF of 60 is applied to the elemental, the 4.85% elemental is reduced to 0.08%. The organic is still twice as great. If the DF is tracked, especially over the first eight hours when control room X/Qs are high, there would certainly be some improvement with respect to the 0.08%. However, during approximately the first two hours of that eight hours, the DF is not an issue because there is still a source, and from two to four hours, the time-dependent DF is still in the range of 200 (i.e., about 0.025%

elemental airborne). So for the case of a.DF of 60, the first two hours of the important eight-hour period would have a gaseous iodine airborne fraction of about 0.3% (elemental is equivalent to organic with a source still present), and the next six hours would have a total gaseous iodine airborne fraction of about0.23% (using an integrated percent fraction versus time metric to gauge importance yields a total of 2hr *0.3% + 6hr

• 0.23% = 2%-hour).

NOC-AE-07002127 Attachment 1 Page 114 of 219 4.3 LOSS OF COOLANTACCIDENT For the case of the transient DF, the first two hours would be the same, the next two hours would have about 0.175% gaseous iodine airborne, and the last four hours would have about 0.15%

gaseous iodine airborne (i.e., essentially organic only) (for a total of about 1.6 %-hour).

Therefore, the transient DF would decrease the gaseous iodine containment leakage control room dose contribution by about a factor of 1.25, and would not decrease the gaseous iodine ESF leakage contribution. Overall, it is estimated that the control room dose would be reduced by less than 0.1 rem TEDE if the transient DF were modeled.

Using a transient DF is too complex, considering the slight reduction in doses; therefore a DF of 60 was selected.

Figure 4.3-3: Iodine Decontamination Factor in the Containment Sump 2500 2000 1500 DF - pH(t) 1000 A000 pH=7 500 0i 0 100 200 300 400 500 600 700 800 Hours

NOC-AE-07002127 Attachment 1 Page 115 of 219 4.3 LOSS OF COOLANT ACCIDENT 4.3.3.2 Radiological Releases from ESF Equipment ECCS leakage is controlled in accordance with Technical Specification 6.8.3.a, "Primary Coolant Sources Outside Containment." As described in UFSAR Table 15.6-12, "Maximum Potential Recirculation Loop Leakage External to Containment," the maximum permitted recirculation loop leakage (i.e., ECCS leakage) is 4,140 cc/hr. These values are assumed in the CLB analysis of dose assessment to the Control Room, with a safety factor of two applied in accordance with Regulatory Guide 1.183. The ECCS leakage rate assumed in the AST LOCA analysis is the same as that for the CLB.

For determination of the dose contribution from ESF leakage, all radionuclides assumed to be released from the core (except noble gases) are assumed to be instantaneously and homogeneously mixed in the containment sump. Actual leakage from the RCS sump through ESF equipment would not start until after the recirculation phase of the accident begins.

However, for conservatism, and to decouple the dose analyses from the actual calculated recirculation start time, ESF leakage is assumed to begin at t=0.

Because the pH of the containment sump falls below 7.0 after one day, a fractional iodine release for ESF leakage greater than 10% must be considered per Regulatory Guide 1.183. A discussion of the pH used for iodine re-evolution from ESF leakage was presented in the preceding section 4.3.3.1.2.2. Using the relative DF as an indication of iodine volatility, a release fraction of 16%

of the iodine in the ESF leakage is assumed to be released for pH = 6.9 (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 480 hours0.00556 days <br />0.133 hours <br />7.936508e-4 weeks <br />1.8264e-4 months <br />) and 25% for pH 6.8 (480 hours0.00556 days <br />0.133 hours <br />7.936508e-4 weeks <br />1.8264e-4 months <br /> to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />).

This leakage is released directly into the Fuel Handling Building (FHB) from the leaked reactor coolant. It is then assumed to be released instantaneously into the environment without benefit of filtration via the FHB Exhaust Air System units.

A calculation was performed to determine the impact of a substantially degraded leakage condition (i.e., degraded condition leak rates assumed to be approximately 10 times greater than the design leak rate) for the ECCS isolation valves, thus allowing a greater than design leakage to migrate back to the Refueling Water Storage Tank (RWST). The leakage values used in the analysis ranged from 180 to 480 cc/hr. The RWST suction line isolation valves, low head/high head safety injection pumps' recirculation line isolation valves, and the containment spray pump's test line isolation valves for the three safety trains are considered in this analysis. The analysis concluded:

1. The motive force for leakage in the containment sump suction line is the high pressure in the containment resulting from the large break LOCA. This pressure is reduced, within the first 3.36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> of an accident, below a pressure capable of forcing water into the RWST. No contaminated sump water will reach the RWST via this leak path.

NOC-AE-07002127 Attachment 1 Page 116 of 219 4.3 LOSS OF COOLANT ACCIDENT

2. The containment spray pumps may be secured up to 13.4 days after initiation of a DBA LOCA and containment water will not reach the RWST via this leak path.

3. The minimum time for leakage from valves assumed to have the degraded leak rate to reach the RWST following the initiation or the recirculation phase of the DBA LOCA is 44.1 days. At this point in time the leakage into the RWST would be 1200 cc/hr.

Therefore, this potential ECCS leakage path to the RWST does not impact the LOCA analyses results.

The surveillance criteria for ESF leakage outside containment accounts for accident leakage.

During normal operations and ECCS testing, leakage is at room temperature. The total ESF leakage for the unit is compared to the 4140 cc/hr limit used in the LOCA analysis. During an accident, ESF leakage would be at a maximum temperature of 212'F. Also, at the time the injection phase of the accident ended and the recirculation phase begins (minimum of 1000 seconds into the accident, UFSAR Table 6.2.1.1-10, Revision 13), the containment building pressure would have dropped from its peak pressure of about 42 psig to about 28 psig. The LOCA analysis assumes the leakage is at room temperature.

The ESF leakage surveillance is part of the STP Contaminated System Leakage Program. During the surveillance, the ESF leakage is room temperature and under a static head from the RWST and possibly an additional dynamic head from SI pump operation (for leakage from mini-flow valves and isolation valves). Under accident conditions, the fluid will be at 212°F and under a diminishing static head from the RWST during the injection phase and a static head from the RCB sump and RCB pressure during the recirculation phase. During both the injection and recirculation phases some leakage will also be under an additional dynamic head from SI pump operation.

To correct surveillance results to accident conditions, a correction factor will be used. The analysis used a constant leak rate of 8280 cc/hr for 30 days (i.e., per Regulatory Guide 1.183, two times the sum of the simultaneous leakage limit from all components in the ESF recirculation systems established by Technical Specifications program requirement 6.8.3.a). A correction factor that considers the time-dependent accident pressure in the RCB plus the RCB emergency sump head to the RWST head during test conditions will be used to correct surveillance conditions to accident conditions.

4.3.4 Radiological Dose Models The RADTRAD 3.03 code was used to calculate the immersion and inhalation dose contributions for both the onsite and the offsite radiological dose consequences. Eight models were needed; four for the Control Room dose analysis and four for the TSC dose analysis. The offsite doses generated for each set are identical. The RADTRAD models for DBA-LOCA are graphically presented on Figures 4.3-4 through 4.3-11. The Case 1 series is for Control Room dose consequence analysis (as well as for offsite), and the Case 2 series is for TSC dose

NOC-AE-07002127 Attachment 1 Page 117 of 219 4.3 LOSS OF COOLANTACCIDENT consequence analysis (as well as for offsite). Table 4.3-4 describes the RADTRAD cases constructed for the LOCA analyses. All of these pathways are part of the CLB with the exception of the electrical penetration pathway.

Table 4.3-4 RADTRAD Models for LOCA Case Designators Figure Purpose 1 4.3-4 General Containment Leakage 2 4.3-8 lpen 4.3-5 Containment leakage into the electrical penetration area 2pen 4.3-9 (primarily to obtain airborne activity for Control Room shine as discussed below) 1esf 4.3-6 ESF leakage into the fuel handling building (FHB) with no 2esf 4.3-10 hold-up or filtration of the re-evolved iodine release lpur 4.3-7 Initial containment purge flow path dose contribution 2pur 4.3-11 In all of these models, the same basic structure is used. In order to set up a model for a specific pathway, certain junctions are actuated (solid lines) and certain junctions are closed (dashed lines). The key junctions which constitute the release path to the environment or from the environment to the Control Room/TSC are shown in heavy solid lines. For a given model graphic, a certain control volume (CV) or junction as shown may. represent one of two actual junctions or CVs. The CV or junction ID for the CV or junction applicable to the model in question is "boxed" to show that applicability. The "intermediate" CV represents the electrical penetration area for Cases 1/2 and for Cases 1/2pen and the FHB for Cases 1/2esf. It is ignored for Cases 1/2pur. The applicable x/Q set is also identified for each model.

The computer code STARDOSE was used to check the RADTRAD results for the DBA LOCA.

The RADTRAD and STARDOSE programs are radiological consequence analysis codes used to determine post-accident doses at offsite and control room locations due to immersion and inhalation. The STARDOSE code is the proprietary property of Polestar Applied Technology, Inc.

NOC-AE-07002127 Attachment 1 Page 118 of 219 Figure 4.3 RADTRAD Model for Case f Heavy lines =active pathways to environment or CR/TSC, dashed lines =closed pathways, boxed text =active control volume/junction where multiples shown Sprayed Cont 3 3]

RFT PWRfor AST 5 CV- o RFT for --- 1 2 Intermediate -

Blowdown 6.-

RFT for ESF Leakage CV ESF 3

__ C Coolant oolant CV2 Unsprayed

-- *]I*,1 CV~or 7 Environment 12 , X/Q =

F[-R *TSC

• -or 1 Icontainment Figure 4.3 RADTRAD Model for Case lpen Heavy lines = active pathways to environment or CR/TSC, dashed lines = closed pathways, boxed text = active control volume/junction where multiples shown

NOC-AE-07002127 Attachment 1 Page 119 of 219 Figure 4.3 RADTRAD Model for Case 1 esf Heavy lines = active pathways to environment or CR/TSC, dashed lines = closed pathways, boxed text = active control volume/junction where multiples shown Figure 4.3 RADTRAD Model for Case Ipur Heavy lines = active pathways to environment or CR!TSC, dashed lines = closed pathways, boxed text = active control volume/junction where multiples shown RFT for- RFT fr --1 -* Intermediate CV4°r5 "

Blowdown RFT for 1V ESF Leakage Unsprayed t

'.C V 3 -- -- -- ----- ---- S 4] CN En ESF Coolant _I, ln CV H or 7 12 X/I I*,l1or TSC vei

NOC-AE-07002127 Attachment 1 Page 120 of 219 Figure Heavy lines = active pathways environmentModel 4.3-8 -toRADTRAD for Case 2 or CR/TSC, dashed lines = closed pathways, boxed text =active control volume/junction where multiples shown CV I Sprayed Cont RFT for PWR AST 1---

CV I_ ,,

k 10Va4or*

5-RFT for ] ..... 2 Intermediate Blowdown 6 RFT for C ESF Leakage Unsprayed ,

"_. CV 3 ...... Environm ent ESF Coolant CV 6 orXQ 8,[ CR or " containment leakage Figure 4.3 RADTRAD Model for Case 2pen Heavy lines = active pathways to environment or CR/TSC, dashed lines = closed pathways, boxed text = active control volume/junction where multiples shown RFTfor CV213 ESF Leakage Unsprayed CV3 ESF Coolant 8,~ CRorý* K,

NOC-AE-07002127 Attachment 1 Page 121 of 219 Figure 4.3 RADTRAD Model for Case 2esf Heavy lines = active pathways to environment or CR/TSC, dashed lines = closed pathways, boxed text - active control volume/junction where multiples shown RFT for PWR AST RFT for Blowdown Figure 4.3 RADTRAD Model for Case 2pur Heavy lines = active pathways to environment or CR/TSC, dashed lines = closed pathways, boxed text = active control volume/junction where multiples shown

NOC-AE-07002127 Attachment 1 Page 122 of 219 4.3 LOSS OF COOLANTACCIDENT 4.3.4.1 Control Room and TSC Dose to operators in the Control Room and to TSC personnel are from two main pathways:

• Dose from airborne contaminants in the Control Room/TSC

• Dose from gamma sources outside the Control Room/TSC These sources are discussed in the following sections.

Consideration of Single Failure Without credit being taken for the FHB filters or for the Control Room make-up filters (and the associated heaters to control intake humidity), the single-failure assessment becomes much simpler for application of the AST than that of the CLB. For the AST DBA LOCA, an electrical division electrical failure is assumed as a single failure to minimize containment mixing via the containment fan-coolers. This assumption maximizes dose. Only two out of three trains of containment ventilation are assumed to operate, and one reactor containment fan-cooler on one of the operating trains is assumed to be out of service, as well. The spray removal lambdas used are also consistent with the loss of one spray train, as are the assumptions regarding Control Room ventilation and filtration.

4.3.4.1.1 CR/TSC Doses from Airborne Contaminants The analytical models used for the Control Room and TSC are described in Sections 4.2.2 and 4.2.3, respectively. Releases are assumed to be drawn into the Control Room/TSC HVAC intakes (points D/H on Figure 4.1-13). The atmospheric dispersion factors are developed in Section 4.1.3. Unfiltered in-leakage and possible "sneak" paths into the CRE are addressed in Section 4.2.2.1.

4.3.4.2 CR/TSC Doses from Gamma Shine The gamma shine dose contribution consists of four parts:

" Gamma shine from the containment airborne activity

• Gamma shine from airborne activity in the electrical penetration area (CR only)

• Gamma shine from activity in the external radioactive cloud surrounding the plant structures

• Gamma shine from trapped activity on filters

NOC-AE-07002127 Attachment 1 Page 123 of 219 4.3 LOSS OF COOLANT A CCIDENT The DBA LOCA radiation dose to personnel in the Control Room includes the gamma shine from the primary containment airborne activity, from airborne activity in the electrical penetration area, from activity in the radioactive cloud surrounding the plant structures, and from trapped activity on filters. Of these four contributors, all but the shine from the electrical penetration area are in the CLB. A tabulation of the dose components for the gamma shine doses is presented in Table 4.3-9.

Gamma shine from the containment airborne activity For shine from the containment, a comparison of source gamma power was made to either justify the use of the calculated dose as-is or to adjust the CLB value for AST application.

It was determined that the time-integrated activity of radioiodine airborne in the containment would be approximately an order of magnitude lower for the AST than for the STP CLB. The airborne noble gas is comparable for both the AST and the STP CLB. While the AST involves the airborne release of significant quantities of additional non-iodine activity in particulate form, this activity is readily removed by filtration and plate-out. The external gamma dose from non-iodine airborne particulate for the AST is only 10% of that for the iodine. Because of this behavior, it is evident that radiation from the containment will be less for the AST than for the STP CLB.

The STP Control Room CLB shine dose due to activity airborne in the containment is 0.101 rem.

By_.a comparison of the basis for the 0.101 rem (in terms of transient airborne activity) to the transient airborne activity for the AST, this value was determined to be bounding. Therefore, it has been used as-is for the AST application as a dose increment for the Control Room dose consequence analysis. In a similar fashion, the CLB shine dose to the TSC due to the RCB airborne activity of 0.004 rem is bounding for the AST analysis.

Gamma shine from airborne activity in the electrical penetration area The Electrical Penetration area is directly between the Control Room Envelope and the containment building (on the bottom of the Control Room Envelope, west of the Relay Room and Computer Room, as depicted in Figure 4.2-1). This exposure source was not considered in the CLB.

For Control Room shine from the electrical penetration area just outside containment, a compartment was added to the plant model for both the RADTRAD and the STARDOSE DBA LOCA analyses. The maximum post-LOCA containment temperature and pressure listed in Table 4.3-11 were used to convert the electrical penetration tested mass leak rate (expressed in sccm) to a volumetric leak rate for use in the dose analysis model. The transient airborne activity within this compartment was calculated using these models. From this transient airborne activity, a dose calculation was performed for shine dose in the Control Room. This calculation was performed using the MicroShield code (Reference 40). MicroShield is a point kernel integration code used for general-purpose gamma shielding analysis. The dose contribution from

NOC-AE-07002127 Attachment 1 Page 124 of 219 4.3 LOSS OF COOLANT ACCIDENT this source is 0.0174 rem. This value has been used for the AST application as a dose increment for the Control Room dose consequence analysis.

Gamma shine from activity in the external cloud surrounding the plant structures For shine from the radioactive cloud, the 30-day shine dose increment for the LPZ (no shielding protection considered and no occupancy factor credited) was adjusted by the ratio of the maximum onsite x/Q value to that for the LPZ for each x/Q averaging period. The result was then reduced by a shielding attenuation factors for the Control Room and TSC (Table 4.3-5).

The final shine dose was obtained by adding the increments for each averaging period., The dose contribution from this source is 0.014 rem to the Control Room and 0.212 rem to the TSC.

Gamma shine from trapped activity on filters Similar to the treatment of the shine from the containment, a comparison of source gamma power was made to either justify the use of the calculated dose as-is or to adjust the CLB value for AST application.

It was determined that the time-integrated activity of radioiodine airborne in the containment would be approximately an order of magnitude lower for the AST than for the STP CLB. The airborne noble gas is comparable for both the AST and the STP CLB. While the AST involves the airborne release of significant quantities of additional non-iodine activity in particulate form, this activity is readily removed by filtration and plate-out. The external gamma dose from non-iodine airborne particulate for the AST is only about 10% of that for the iodine. Because of this behavior, it is evident that radiation from the activity trapped on filters will be less for the AST than for the STP CLB.

This source of exposure is part of the CLB for both the Control Room and the TSC. Three of the filter shine contributions assessed as part of the CLB are negligible: (1) Control Room shine dose due to Control Room make-up filters, (2) Control Room shine dose due to TSC make-up/recirculation clean-up filter, and (3) TSC shine dose due to TSC make-up/recirculation clean-up filter. The Control Room shine dose due to the Control Room recirculation make-up filters is 0.00218 rem, and the TSC shine dose due to the Control Room make-up filters is 0.844 rem.

The CLB activity trapped on one Control Room make-up filter is shown on Table 4.3-6. The gamma power due to activity on two Control Room make-up filters (the basis for the CLB TSC shine dose contribution) is shown on Table 4.3-7.

For the CLB, it is assumed that the activity trapped on the Control Room recirculation clean-up filters is one-tenth of that on the Control Room make-up filters (i.e., one-tenth of Table 4.3-6) due to the iodine removal by the makeup filters. The CLB Control Room shine dose due to the Control Room recirculation clean-up filters is based on one-tenth of Table 4.3-7.

NOC-AE-07002127 Attachment 1 Page 125 of 219 4.3 LOSS OF COOLANT ACCIDENT The AST gamma power due to activity trapped on two Control Room make-up filters is presented in Table 4.3-7, and the AST gamma power due to activity trapped on two Control Room recirculation clean-up filters is presented in Table 4.3-8.

Note that in the AST analysis, the Control Room inhalation and immersion doses were calculated without benefit of the Control Room make-up filters. The Control Room recirculation filter loading was taken from Table 4.3-11. A separate calculation for the Control Room make-up filter loading was done solely for the purpose of evaluating the gamma shine contribution to the TSC. This second analysis assumed 100% filter efficiency for the Control Room make-up filters.

As can be seen from Table 4.3-7, the CLB Control Room make-up filter loading is about an order of magnitude greater than that for the AST. On this basis, the CLB 0.844 rem Control Room make-up filter shine dose contribution to the TSC has been. used as-is for the AST application as a dose increment for the TSC dose consequence analysis.

As can be seen from a comparison of one-tenth of Tables 4.3-7 and 4.3-8, the maximum ratio.of AST Control Room recirculation clean-up filter loading to the CLB Control Room recirculation clean-up filter loading is 1.74 at 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />. On this basis, the CLB 0.00218 rem Control Room recirculation clean-up filter shine dose contribution to the Control Room has been increased by a factor of 1.74 to 0.0038 rem for the AST application as a dose increment for the Control Room dose consequence analysis.

Table 4.3-5 CR and TSC Gamma Shine Dose Analysis Inputs for DBA LOCA Input/Assumption CLB Analysis AST Analysis Attenuation factor for Control Room shine from 1.03E-3 1.03E-3 atmospheric activity Attenuation factor for TSC shine from 1.03E-3 1.56E-2 atmospheric activity Activity on one Control Room make-up filter Table 4.3-6 N/A Gamma power due to activity on two Control Table 4.3-7 Table 4.3-7 Room make-up filters

NOC-AE-07002127 Attachment 1 Page 126 of 219 4.3 LOSS OF COOLANT ACCIDENT Table 4.3-5 CR and TSC Gamma Shine Dose Analysis Inputs for DBA LOCA Input/Assumption CLB Analysis AST Analysis Gamma power by photon energy due to activity One-tenth of One-tenth of on Control Room recirculation clean-up filters Table 4.3-7 Table 4.3-7 AST gamma power by photon energy due to activity on Control Room recirculation clean-up N/A Table 4.3-8 filters Maximum ratio of one-tenth of Table 4.3-7 to TableN/A 1.74 at 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> Table 4.3-6 Current Licensing Basis Activity on One Control Room Make-Up Filter (Ci)

Time Radionuclide (hours) 1-131 1-132 1-133 1-134 1-135 0.05 8.76E-2 1.26E-1 1.88E-1 1.96E-1 1.71E-1 0.7882 4.41E-1 5.1OE-1 9.23E-1 5.51E-1 8.OOE-1 8 1.65 2.19E- 1 2.78 6.96E-3 1.44 24 3.24 3.57E-3 3.41 - 5.61E-1 44 3.63 - 2.10 - 8.31E-2 64 3.97 - 1.28 - 1.19E-2 84 4.22 - 7.54E-1 - 1.68E-3 96 4.36 - 5.45E-1 - 5.14E-4 150 4.07 - 1.04E-1 - -

400 2.58 - - -

720 1.18 -

NOC-AE-07002127 Attachment 1 Page 127 of 219 4.3 LOSS OF COOLANTACCIDENT Table 4.3-7 Gamma Power from Activity on Two Control Room Make-Up Filters (MeV/sec)

Time (hours) CLB AST  % Difference 0.05 9.03E+10 2.78E+09 -96.9%

0.7882 3.41E+ 1 2.16E+l 0 -93.7%

8 3.8E+11 4.84E+10 -87.3%

24 3.1E+11 2.82E+10 -90.9%

44 2.1E+11 2.21E+10 -89.5%

64 1.7E+l 1 1.97E+10 -88.4%

84 1.5E+l 1 1.84E+10 -87.7%

96 1.5E+l1 1.79E+10 -88.1%

150 1.2E+1I 1.58E+10 -86.8%

400 7.3E+10 1.11E+10 -84.8%

720 3.3E+10 7.39E+09 -77.6%

NOC-AE-07002127 Attachment 1 Page 128 of 219 4.3 LOSS OF COOLANT ACCIDENT Table 4.3-8 Gamma Power from Activity on Two Control Room Cleanup Filters (Mev/sec)

Time (hours) CLB AST  % Difference 0.05 4.14E+10 2.19E+08 -99.5%

0.7882 1.57E+1 1 7.54E+09 -95.2%

8 1.76E+1 1 3.66E+10 -79.2%

24 1.45E+ 11 2.12E+10 -85.4%

44 9.68E+10 1.65E+10 -82.9%

64 8.02E+10 1.48E+10 -81.5%

84 7.23E+10 1.39E+10 -80.8%

96 6.96E+10 1.36E+10 -80.5%

150 5.67E+10 1.21E+10 -78.7%

400 3.45E+10 8.66E+09 -74.9%

720 1.58E+10 5.79E+09 -63.4%

Table 4.3-9 Gamima Shine Component Doses (rem)

Source Receptor CLB AST RCB Activity CR 0.101 0.101 External Cloud CR 0.464 0.014 CR Makeup Filters CR Negligibl e Negligible CR Recirc Filters CR 0.00218 0.004 Electrical Penetration Room CR N/A 0.017 TOTAL 0.567 0.136 RCB Activity TSC 0.004 0.004 External Cloud TSC 0.464 0.212 CR Makeup Filters TSC 0.844 0.844 TSC Makeup/Recirc Filters TSC Negligible Negligible Electrical Penetration Room TSC N/A Negligible TOTAL 1.312 1.060

NOC-AE-07002127 Attachment I Page 129 of 219 4.3 LOSS OF COOLANT ACCIDENT 4.3.5 Inputs and Assumptions The following assumptions are made in the LOCA analyses.

Release into Containment

1. The source term is based upon a power level of 4100 MW thermal, 5 w/o enrichment, and a three region core with equilibrium cycle core at end of life. The three regions have operated at a specific power of 39.3 MW/MTU for 509, 1018, and 1527 EFPD, respectively. The assumed power level is greater than the Rated Thermal Power of 3853 MWth plus a 0.6% measurement uncertainty. The AST requires the consideration of additional radionuclides to ensure that the TEDE dose (which considers organs other than thyroid) is properly calculated. For the DBA LOCA, all fuel assemblies in the core are assumed to be affected, and the core average inventory is used.

2. A total of 100 percent of the core noble gas inventory and 50 percent of the core iodine inventory is assumed to be immediately available for leakage from the Containment.

3. For the AST, in accordance with Regulatory Guide 1.183, of the radioiodine released from the reactor core, 95 percent of the iodine released is assumed to be particulate in the form of cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. This includes releases from the gap and the fuel pellets. For the CLB, in accordance with Regulatory Guide 1.4 (Reference 41), of the iodine activity released to the Containment, it is assumed that 95.5 percent is in the elemental form, 2 percent is in the organic or methyl iodine from, and 2.5 percent is in particulate form.

4. The radioactivity released from the fuel is assumed to mix instantaneously and homogeneously throughout the containment air space as it is released. The distribution is not adjusted for internal compartment effects.

Release via Containment Supplemental Purge

5. The Containment Supplementary Purge System is assumed to be in operation and the purge is assumed to be isolated within 23 seconds of the generation of the Containment Pressure-High 1 signal. This includes the signal and sequencer delays, Standby Diesel Generator startup time, and the valve closing time. Thistime dose not include the 1.2 seconds between the postulated instantaneous break and the containment pressure reaching the High-I setpoint. However, the constant value used for the choke flow through the ventilation system bounds the effect of neglecting these 1.2 seconds. During normal power operation, the Containment Supplementary Purge System vents the containment at 4,500 ft3/min. However, for this analysis, the maximum flow rate due to the pressure spike inside the Containment was used (83,200 ft3/min for each purge line, intake and exhaust).

6. The coolant activity in the AST analysis does not include iodine spiking (per the guidance of Regulatory Guide 1.183, Appendix A, Section 3.8). However, the RCS

NOC-AE-07002127 Attachment 1 Page 130 of 219 4.3 LOSS OF COOLANT ACCIDENT concentrations are based on 1% failed fuel, which are greater than those corresponding to the 1.0 gCi/gm Technical Specification limit on DE 1-131. The CLB assumes the Containment airborne iodine inventory available for release is the flashed portion of the total primary coolant iodine inventory based upon a preexisting iodine spike level of 60 p.Ci/g dose equivalent 1-131. For noble gases, 100 percent of the primary coolant inventory based upon 1% failed fuel is assumed to be available for release. In both cases, no failed fuel due to the accident is assumed to have occurred because isolation occurs prior to the core reaching a temperature that could cause a fuel failure.

Fission Product Removal Inside Containment

7. A two-volume model of the Containment is used to represent sprayed and unsprayed regions of the Containment.

8. Of the six Reactor Containment Fan Cooler (RCFC) units, only three are assumed to function. Two are failed due to the assumed failure of a Standby Diesel Generator to start upon Loss of Offsite Power and one is assumed down for maintenance.

9. The transfer rate between the sprayed and unsprayed regions is assumed to be limited to the forced convection induced by the RCFC units. The assumed minimum flow rate conservatively neglects the effects of natural convection, steam condensation, and diffusion, although these effects are expected to enhance the mixing rate between the sprayed and unsprayed volumes. The majority of the RCFC air supply, except a small portion discharged to the dome, is discharged to the space within the secondary shield wall, where it is relieved to the balance of the Containment volume through the vent areas.

10. For fission products other than iodine, the only removal processes considered are radioactive decay and leakage. Iodine is assumed to be removed by radioactive decay and leakage, plateout, and also by the Containment Spray System (CSS).

11. The AST analyses change the ultimate DF to 60 instead of a value of 100 as used in the CLB. This change is based on pH decreasing below 7.0 at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

.12. Since the AST elemental iodine activity release to the containment is different in magnitude and timing as compared to the CLB, the elemental iodine DF of 60.is reached at a different time. Also, once the elemental iodine DF of 60 is reached, all elemental iodine removal, both natural removal and removal by spray, is terminated. For the CLB, a spray removal rate of 20 per hour is assumed until the airborne elemental iodine is reduced by a factor of 60. After this time, the elemental spray removal rate is assumed to be zero.

13. For the AST, the natural removal rate in the containment for elemental iodine is changed to 4.5 per hour. This is due to the application of an ultimate DF of 60 (instead of 100) based on the sump pH decreasing below 7.0 after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The CLB uses 3.59 per hour

NOC-AE-07002127 Attachment 1 Page 131 of 219 4.3 LOSS OF COOLANT ACCIDENT for the sprayed region and 0.91 per hour for the unsprayed region. This is based on partitioning the 4.50 per hour total removal rate between the regions based on relative volumes. However, if it is assumed that the volumes have the same surface area to volume ratio, then 4.5 may be used for both volumes. For the CLB, the deposition removal rate for elemental iodine is assumed to be 4.5 per hour which is reduced to 5% of this value once a DF of 100 is reached and no additional credit is taken for deposition after a DF of 200 is reached.

14. For the CLB, for particulate iodine, a spray removal rate of 6.9 per hour is assumed until a DF of 50 is reached and it is then reduced to 10% of this value until a DF of 1000 is.

reached.

15. Since the AST particulate activity release to the containment is different in magnitude and timing as compared to the CLB, the particulate DF of 50 (the time at which the spray removal rate is reduced by a factor of ten) is reached at a different time. For the CLB, the analysis assumes containment spray for 380 minutes for removal of particulate iodine.

Release from Containment

16. The Containment leak rate to the atmosphere used in the analysis is the design-basis leak rate indicated in the Technical Specifications. For the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the leak rate is assumed to be 0.30 percent per day, while for the remainder of the 30-day period the leak rate is assumed to be 0.15 percent per day. This Containment leakage is assumed to leak directly to the environment.

17. To support the revised analysis consideration of the gamma "shine" dose to the Control Room, leakage from the containment into the adjacent electrical penetration room is assumed based on the relative number of penetrations in the penetration area.

ESF Leakage Release

18. The amount of water in the Containment sumps at the start of recirculation is the total of the RCS water and the water added due to operation of the engineered safeguards, i.e.,

the ECCS and CSS. This amount has been calculated to be 512,494 gallons. This value is conservatively low to maximize iodine concentration in the sump water.

19. Since most of the radioiodine released during the LOCA would be retained by the Containment sump water dueto operation of the CSS and the ECCS, it is conservatively assumed that 50 percent of the core iodine inventory is introduced to the sump water to be recirculated through the external piping systems. Because noble gases are assumed to be available for leakage from the Containment atmosphere and are not readily entrained in water, the noble gases are not assumed to be part of the source term of this contribution to the total LOCA dose.

20. For the fractional release of iodine from the ESF leakage, the Regulatory Guide 1.183 recommended value of 10% is used only for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Beyond that time, the

NOC-AE-07002127 Attachment 1 Page 132 of 219 4.3 LOSS OF COOLANTACCIDENT release fraction is increased to 16% (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 480 hours0.00556 days <br />0.133 hours <br />7.936508e-4 weeks <br />1.8264e-4 months <br />) and then to 25% (480 hours0.00556 days <br />0.133 hours <br />7.936508e-4 weeks <br />1.8264e-4 months <br /> to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />) based on the calculated volatility of the iodine for the pH values over those intervals relative to the volatility for a pH of 7.0. The CLB assumes that 10% of the iodine is released into the Fuel Handling Building.

21. The maximum potential recirculation loop leakage is 4140 cc/hr. This value represents expected leakages from ESF equipment and is the total leakage from all three trains of ESF equipment. The radiological dose model does not distinguish between the specific source, component, or train of the ESF leakage. The radiological dose model conservatively uses twice the total leakage.

22. The iodine activity released from the ESF leakage, once released to the atmosphere of the FHB, is assumed to be quickly transported by the ventilation system through the exhaust filters and released to the environment at ground level. The AST analysis assumes no filtration of the iodines released to the environment. The CLB assumes the iodine filtration efficiency to be 95 percent.

Control Room HVAC

23. The Control Room ventilation system is assumed to automatically transfer to the emergency mode of operation after the initiation of safety injection.

24. The AST analyses use .the nominal Control Room HVAC flow rates plus uncertainties to more conservatively model the Control Room HVAC system. The Control Room make-up flow is increased from nominal 2000 cfm to 2200 cfm to allow for tolerances, and the Control Room recirculation flow is decreased from 9500 cfln to 8600 cfm to allow for tolerances. The Control Room make-up filter is conservatively ignored except for determining the filter shine dose to the TSC. 100 cfm of unfiltered in-leakage is assumed in addition to the 2200 cfln of make-up flow that is assumed to experience no filtration at all. This yields a total of 2300 cfm of unfiltered in-leakage.

Miscellaneous

25. Offsite Power is lost. One Standby Diesel Generator fails to start. This causes the loss of one train of Control Room emergency HVAC and the loss of one train of RCFC's (two RCFC units).

26. For determination of offsite doses, all activity is released to the environment with no consideration given to cloud depletion by ground deposition during transport to the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ).

Input parameters used for the LOCA analysis are given in Table 4.3-11. Conformance with Regulatory Guide 1.183 guidance addressing LOCA analysis is provided in Attachment 6, Tables A and B.

NOC-AE-07002127 Attachment 1 Page 133 of 219 4.3 LOSS OF COOLANT ACCIDENT Table 4.3-10 LOCA Time-Dependent Release Fractions Time Period (sec) Fraction of core inventory37 0-30 No Release 30-1830 Gases Xe, Kr- 0.1/hr (0.05 total)

Elemental I - 4.9E-3/hr (2.4E-3 total)

Organic I - 1.5E-4/hr (7.5E-5 total)

Aerosols I, Br - 0.095/hr (0.0475 total)

Cs, Rb - 0.1/hr (0.05 total) 1830 - 6510 Gases Xe, Kr - 0.73/hr (0.95 total)

Elemental I- 1.3E-2/hr (1.7E-2 total)

Organic I - 4.OE-4/hr (5.3E-4 total)

Aerosols I,,Br - 0.256/hr (0.3325 total)

Cs, Rb - 0.1 92/hr (0.25 total)

Te Group - 0.038/hr (0.05 total)

Ba, Sr - 0.015/hr (0.02 total)

Noble Metals - 1.9E-3/hr (2.5E-3 total)

La Group - 1.5E-4/hr (2E-4 total)

Ce Group - 3.8E-4/hr (5E-4 total) 37 From RG 1.183 Table 2 considering the chemical form described in RG 1.183, Section 3.5.

NOC-AE-07002127 Attachment 1 Page 134 of 219 4.3 LOSS OF COOLANTACCIDENT Table 4.3-11 Dose Analysis Inputs for LOCA Input/Assumption Current Licensing Basis Proposed AST Core power level 4100 MWt (for radiological source terms)

Core power level 3876 MWt (for RCS steam releases for (3853 MWt + 0.6%)

supplemental purge)

Core inventory per MWt Table 4.3-1 Activity in coolant blowdown Table 4.2-14 (1% failed fuel)

Coolant blowdown mass (parameter not used) 9.3E5 Ibm Activity release from overheated fuel Table 4.2-9 Table 4.2-9 (Iodines and noble gases only)

Volume of containment sprayed region 2.7E6 ft3 Volume of containment unsprayed 6.8E5 ft3 region Volume of water in containment sump 61,486 ft3 Volume of electrical penetration area N/A 101,477 ft3 Volumetric flowrate due to open purge 142,000 cfm valves Duration of flow through open purge 23 seconds valves Volumetric flowrate between sprayed i 38 152,475 cfrn and unsprayed regions of containment (3 of 6 coolers)

Volumetric leakrate from containment 0.3%/day, first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.15%/day,24-720 hours ESF leakrate 4140 cc/hr (analyzed at 8280 cc/hr) 38 Values given are for full flow. Less than 10% of this total will recirculate in the unsprayed region, and dose is insensitive to mixing flow bypass of this magnitude.

NOC-AE-07002127 Attachment 1 Page 135 of 219 4.3 LOSS OF COOLANTACCIDENT Table 4.3-11 Dose Analysis Inputs for LOCA Input/Assumption Current Licensing Basis Proposed AST Fraction of radioiodine released from 10% (0-24 hours)

ESF leakage 10% 16% (24-480 hours) 25% (480 hour0.00556 days <br />0.133 hours <br />7.936508e-4 weeks <br />1.8264e-4 months <br />s-720 hours)

Fraction of core iodine inventory 50% See Table 4.3-10 released to RCB Fraction of Iodines released into the 50% See Table 4.3-10 RCB which is available for release Iodine Species for the lodines Released to RCB 91%/4%/5% 4.85%/0.15%/95%

(elemental/organic/particulate)

Iodine Species in ESF Leakage 97%/3%/0%

(elemental/organic/particulate)

Containment electrical penetration N/A 100 sccm per penetration leakrate Number of containment electrical N/A 18 penetrations Ventilation exhaust rate for electrical N/A 833 cfm penetrdtion area Spray start time 2.34 minutes Maximum post-LOCA containment N/A 41.2 psig pressure Maximum post-LOCA containment N/A 330 F temperature Assumed FHB exhaust rate (for ESF Infinite leakage)

Assumed FHB filter efficiency Elemental 95% 0%

Organic 95% 0%

Particulate 99% 0%

Dose Conversion Factors Table 4.2-6 Decay Constants and Decay Daughter Table 4.2-7 Fractions Offsite breathing rates Table 4.2-1

NOC-AE-07002127 Attachment 1 Page 136 of 219 4.3 LOSS OF COOLANTACCIDENT Table 4.3-11 Dose Analysis Inputs for LOCA Input/Assumption Current Licensing Basis Proposed AST Offsite X/Q's Table 4.1-24 Control Room HVAC Parameters Table 4.2-3 Control Room HVAC Flow Rates Table 4.2-2 TSC HVAC Parameters Table 4.2-5 TSC HVAC Flow Rates Table 4.2-4 Control Room and TSC X/Q's Table 4.1-37 Table 4.3-12 CLB Spray Removal Parameters Elemental 39 Particulate/Aerosol Time Sprayed Unsprayed Sprayed Unsprayed Event (hr) Region Region Region Region Break 0.0000 3.59 0.91 0.0 0.0 CSS Start 0.039 23.59 0.91 6.9 0.0 Elemental DF of 60 Reached 0.2808 3.59 0.91 6.9 0.0 Elemental DF of 100 Reached 0.62 0.0 0.0 6.9 0.0 Particulate DF of 50 Reached 0.7559 0.0 0.0 0.7 0.0 Particulate DF of 1000 Reached 6.335 0.0 0.0 0.0 0.0 30 days 720. 0.0 0.0 0.0 0.0 Table 4.3-13 AST Spray Removal Parameters Elemental 40 Particulate/Aerosol Time Sprayed Unsprayed Sprayed Unsprayed Event (hr) Region Region Region Region Break 0.0000 4.5 4.5 0.0 0.0 CSS Start 0.039 24.5 4.5 6.9 0.0 Elemental DF of 60 Reached 1.855 0.0 0.0 6.9 0.0 Particulate DF of 50 Reached 2.185 0.0 0.0 0.7 0.0 30 days 720. 0.0 0.0 0.0 0.0 39 Includes removal by deposition 40 Includes removal by deposition

NOC-AE-07002127 Attachment 1 Page 137 of 219 4.3 LOSS OF COOLANT ACCIDENT 4.3.6 Summary and Conclusions Radiological doses resulting from a design-basis LOCA to a Control Room operator and aperson located at the EAB or LPZ are to be less than the regulatory dose limits given 10CFR50.67.

Table 4.3-14 presents the results of the LOCA radiological consequence analysis.

Table 4.3-14 LOCA Dose Results (rem TEDE)

EAB Control Room/TSC (worst 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) LPZ (30 days)

Dose Component Result Limit Result Limit I Control Room TSC Limit Containment Leakage41 5.49 2.52 1.93 0.11 Elec. Penetration Room 0.01 0.01 0.02 Negligible ESF Leakage 0.10 0.27 1.57 0.04 Supplemental RCB Purge 0.02 0.01 0.02 Negligible Shine dose N/A N/A 0.14 1.06 TOTAL 5.62 25 2.81 25 3.68 1.21 I 5j1 The worst 2-hour dose at the EAB is between t=0 and 3.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> and is less than 5.55 rem TEDE.

This is developed from the worst 2-hour EAB dose from four separate RADTRAD cases. The time frame represents the earliest start of a worst 2-hour interval and the latest end of a worst 2-hour interval from these four runs.

41 The containment release is direct to the environment.

NOC-AE-07002127 Attachment 1 Page 138 of 219 4.4 FUEL HANDLING A CCIDENT 4.4 Fuel Handling Accident (FHA) Radiological Assessment 4.4.1 Methodology Overview This postulated refueling accident involves the drop of a fuel assembly on top of other fuel assemblies during refueling operations. The mechanical part of the analysis remains unchanged from the CLB; the total number of failed fuel rods is 314 (out of 50952 for an entire core). The depth of water over the damaged fuel is not less than 23 feet.

Following reactor shutdown, decay of short-lived fission products greatly reduces the fission product inventory present in irradiated fuel. The proposed amendment takes credit for the normal decay of irradiated fuel rather than crediting certain active mitigative systems (e.g.,

ventilation filtration systems). Since radioactive decay is a natural phenomenon, it has a reliability of 100 percent in reducing the potential radiological release from the fuel assemblies.

In addition, the water level that covers the fuel assemblies is another natural method that provides an adequate barrier to a significant radiological release. This defense-in-depth method will continue to be enforced by Technical Specification controls. Technical Specifications 3/4.9.10 and 3/4.9.11 control water level over irradiated fuel assemblies.

The analysis is fully compliant with Regulatory Guide 1.183. The analysis was performed for 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> after shutdown assuming a ground-level release through the plant vent. An accident in either the containment or the FHB would involve a release via the plant vent or possibly directly from the containment. However, a release directly from the containment would experience more favorable atmospheric dispersion on the path to the Control Room and TSC air intake than a release frbm the plant vent because of the greater distance involved.

4.4.2 Analytical Model A simple analytical model is used for the proposed analysis. The activity released from the fuel assembly is scrubbed by the pool water and then immediately released into the environment. No credit is taken for building holdup or HVAC filtration. The activity release is multiplied by the appropriate x/Q (EAB, LPZ or Control Room/TSC air intake) to obtain the activity concentration at that location. The offsite X/Qs and Control Room/TSC x/Qs are provided in Tables 4.1-24 and 4.1-37, respectively.

NOC-AE-07002127 Attachment 1 Page 139 of 219 4.4 FUEL HANDLING ACCIDENT The air concentrations are multiplied by the DCFs taken from the RADTRAD AST default file for CEDE and EDE WB (Effective/Inhaled Chronic and Effective/Cloudshine, respectively, with the appropriate conversion from Sv/Bq to rem/Ci) and by the assumed breathing rate of 3.5E-4 m 3/sec from Regulatory Guide 1.183 for the CEDE. The CEDE and EDE WB doses are combined for each location to obtain the TEDE. For the Control Room, the EDE WB is reduced by the finite volume correction factor described in Section 4.2.7 of Regulatory Guide 1.183 prior to calculating the TEDE. Because this correction factor reduces the dose and varies with (volume ratio)0 "338 and because the TSC volume is smaller than that of the Control Room, the FHA Control Room dose is limiting for the TSC.

No credit is taken for filtration by the FHB filters or for hold-up in either the containment or the FHB. No credit is taken for any filtration (make-up or recirculation clean-up) for either the Control Room or the TSC. The only benefit afforded by the Control Room or TSC envelope is the finite volume EDE WB dose correction. Because of the simplicity of this model, it applies to a FHA both in the containment and in the FHB.

4.4.3 Radiological Source Term Following accident initiation at 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> after shutdown, the radionuclide inventory from the damaged fuel pins is assumed to leak out to the environment instantaneously (even though releases to the environment could be assumed to occur over a 2-hour period according to the NRC regulatory guidance). The core radial peaking factor used is 1.7 (the CLB value). This is a conservatively high peaking factor which bounds past operating experience at STP and is expected to bound future core designs as well. The cycle-specific peaking factor limits are stated ineach cycle's Core Operating Limits Report (COLR).

The ORIGEN 2.1 code was used to calculate plant-specific fission product inventories for use in the FHA dose analyses. The fraction of the core that is damaged is assumed to be one fuel assembly (264 fuel pins) plus an additional 50 pins in an impacted assembly (314 pins total) out of 50,952 pins in the core. A peaking factor of 1.7 was applied to the fission product inventory of these pins. This peaking factor value is a practical bounding value for the peaking factors found in the cycle-specific Core Operating Limits Repot (COLR), based on previous core design history and future projections. The gap fractions of Table 3 of Regulatory Guide 1.183 were also applied. This result is the activity release from the damaged fuel.

The core inventory of relevant radionuclides is shown on Table 4.4-1. The CLB activities are based on the discharge batch of a three-region core. The AST activities are based on the core average inventory. Note that the AST activities bound the CLB activities. The values in the table reflect a core average gap inventory and do not include the 1.7 power peaking factor, the 314/50952 pin fraction, or any effects from pool water scrubbing. Since alkali metal releases (as particulates) are assumed to experience an infinite DF due to the water submergence (per Regulatory Guide 1.183), no alkali metals (e.g., Cs and Rb) are included.

NOC-AE-07002127 Attachment 1 Page 140 of 219 4.4 FUEL HANDLING ACCIDENT Table 4.4-1 42 Base Fission Product Gap Inventory for the FHA CLB AST Ci/MW Ci/MW Ci/MWt Ci/MWt @t = 42 hr @t = 0 @t =42 hr Isotope @t = 0 (shutdown) (accident) (shutdown) (accident)

Kr83m - 3.41E+03 5.05E-04 Kr85m - 7.07E+03 9.26E+00 Kr85 1.2E+02 1.23E+02 5.86E+02 5.86E+02 Kr87 - 1.34E+/-04 1.40E-06 Kr88 - 1.90E+04 5.77E-01 Kr89 - 2.32E+04 02 Xel31m 2.OE+01 1.81E+01 2.68E+02 2.79E+02 Xel33m 8.2E+02 4.73E+02 1.66E+03 9.79E+02 Xel33 5.6E+03 4.43E+03 5.37E+04 4.87E+04 Xe135m 1.2E+03 3.08E-47 1.02E÷04 1.08E+02 Xel35 1.4E+03 5.76E+01 1.34E+04 3.83E+03 Xe137 4.63E+04 0 Xe138 4.39E+04 9.77E-41 1131 3.5E+03 2.98E+03 4.14E+04 3.56E+04 1132 4.2E+03 1.19E-02 3.71E+04 1.38E-01 1133 5.6E+03 1.38E+03 5.37E+04 1.33E+04 1134 5.85E+04 1.33E-10 1135 5.1E+03 6.09E+01 4.88E+04 6.46E+02 42 Reflects core inventory without 1.7 peaking factor or pool DFs applied.

NOC-AE-07002127 Attachment 1 Page 141 of 219 4.4 FUEL HANDLING ACCIDENT 4.4.4 Radiological Releases The analysis assumes 23 feet of water above damaged fuel. This value corresponds to the minimum depth of water coverage over the top of irradiated fuel assemblies seated in the spent fuel pool racks as required by TS 3/4.9.11, Water Level - Storage Pools Spent Fuel Pool.

Twenty-three feet of water is also assumed for an assembly drop in the core. TS 3/4.9.10, "Water Level - Refueling Cavity", requires maintaining at least 23 feet of water above the top of the reactor vessel flange during movement of irradiated fuel assemblies within containment. This assumption is consistent with Regulatory Guide 1.183. Due to the submergence of the damaged fuel, the iodine release is assumed to experience a DF of 200 per Regulatory Guide 1.183. The assumed iodine chemical form after decontamination by the water pool is 43% organic and 57%

elemental. No DF is applied to the noble gas. As previously noted, the DF for particulates is assumed to be infinite.

Releases from the Fuel Handling Building are vented to the atmosphere via the Plant Vent. The RCB purge is also from the same Plant Vent. Therefore, for the FHA releases, the Plant-Vent-to-Control-Room X/Q is used. Releases from the RCB Personnel Airlock are also exhausted via this Plant Vent. The Plant-Vent-to-Control-Room x/Q also bounds a release from the RCB Equipment Hatch opening since the Plant Vent is much closer to the Control Room air intake than the Equipment Hatch.

4.4.5 Assumptions and Inputs Assumptions and inputs utilized in the analysis are:

1. The bounding core inventory is based on a DBA power level of 4100 MWth compared to the Rated Thermal Power (RTP) level of 3853 MWth with a 0.6% measurement uncertainty.

2. The release consists of the gap activity in the 264 fuel pins in the dropped assembly and 50 pins in an impacted fuel assembly, for a total of 314 fuel pins. Since there are 193 fuel assemblies in the core, there are 50,952 fuel pins in a core.

3. The dropped assembly and the impacted assembly are assumed to have peaking factors of 1.7.

4. A water depth above the damaged fuel of 23 feet is the limiting case.

5. The activity is assumed to be released directly to the Control Room HVAC intake from the Plant Vent (using the Plant Vent to Control Room X/Q). The Control Room internal air is assumed to be in equilibrium with the air outside the Control Room HVAC intake.

Therefore, the Control Room is not assumed to be pressurized during the accident, nor are any assumptions made as to the functioning of the Control Room HVAC systems.

Input parameters used for the FHA analysis are given in Table 4.4-2. Conformance with Regulatory Guide 1.183 guidance addressing FHA analysis is provided in Attachment 6, Tables A and C.

NOC-AE-07002127 Attachment 1 Page 142 of 219 4.4 FUEL HANDLING ACCIDENT Table 4.4-2 Fuel Handling Accident Inputs Input/Assumption CLB AST Previous Reactor Power 4100 MWt Fuel Decay Period 42 hrs Radial Peaking Factor 1.7 Release Fractions Noble Gases (except Kr-85) 10% 5%

Kr-85 30% 10%

Iodines (except 1-131) 10% 5%

1-131 12% 8%

Number of Failed Rods 314 of 50952 (Equivalent Assemblies)

Minimum water depth over 23 feet damaged fuel Pool Water Iodine 100 200 Decontamination Factor Release Period 2hr Instantaneous Release Location Plant Vent Credit for Filtration on Accident in FHB: Yes Release Accident in RCB: No No Credit for Control Room Yes No Filtration Dose Conversion Factors Table 4.2-6 Decay Constants and Decay Daughter Fractions Table 4.2-7 Offsite X/Q's Table 4.1-24 Offsite breathing rates Table 4.2-1 Control Room and TSC x/Q's Table 4.1-37

NOC-AE-07002127 Attachment 1 Page 143 of 219 4.4 FUEL HANDLING A CCIDENT 4.4.6 Summary and Conclusions The radiological consequences of the design-basis refueling accident were analyzed using the simplified and conservative assumptions described above. A spreadsheet calculation was carried out to obtain the results for 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> of decay and the results are presented in Table 4.4-3. The spreadsheet results were verified with STARDOSE code. The dose agreement between the spreadsheet and STARDOSE was excellent.

Control Room and EAB doses are explicitly calculated. Because the release occurs within two hours, the 0-2 hour EAB dose is bounding for the 0-8 hour LPZ dose. The Control Room dose bounds that for the TSC since the TSC uses the same air intake as the Control Room, and the TSC has a smaller volume. A smaller finite volume correction factor gives a smaller immersion dose.

Table 4.4-3 Fuel Handling Accident Dose Results (rem TEDE)

Dose Receptor (0-2 hours) Limit43 EAB 0.83 6.3 LPZ 0.30 6.3 Control Room 3.39 5 TSC 3.39 5 Radiological doses to a Control Room operator and a person located at the EAB or LPZ resulting from a design basis FHA are less than the regulatory dose limits given in 10CFR50.67.

4.4.7 Core Alterations CORE ALTERATIONS are defined as the movement of any fuel, sources, or reactivity control components (excluding rod cluster control assemblies locked out in the integrated head package) within the reactor vessel with the reactor head removed and fuel in the vessel. As described in TSTF-51, Revision 2, accidents postulated to occur during core alterations include inadvertent criticality (due to control rod removal error or continuous rod withdrawal error during refueling or boron dilution), fuel handling accident, and the loading of a fuel assembly or control component in an incorrect location. Generically, it was concluded that of these off-normal 41 10CFR50.67 for offsite and 10CFR50.67, as modified by Regulatory Guide 1.183 in Table 6, page 1.183-20, for the Control Room and TSC.

NOC-AE-07002127 Attachment 1 Page 144 of 219 4.4 FUEL HANDLING ACCIDENT occurrences, only the fuel handling accident results in cladding damage and potential radiological release. Consequently, it is being proposed that the APPLICABLE MODE "during core alterations" be deleted from TS 3/4.3.2, Table 3.3-3, Functional Unit 3.b.4), TS 3/4.3.2, Table 4.3-2, Functional Unit 3.b.4), and TS 3/4.7.7. Functional Unit 3.b.4) is the Containment Ventilation Isolation RCB Purge Radioactivity instrument. In addition, the ACTION to "immediately suspend core alterations" if the Limiting Condition for Operations is not met is deleted from TS 3/4.3.2, Table 3.3-3, Functional Unit 10.d and from TS 3.7.7, Modes 5 and 6.

The affected system by TS 3/4.7.7 is the Control Room Makeup and Cleanup Filtration System.

TS Limiting Conditions for Operation (LCO) requirements remain unaffected for other Technical Specifications that are needed to prevent or mitigate CORE ALTERATION events other than the fuel handling accident. This includes Technical Specifications such as the required boron concentration for refueling operations (Specification 3/4.9.1), and the required nuclear instrumentation for refueling operations (Specification 3/4.9.2).

The LCO APPLICABILITY requirements for operations with a potential for draining the reactor vessel are unaffected by the proposed changes. Also, APPLICABILITY requirements are unaffected for decay heat removal systems during shutdown condition specifications, and for specifications that require maintenance of high water levels over irradiated fuel.

4.4.8 Shutdown Safety Assessment/Defense-in-Depth In previous amendments for similar relaxations at other facilities, the NRC staff requested that licensees make appropriate commitments to implement administrative controls to facilitate restoration of containment or fuel building closure, and to provide a filtered and monitored release path as a defense-in-depth measure to mitigate the consequences of a postulated FHA.

TSTF-51, Revision 2, requires licensees that incorporate this generic change to commit to NUMARC 93-01, Revision 3, Section 11.2.6, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants", subheading "Containment - Primary (PWR)/Secondary (BWR)". The commitment in TSTF-5 1, Revision 2, is based on a draft version of NUMARC 93-01, Revision 3. When NUMARC 93-01, Revision 3, was approved in July 2000, the guidelines referred to in TSTF-51, Revision 2, were designated as Section 3.6.5.

Section 3.6.5 of NUMARC 93-01 states:

"...for plants which obtain license amendments to utilize shutdown safety administrativecontrols in lieu of Technical Specification requirements on primary or shutdown containmentoperability and ventilation system operabilityduringfuel handling or core alterations,the following guidelines should be included in the assessment of systems removedfrom service:

Duringfuel handling/corealterations,ventilation system and radiationmonitor availability (as defined in NUMARC 91-06) should be assessed, with respect to filtration and monitoring of releasesfrom the fuel. Following shutdown, radioactivity in the RCS decaysfairly rapidly. The basis of the Technical Specification operability amendment is the reduction in doses due to such decay. The goal of maintainingventilation system and

NOC-AE-07002127 Attachment 1 Page 145 of 219 4.4 FUEL HANDLING ACCIDENT radiation monitor availabilityis to reduce doses even further below thatprovided by the naturaldecay, and to avoid unmonitoredreleases.

A single normalor contingency method to promptly closeprimary or secondary containmentpenetrationsshould be developed. Such prompt methods need not completely block the penetrationor be capable of resistingpressure. The purpose is to enable ventilation systems to draw the releasefrom a postulatedfuel handling accidentin the proper direction such that is can be treatedand monitored.

The NUMARC 93-01 guidance is built upon two basic premises: avoiding unmonitored releases and using available (although not necessarily "Technical Specification OPERABLE") filtration capabilities to reduce doses below those achieved from the decay of the source term and the scrubbing of the water.

STP License amendments 69 and 139 for Unit 1 and 58 and 128 for Unit 2 were approved based on the premise that administrative controls will be in place to shut a Personnel Airlock Door and the Equipment Hatch in the Reactor Containment Building in the event of a fuel handling accident. Administrative controls will be in place to close penetrations providing direct access from the containment atmosphere to the outside atmosphere and entrances to the Fuel Handling Building in the event of a FHA when integrity of this building is not required. These controls do not need to result in completely blocking the penetration or being capable of resisting pressure.

To support the purpose stated in the preceding paragraph, additional administrative controls will be in place for controlling the removal from service of ventilation filtration and radiation monitoring systems. These controls will be in place so that ventilation filtration and radiation monitoring remains "available" (not necessarily OPERABLE) in the Containment, the Control Room and Fuel Handling Buildings whenever handling irradiated fuel or loads over irradiated fuel to ensure that the release is treated and monitored. If for any reason the ventilation requirements can not be met, fuel movement within the affected building shall be discontinued until the flow path(s) become available. Attachment 4 provides a description of the planned changes to the STP Technical Requirements Manual to close containment penetrations and to maintain ventilation systems available so that releases from a FHA can be treated and monitored. provides a List of Commitments to maintain these systems available.

NOC-AE-07002127 Attachment 1 Page 146 of 219 4.5 MAIN STEAM LINE BREAK 4.5 Main Steam Line Break Radiological Assessment 4.5.1 Methodology Overview The Main Steam Line Break (MSLB) accident is postulated as a break of one of the large steam lines outside the containment leading from a SG. For the three intact SGs loops, primary-to-secondary coolant leakage transfers activity into the secondary coolant. This makes it available for release into the environment via steaming through the SG PORV. For the coolant loop with the broken steam line (i.e., faulted SG), primary-to-secondary coolant leakage is assumed to be released from the RCS directly into the environment without passing through any secondary coolant. This is due to assumed "dry-out" conditions in the faulted SG.

Consistent with Regulatory Guide 1.183, two reactor transients that maximize the radioactivity available for release were modeled.

Pre-accident Iodine Spike A pre-accident iodine spike raises the primary coolant iodine concentration to the Technical Specification maximum 60 gCi/gm assumed DE 1-131 value at full power operations. It is assumed that all of the spike activity is homogeneously mixed in the primary coolant, prior to accident initiation.

Note that the equilibriumf' secondary coolant system iodine activity must also be evaluated. This consists of the 0.1 pCi/gm DE 1-131 equilibrium secondary coolant activity concentration, as allowed by the TS. This activity is used to determine the dose contribution that results from the initial blowdown of all fluid in the faulted SG, and the SG PORV release of secondary coolant through the intact SGs.

In addition, the MSLB analysis was performed to determine the release path through the above seat main steam line orifices. This steam release rate is assumed to be 1.93 Ibm/second per SG orifice and conservatively continues for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Accident-Initiated Concurrent Iodine Spike It is assumed that the MSLB event causes a primary reactor system transient concurrent with the release of fluid from the primary and secondary coolant systems. This transient, in turn, is associated with an iodine spike which assumes that the iodine release rate from the fuel rods to the primary coolant increases to a value 500 times greater than the release rate corresponding to the 1.0 tCi/gmn DE 1-131 RCS equilibrium iodine concentration. The elemental and particulate iodines release rate spike is assumed to occur for a duration of eight hours. Since no partitioning is assumed for the organic iodines, they are released, along with the noble gases, as an instantaneous release.

NOC-AE-07002127.

Attachment I Page 147 of 219 4.5 MAIN STEAM LINE BREAK As described for the Pre-accident Iodine Spike case above, the dose due to the equilibrium secondary coolant system iodine activity (0.1 jtCi/grn DE 1-131) must also be determined. The release path through the above seat main steam line orifices is also modeled.

4.5.1.1 Comparison of Modeling with Regulatory Guide 1.183, Appendix E Regulatory Guide 1.183 Appendix E, Position 5.5.1:

A portion of the primary-to-secondaryleakage willflash to vapor, based on the thermodynamic conditions in the reactorand secondary coolant.

" Duringperiods of steam generatordryout, all of the primary-to-secondary leakage is assumed to flash to vapor and be releasedto the environment with no mitigation.

" With regardto the unaffected steam generatorsusedforplantcooldown, the primary-to-secondaryleakage can be assumed to mix with the secondary water withoutflashing duringperiods of total tube submergence.

Treatment for the MSLB analysis:

In the faulted SG, all of the primary-to-secondary leakage is assumed to flash to vapor and be released to the environment with no mitigation. In the intact SGs, the primary-to-secondary leakage is assumed to mix with the secondary water without flashing. The SG tubes in the intact SGs are assumed to not be uncovered during the accident.

Regulatory Guide 1.183 Appendix E, Position 5.5.2:

The leakage that immediatelyflashes to vapor will rise through the bulk water of the steam generatorand enter the steam space. Credit may be taken for scrubbing in the generator,using the models in NUREG-0409, "Iodine Behavior in a PWR Cooling System Following a PostulatedSteam GeneratorTube Rupture Accident" (Ref E-2), duringperiods of total submergence of the tubes.

Treatment for the MSLB analysis:

This assumption is not used. It is assumed that the primary-to-secondary leakage does not flash in the intact SGs. The nuclides in the primary-to-secondary leakage are added to the bulk fluid in the intact SGs.

Regulatory Guide 1.183 Appendix E, Position 5.5.3:

The leakage that does not immediatelyflash is assumed to mix with the bulk water.

Treatment for the MSLB analysis:

It is assumed that the primary-to-secondary leakage does not flash in the intact SGs. The nuclides in the primary-to-secondary leakage are added to the bulk fluid in the intact SGs.

NOC-AE-07002127 Attachment 1 Page 148 of 219 4.5 MAIN STEAM LINE BREAK Regulatory Guide 1.183 Appendix E, Position 5.5.4:

The radioactivityin the bulk water is assumed to become vapor at a rate that is the function of the steaming rate and the partition coefficient. A partitioncoefficient for iodine of 100 may be assumed. The retention ofparticulateradionuclidesin the steam generatorsis limited by the moisture carryoverfrom the steam generators.

Treatment for the MSLB analysis:

A partition coefficient of 100 is assumed for elemental and particulate iodines released from the intact steam generators. Organic iodine is not partitioned. Organic iodine is assumed to migrate directly to the steam space and become immediately available for release.

Regulatory Guide 1.183 Appendix E, Position 5.6:

Operating experience and analyses have shown thatfor some steam generatordesigns, tube uncovery may occurfor a short periodfollowing any reactortrip (Ref E-3). The potential impact of tube uncovery on the transportmodel parameters (e.g., flashfraction, scrubbingcredit) needs to be considered.The impact of emergency operatingprocedure restorationstrategies on steam generatorwater levels should be evaluated.

Treatment for the MSLB analysis:

Tube uncovery does not occur in the intact SGs following this event and the subsequent reactor trip.

4.5.2 Analytical Model The RADTRAD computer code is used to determine the MSLB accident doses at the EAB, LPZ, and Control Room, consistent with Regulatory Guide 1.183. For each spiking scenario, two models were designed for two different release paths, i.e., the intact SG PORVs and the broken steam line. (Note that all releases from the PORVs of different SGs and from a broken steam line are postulated to occur at the location of the PORV closest to the Control Room HVAC intake.)

Following a main steam line break, auxiliary feedwater to the faulted loop is isolated and the steam generator is allowed to steam dry. Thus, radionuclides carried from the primary coolant to the faulted steam generator via leaking tubes are assumed to be released directly to the environment. Radionuclides released from the generators in the intact loops are assumed to be mixed with the secondary coolant and partitioned between the generator liquid and steam before releasing to the environment. The intact steam generator iodine partition coefficient (PC) is 100.

The iodine partition is modeled using a reduced release flow rate. The steam release to the environment through relief valves is assumed to last for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The steam release through the above seat valve is conservatively assumed to last for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. For the radiological evaluation of the postulated MSLB, the following two scenarios were considered:

NOC-AE-07002127 Attachment 1 Page 149 of 219 4.5 MAIN STEAM LINE BREAK

1) A pre-existing iodine spike has raised the concentration in the RCS to 60 pCi/gm DEI 131.

2) An accident-induced iodine spike which increases the release rate to the RCS to a value 500 times greater than the release rate corresponding to an RCS iodine concentration of 1 pCi/gm DEI 131.

The CLB uses the ICRP-30 dose conversion factors. For the application of AST and TEDE dose criteria, dose conversion factors from Federal Guidance Reports 11 and 12 are used. A schematic of the analytical model is provided in Figure 4.5-1.

Iodine Spike Release Model:

The reactor trip and the primary system depressurization associated with the MSLB create an iodine spike in the primary system. The increase in primary coolant iodine concentration is estimated using a spike model which assumes that the iodine release rate from the fuel rods to the primary coolant increases to a value 500 times greater than the release rate corresponding to the iodine concentration of 1 pCi/gm dose equivalent 1-131 in the RCS. The release rate is calculated using the following equation.

= N{A* + JL?+LRJ Pi = Production Rate for Nuclide i (pCi/grn-sec)

ýi = Radioactive Decay Constant for Nuclide i (sec-1)

Ni = Concentration of Nuclide i (pCi/gm) fL = Letdown Flow (gm/sec)

MRcs RCS volume (gin) 1 = Letdown Demineralizer Efficiency/100 (unitless)

LR = Rate of Reactor Coolant System Identified and unidentified Leakage (as allowed by plant Technical Specifications). (gm/sec)

This is the same modeling technique as used in the Current Licensing Basis, Per Appendix E (Section 2.2) of Regulatory Guide 1.183, the assumed iodine spike duration is 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. For conservatism and simplicity of RADTRAD modeling, the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of total radioactivity release is assumed to be instantaneously released at the beginning of the event. The spike is assumed to also increase the RCS concentration of Alkali metals (Cs and Rb).

NOC-AE-07002127 Attachment 1 Page 150 of 219 4.5 MAIN STEAM LINE BREAK Figure 4.5-1 MSLB RADTRAD Model 1

RCS 2 3 SG - Intact SG - Faulted 4

Environment 5 Recir Control Room Filtration Exhaust to Env

NOC-AE-07002127 Attachment 1 Page 151 of 219 4.5 MAIN STEAM LINE BREAK 4.5.3 Radiological Source Term For this analysis only the iodine and noble gas activities, which are conservatively characterized by operation with 1% core fuel defects and the equilibrium and spiked release rates from that fuel, define the source terms. RADTRAD uses these activities, in curies per megawatt, and then applies nuclide release fractions and a specified core power to calculate the source term for a given case. The AST release fractions associated with iodines and noble gases are assumed to be 100%, and are released to the reactor coolant system.

No additional fuel damage is assumed due to this accident. Two different cases of iodine spiking are analyzed, in accordance with regulatory guidance as previously described.

4.5.3.1 Reactor Coolant System Source Term 4.5.3.1.1 RCS Iodine Concentrations Table 4.2-14 shows the calculation for the Reactor Coolant System (RCS) iodine concentration, based on thyroid DCFs, for 1% failed fuel. Table 4.2-17 shows the calculation for the RCS iodine concentration, based on Thyroid DCFs, for a Pre-existing Iodine Spike.

For the accident-induced iodine spike, the iodine release rates corresponding to a RCS concentration of 1 ýtCi/gm are calculated using methodology described in Section 4.5.2. The release rates are then multiplied by the RCS mass and a factor of 500 to yield a release rate in units of Ci/minute. For conservatism, the modeling assumes that the total iodines released from the gap during the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period are instantaneously released to the RCS (i.e. puff release) following the initiation of the event. Table 4.5-1 shows the total iodine spike activity.

Table 4.5-1 RCS Iodine Inventory Due to Accident-Induced Spike (500x Release Rate)

CLB AST Isotope (Ci) (Ci)  % Difference 1-131 1.86E+05 1.73E+05 -6.7%

1-132 4.82E+05 5.62E+05 16.5%

1-133 3.31E+05 3.23E+05 -2.4%

1-134 1.84E+05 2.35E+05 27.9%

1-135 2.36E+05 1.12E+06 375.3%

NOC-AE-07002 127 Attachment 1 Page 152 of 219 4.5 MAIN STEAM LINE BREAK 4.5.3.1.2 RCS Noble Gas Concentrations Table 4.2-14 shows the calculation for the RCS noble gas concentration for 1% failed fuel.

However, Kr-89 and Xe-137 were not used in the AST analysis.

4.5.3.1.3 RCS Cesium and Rubidium Concentrations Iodine spikes are conservatively assumed to cause an increase in Cesium and Rubidium activities, along with the increase in iodine concentrations. Table 4.2-18 shows the total activities released from the pre-accident spike. Table 4.5'2 shows the activities for an accident-induced spike. For the MSLB, the spike is modeled as an instantaneous release at time 0 of the total number of curies that would be released into the RCS over an 8-hour period.

Table 4.5-2 Total RCS Cs and Rb Activity for an Accident-Induced Iodine Spike (Ci)

Isotope CLB AST Rb-86 N/A 2.06E+03 Rb-88 N/A 5.07E+06

,Rb-89 N/A 2.65E+05 Cs-134 N/A 1.68E÷05 Cs-136 N/A 3.04E+05 Cs-137 N/A 1.32E+05 4.5.3.2 Secondary System Source Terms 4.5.3.2.1 Secondary System Iodine Concentrations The secondary systems iodine concentrations corresponding to the Technical Specification limit of 0.10 pCi/gm are given in Table 4.2-19.

4.5.3.2.2 Secondary System Noble Gas Concentrations The secondary systems noble gas concentrations corresponding to 1.0% failed fuel are given in Table 4.2-20.

NOC-AE-07002127 Attachment 1 Page 153 of 219 4.5 MAIN STEAM LINE BREAK 4.5.4 Radiological Releases The activity release model is consistent with the model given on Figure E-1 of Regulatory Guide 1.183. Activity that originates in the RCS is released to the secondary coolant by means of the primary-to- secondary coolant leak rate. This design basis leak rate value is 0.65 gpm from the three intact SGs, and 0.35 gpm for the faulted SG with the broken steam line.

Primary-to-secondary coolant leakage through the faulted SG conservatively goes directly to the environment, without mixing with any secondary coolant. Therefore, under the assumed dry-out conditions, no partitioning of any nuclides is expected to occur in this release pathway.

For all post-accident releases through the PORVs of the intact SG loops, the mechanism for release to the environment is steaming of the secondary coolant. Because of this release dynamic, Regulatory Guide 1.183 allows for a reduction in the amount of activity released to the environment based on partitioning of nuclides between the liquid and gas states of water. For iodine, Regulatory Guide 1.183 allows a partition coefficient of 100 for all iodines. However, organic iodines are assumed to be released directly to the environment. Reviewing the specified AST release fractions, it is concluded that the only nuclides other than iodines to be released from the core source term are noble gas nuclides. Because of their volatility, 100% of the noble gases are assumed to be released.

The methodology used to model steaming of activity through PORVs following the postulated MSLB event assumes an average cumulative release rate through the SG PORVs. The partition factors are applied to these release rates. This data was then converted using the assumption of cooled liquid conditions (i.e., 62.4 lbm/ft3 ), as specified by the applicable guidance of Regulatory Guide 1.183. The steaming release and primary-to-secondary coolant leakage is postulated to end at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when the RCS and secondary loop have reached equilibrium. Steam release through the above seatmain steam line orifices is assumed to conservatively continue for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Per Appendix E of Regulatory Guide 1.183, the chemical form of radionuclide released from the fuel is assumed to be 95% cesium iodine (CsI), 4.85% elemental iodine, and 0.15% organic iodine. This analysis assumes that iodine released from the steam generators to the environment is 4.2% elemental, 13.1% organic, and 82.7% particulate (see Section 4.2.5).

Different forms of radionuclide have different transport behaviors. Based on Regulatory Guide 1.183, the particulate form of radioiodine (CsI) is not released from the steam generator to the environment. However, for conservatism, particulate iodine released from the intact steam generators is assumed to have the same partition coefficient as elemental iodine. Also, RCS leakage to the faulted steam generator is assumed to be directly released to the environment. This release includes cesium and rubidium particulates and all chemical forms of radioiodine.

All releases from the SG PORVs (i.e., from the intact SGs) and the faulted SG are considered ground releases.

NOC-AE-07002127 Attachment 1 Page 154 of 219 4.5 MAIN STEAM LINE BREAK 4.5.5 Assumptions and Inputs The following inputs and assumptions were used in the MSLB analysis.

1. The source term is based upon a power level of 4100 MW thermal, 5 w/o enrichment, and a three region core with equilibrium cycle core at end of life. The three regions have operated at a specific power of 39.3 MW/MTU for 509, 1018, and 1527 EFPD, respectively. The assumed power level is greater than the Rated Thermal Power of 3853 MWth plus a 0.6% measurement uncertainty.

2. The equilibrium secondary activity before the accident is based upon a pre-accident primary-to-secondary leakage of 1 gpm. This is conservative since the Technical Specifications limit the pre-accident leakage to 150 gpd per steam generator or 600 gpd (0.42 gpm) total. The secondary coolant activityis based on 0.1 ýICi/gm of dose equivalent 1-131. Noble gas activity in the secondary coolant is based on 1% failed fuel.

3. Primary-to-secondary leakage through the steam generator tubes prior to the accident and during the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the transient is 1 gpm. Eight hours after the accident, the residual heat removal system starts and primary-to-secondary leakage is stopped. Primary-to-secondary leakage is modeled as 0.65 gpm for the three intact steam generators and at 0.35 gpm for the faulted steam generator.

4. No fuel failures are assumed to be caused by the main steam line break.

5. For a pre-accident iodine spike, the activity in the reactor coolant is based upon an iodine spike which has raised the reactor coolant concentration to 60 pCi/gm of dose equivalent 1-131. Noble gas activity is based on 1% failed fuel.

6. For an accident-induced iodine spike, the accident initiates an iodine spike in the RCS which increases the iodine release rate from the fuel to a value 500 times greater than the release rate corresponding to a RCS concentration of 1 ýtCi/gm dose equivalent I-131. Iodine is assumed to be released at this rate for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> into the RCS. The iodine activity released from the fuel to the RCS is conservatively assumed to mix instantaneously and uniformly in the RCS. The accident-induced spike is modeled as an instantaneous release at t=0 of the 0-8 hour integrated iodine release.

Since Regulatory Guide 1.183 specifies that the chemical form of particulate iodine is cesium iodide (CsI), the spike is also assumed to increase the Alkali metal (Cs and Rb) in the RCS in relative amounts. Noble gas activity is conservatively based on 1% failed fuel.

7. Following the rupture, auxiliary feedwater to the faulted loop is isolated and the steam generator is allowed to steam dry. Thus, the iodine partition factor for the faulted steam generator is 1.

8. The activity released from the fuel from the gap is assumed to be instantaneously mixed with the reactor coolant within the pressure vessel per Regulatory Guide 1.183.

NOC-AE-07002127 Attachment 1 Page 155 of 219 4.5 MAIN STEAM LINE BREAK

9. Tube uncovery does not occur in the three intact SGs. Primary-to-secondary leakage in these SGs is added to the bulk fluid in the SGs and does not flash directly to the environment.

10. A partition coefficient of 100 is assumed for elemental iodine released from the intact steam generators (Regulatory Guide 1.183, Appendix E, Section 5.5.4). Organic iodine is not partitioned. Organic iodine is assumed to migrate directly to the steam space and become immediately available for release.

11. Similar to the CLB, operator action is taken to isolate the faulted SG within 30 minutes of the event. The total release from the faulted SG is 214,000 Ibm initially plus a subsequent release of 385,000 lbm from the Main Feedwater System'and the Auxiliary Feedwater System, for a total of 599,000 lbm.

12. Steam releases from the faulted and the intact SGs are assumed to occur at a constant rate for the time period of interest.

13. Eight hours after the accident, the residual heat removal system is in operation and no further steam containing radionuclides is released from steam generators to the environment except the leakage through the MSIV above seat orifices. The release from the orifices continues until 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after the start of the accident. This is conservative since all releases would terminate in less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when the RHR system is in operation.

14. The break and the above-seat drain releases occur in the Isolation Valve Cubicle next to the PORVs. Therefore, the PORV-to-Control Room x/Qs are used for the Control Room and TSC dose analyses.

15. Offsite Power is lost. The condensers are unavailable for steam dump.

16. The Control Room ventilation system is assumed to automatically transfer to the emergency mode of operation after the initiation of safety injection.

17. All activity is released to the environment with no consideration given to cloud depletion by ground deposition during transport to the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ).

18. Reactor coolant density is 8.33 lbs/gal.

Input parameters used for the MSLB analysis are given in Table 4.5-3.

NOC-AE-07002127 Attachment 1 Page 156 of 219 4.5 MAIN STEAM LINE BREAK Table 4.5-3 Inputs for MSLB Analysis Parameter CLB AST Core power (for radiological source terms) 4100MWt Core power (for steam releases) 3876 MWt (3853MWt + 0.6%)

RCS density 8.33 lbm/gallon RCS Mass 44 2.658E+8 gm SG Node Volume Intact 5.937E+4 gal 5.937E+4 gal 7.94E+3 ft3 Faulted 1.979E+4 gal 1.979E+4 gal3 2.65E+3 ft Primary-to-Secondary Leakage Intact 0.65 gpm Faulted 0.35 gpm Release from Intact SGs 0-2 hours 452,000 lbm 4

2-8 hours 1,080,000 Ibm Release from Faulted SG 5 0-0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 599,000 Ibm Release from Above (MSIV) Seat Drains Intact SGs 5.79 lbm/sec Faulted SG 1.93 Ibm/sec Iodine appearance rate into the RCS for the 500 accident-induced spike Iodine Species Released from the RCS (%) 91/4/5 4.85/0.15/95 (elemental/organic/particulate)

Iodine Species Released from flashed RCS primary-to-secondary leakage flow to the 91/4/5 4.85/0.15/95 environment (%) (elemental/organic/particulate)

Iodine Partition Factors for Releases from the Secondary Side 100/100/100 100/1/100 (elemental/organic/particulate)

For the CLB analyses, a high value of 2.658E+8 gm was used to maximize the total activity in the RCS and a smaller value of 2.6E+8 was used to determine the RCS volume.

45 The total release from the faulted SG is 214,000 lbm initially plus a subsequent release of 385,000 Ibm from the Main Feedwater System and the Auxiliary Feedwater System, for a total of 599,000 lbm. Operator action is taken to isolate feedwater to the faulted SG within 30 minutes of the event.

NOC-AE-07002127 Attachment 1 Page 157 of 219 4.5 MAIN STEAM LINE BREAK Table 4.5-3 Inputs for MSLB Analysis Parameter CLB AST Resulting Iodine Species Released from the Secondary Side to Environment 91/4/5 4.2/13.1/82.746

% (elemental/organic/particulate)

Steam Flow rate 1.574E+7 lbmihr Dose Conversion Factors Table 4.2-6 Decay Constants and Decay Daughter Fractions Table 4.2-7 Offsite breathing rates Table 4.2-1 Offsite X/Q's Table 4.1-24 Control Room HVAC Parameters Table 4.2-3 Control Room HVAC Flow Rates Table 4.2-2 TSC HVAC Parameters Table 4.2-5 TSC HVAC Flow Rates Table 4.2-4 Control Room and TSC X/Q's Table 4.1-37 46 See Section 4.2.5

NOC-AE-07002127 Attachment 1 Page 158 of 219 4.5 MAIN STEAM LINE BREAK 4.5.6 Summary and Conclusions Table 4.5-4 provides the results from the analyses.

Table 4.5-4 MSLB Dose Results (rem TEDE)

Pre-Existing Iodine Spike Accident-induced Iodine Spike Receptor Result Limit Result Limit I

EAB (worst 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) 0.052 254/ 0.85 2.548 LPZ 0.041 2547 0.66 2.548 Control Room 0.109 547 1.70 547 TSC 0.106 5 1.65 5 All doses are well below their respective acceptance criteria, so it is verified that this design basis MSLB accident is sufficiently mitigated.

10CFR50.67 48 10CFR50.67 as modified by Regulatory Guide 1.183 in Table 6 on Page 1.183-20.

NOC-AE-07002127 Attachment 1 Page 159 of 219 4.6 STEAM GENERA TOR TUBE R UPTURE 4.6 Steam Generator Tube Rupture Radiological Assessment 4.6.1 Methodology Overview The Steam Generator Tube Rupture (SGTR) accident is postulated as a complete severance of a single SG tube. This is a conservative assumption because tube material is a highly ductile metal alloy, and the most probable mode of failure would be one or more minor tube leaks of varying sizes and undetermined origin.

The tube rupture results in the release of radioactive material from the containment. For the three intact SGs, primary-to-secondary coolant leakage continues to transfer activity into the secondary coolant side. This makes it available for release into the-environment via steaming through the SG PORVs. For the SG with the ruptured tube, referred to as the ruptured SG, coolant release will take two forms:

1. Break Flow - un-flashed release of RCS coolant directly into the secondary loop, and made available for steaming release to the environment through the PORV.

2. Flashed Break Flow - RCS coolant that flashes directly to steam when released from the ruptured tube, and is sent through the PORV to the environment.

Operators are assumed to identify the ruptured steam generator and attempt to close the PORV on the ruptured steam generator in 10 minutes. However, the PORV is assumed to fail open (the single failure for this accident scenario) at that time. It is assumed that the failed PORV is isolated by manually closing the PORV block valve within 15 minutes of the PORV failure.

Therefore, the steam release via the ruptured steam generator's PORV is assumed to continue for a total of 25 minutes. These assumptions are consistent with the current licensing basis.

Consistent with Regulatory Guide 1.183, two reactor transients (i.e., cases) that maximize the radioactivity available for release were modeled.

Pre-accident Iodine Spike A pre-accident iodine spike raises the primary coolant iodine concentration to the Technical Specification maximum 60 ptCi/gm assumed DE 1-131 value at full power operations. It is assumed that all of the spike activity is homogeneously mixed in the primary coolant, prior to accident initiation.

The equilibrium secondary coolant system iodine activity must also be evaluated. The total activity available for release from both the intact SGs and ruptured SG is the Technical Specification limit of 0.10 gCi/gm dose equivalent of 1-131.

NOC-AE-07002127 Attachment 1 Page 160 of 219 4.6 STEAM GENERA TOR TUBE R UPTURE Accident-Initiated Concurrent Iodine Spike It is assumed that the SGTR event causes a primary reactor system transient concurrently with the release of fluid from the primary and secondary coolant systems. This transient, in turn, is associated with an iodine spike which assumes that the iodine release rate from the fuel rods to the primary coolant increases to a value 335 times greater than the release rate corresponding to the 1.0 pCi/gm DE 1-131 RCS equilibrium iodine concentration. The elemental and particulate iodines release rate spike is assumed to occur for a duration of eight hours. Since no partitioning is assumed for the organic iodines, they are released, along with the noble gases, as an instantaneous release.

The doses due to the equilibrium secondary coolant system iodine activity (0.1 pCi/gm DE I-131) and the release path through the above seat main steam line orifices are also determined.

4.6.1.1 Comparison of Modeling with Regulatory Guide 1.183, Appendix E Regulatory Guide 1.183 Appendix E, Position 5.5.1:

A portion of the primary-to-secondaryleakage will flash to vapor, based on the thermodynamic conditions in the reactorand secondary coolant.

" Duringperiods ofsteam generatordryout, all of the primary-to-secondary leakage is assumed toflash to vapor and be released to the environment with no mitigation.

• With regardto the unaffected steam generators usedforplantcooldown, the primary-to-secondaryleakage,can be assumed to mix with the secondary water withoutflashing duringperiods of total tube submergence.

Treatment for the SGTR analysis:

In the ruptured SG, the portion of the break flow that is assumed to flash to vapor is released to the environment with no mitigation. The unflashed portion and the primary-to-secondary leakage are assumed to mix with the secondary water without flashing. In the intact SGs, the primary-to-secondary leakage is assumed to mix with the secondary water without flashing. The SG tubes in the intact SGs are assumed to not be uncovered during the accident.

Regulatory Guide 1.183 Appendix E, Position 5.5.2:

The leakage that immediatelyflashes to vapor will rise through the bulk water of the steam generatorand enter the steam space. Credit may be taken for scrubbing in the generator,using the models in NUREG-0409, "Iodine Behavior in a PWR Cooling System Following a'PostulatedSteam GeneratorTube Rupture Accident" (Ref E-2), duringperiods of total submergence of the tubes.

NOC-AE-07002127 Attachment I Page 161 of 219, 4.6 STEAM GENERA TOR TUBE R UP TURE Treatment for the SGTR analysis:

This assumption is not used. It is assumed that the primary-to-secondary leakage does not flash in the intact SGs. The nuclides in the primary-to-secondary leakage are added to the bulk fluid in the intact SGs.

Regulatory Guide 1.183 Appendix E, Position 5.5.3:

The leakage that does not immediatelyflash is assumed to mix with the bulk water.

Treatment for the MSLB analysis:

In the ruptured SG, the unflashed portion and the primary-to-secondary leakage are assumed to mix with the bulk SG water without flashing. In the intact SGs, the primary-to-secondary leakage is assumed to mix with the bulk SG water without flashing.

Regulatory Guide 1.183 Appendix E, Position 5.5.4:

The radioactivityin the bulk water is assumed to become vapor at a rate that is the function of the steaming rate and the partition coefficient. A partition coefficient for iodine of 100 may be assumed. The retention ofparticulateradionuclidesin the steam generators is limited by the moisture carryoverfrom the steam generators.

Treatment for the SGTR analysis:

A partition coefficient of 100 is assumed for elemental iodine released from the intact steam generators. Organic iodine is not partitioned. Organic iodine is assumed to migrate directly to the steam space and become immediately available for release.

Regulatory Guide 1.183 Appendix E, Position 5.6:

Operatingexperience and analyses have shown thatfor some steam generatordesigns, tube uncovery may occurfor a shortperiodfollowing any reactortrip (Ref E-3). The potential impact of tube uncovery on the transportmodel parameters(e.g., flashfraction, scrubbingcredit) needs to be considered. The impact of emergency operatingprocedure restorationstrategies on steam generatorwater levels should be evaluated.

Treatment for the SGTR analysis:

Tube uncovery does not occur in the intact SGs following this event and the subsequent reactor trip.

NOC-AE-07002127 Attachment 1 Page 162 of 219 4.6 STEAM GENERATOR TUBE R UPTURE 4.6.2 Analytical Model The RADTRAD computer code is used to determine the accident doses, consistent with Regulatory Guide 1.183.

For the analysis of radionuclides other than noble gas, calculation of doses in the accident-initiated (AI) iodine spike case, the appearance rate of the nuclides from the reactor core is a significant factor in the analysis. Each nuclide has a unique appearance rate based on preaccident production and spiking assumptions. Complicating this case are modeling limitations in the RADTRAD code that do not allow the code to either explicitly model appearance rates, source distribution, and application of partition factors to nodes or nuclides. In order to accurately model these behaviors in a reasonably limited number of computer calculations, some simplifying assumptions were made:

I) The 1-131 appearance rate is arbitrarily selected as the base core to RCS appearance rate. The core is modeled as a 1 cubic foot volume, and the core to RCS flow rate corresponded to the 1-131 appearance rate. This only applies to the Al spike cases.

2) The Al spike source terms for non-noble gas nuclides was scaled by the ratio of the nuclide's appearance rate to the 1-131 appearance rate. This modified the nuclide source so that the curie count appearing for release remains correct despite the difference between the actual and modeled flow from fuel to RCS to release.

3) The organic iodine available for release is calculated based on each nuclide's appearance rate and the 8-hour time period. This iodine was modeled with the noble gases as a puff release because neither set of nuclides are assumed to be partitioned.

4) For pre-existing (PE) iodine spiking, the source term in the RCS is calculated directly and appearance rates from the fuel were not a part of that model.

5) Because the code does not allow multiple sources, the primary source was assumed present in an initial node, with additional activity assigned via a "fraction" term. For core release sources (Al spikes), this fraction for the RCS and SGs is based on the core activity level (with the core assigned a fraction of

-1.0). For all RCS release sources (both Al and PE spikes), the source fractions in the steam generators are based on the conservatively high PE iodine spike source distribution.

These modeling assumptions allow the control room and offsite doses for the either iodine spike to be modeled in three computer runs:

" Iodine, cesium, rubidium modeled from core to RCS to SG to environment with steam generator flows partitioned;

• Noble gases and organic iodine eight-hour puff release from RCS to SG to environment without partitioned steam generator flows; and,

NOC-AE-07002127 Attachment 1 Page 163 of 219 4.6 STEAM GENERATOR TUBE RUPTURE 0 Core model with spike-increased isotopes (335x) to RCS (I, Cs, & Rb) directly released from the core to the RCS to the environment without flow partitioning.

These models were run separately for Control Room and TSC modeling, and the analysis was performed for the accident-initiated iodine spike and the preexisting iodine spike.

A schematic of the analytical model is provided in Figure 4.6-1.

Figure 4.6-1: SGTR RADTRAD Model Core 2

RCS 3 F as.[.............. .................................

SG - IntactFlow[ t- E nvi o n"t met5 SG l.................

Faulted 5

6 Recirc Control Room Filtration Exhaust to Env

NOC-AE-07002127 Attachment 1 Page 164 of 219 4.6 STEAM GENERA TOR TUBE R UPTURE 4.6.3 Radiological Source Term For this analysis, only the iodine and noble gas activities, which are conservatively characterized by operation with 1% core fuel defects and the equilibrium and spiked release rates from that fuel, define the source terms. RADTRAD uses these activities, in curies per megawatt, and then applies nuclide release fractions and a specified core power to calculate the source term for a given case. The AST release fractions associated with iodines and noble gases are assumed to be 100%, and are released to the reactor coolant.

No additional fuel damage is assumed due to this accident. Two different cases of iodine spiking are analyzed, in accordance with regulatory guidance as previously described.

4.6.3.1 Reactor Coolant System Source Term 4.6.3.1.1 RCS Iodine Concentrations Table 4.2-14 shows the calculation for the Reactor Coolant System (RCS) iodine concentration, based on Thyroid DCFs, for 1% failed fuel that was used in this analysis. Table 4.2-17 shows the calculation for the RCS iodine concentration, based on Thyroid DCFs, for a Pre-existing Iodine Spike.

For the Accident-induced iodine spike, the iodine release rates corresponding to a RCS concentration of 1 pCi/gmn are calculated using the methodology described in Section 4.5.2. The release rates are then multiplied by the RCS mass, and a factor of 335 to yield a release rate in units of Ci/minute. The iodines are assumed to be 95% particulate, 4.85% elemental, and 0.15%

organic. The elemental and particulate iodines are released to the RCS at the 1-131 release rate.

For conservatism, no partition factor is assigned to organic iodines and the modeling assumes that the total organic iodines released from the gap during the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period is instantaneously released to the RCS (i.e. puff release) following the initiation of the event. Table 4.6-1 shows the total iodine spike activity.

Table 4.6-1 RCS Iodine Inventory Due to an 8-Hour Accident-Induced Spike (335x Release Rate)

CLB@ 500x AST @ 335x Isotope (Ci) (Ci)  % Difference

!-131 1.84E+05 1.16E+05 -37.0%

1-132 4.95E+05 3.77E+05 -23.8%

1-133 3.30E+05 2.16E+05 -34.5%

1-134 1.95E+05 1.57E+05 -19.4%

1-135 2.33E+05 7.44E+05 219.1%

NOC-AE-07002127 Attachment 1 Page 165 of 219 4.6 STEAM GENERATOR TUBE RUPTURE 4.6.3.1.2 RCS Noble Gas Concentrations Table 4.2-14 shows the calculation for the RCS noble gas concentration for 1% failed fuel.

However, Kr-89 and Xe-137 were not used in the AST analyses. Because noble gases are not subject to spiking, the same source terms are used in both spiking cases. Also, the noble gases released from the RCS are modeled as an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> integrated puff release.

4.6.3.1.3 RCS Cesium and Rubidium Concentrations Iodine spikes are conservatively assumed to cause an increase in Cesium and Rubidium activities, along with the increase in iodine concentrations. Table 4.2-18 shows the activity in the RCS due to a pre-accident spike. The Cs and Rb activity released due to an accident-induced spike are modeled as a time release into the RCS at the rate of the 1-131 release rate.

4.6.3.2 Secondary System Source Terms 4.6.3.2.1 Secondary System Iodine Concentrations The secondary systems iodine concentrations corresponding to the Technical Specification limit of 0.10 gCi/gm are given in Table 4.2-19 4.6.3.2.2 Secondary System Noble Gas Concentrations The secondary systems noble gas concentrations corresponding to 1.0% failed fuel are given in Table 4.2-20.

4.6.4 Radiological Releases The activity release model is consistent with the model given on Figure E-1 of Regulatory Guide 1.183. Activity that originates in the RCS is released to the secondary coolant by means of the RCS break flow and the primary-to- secondary coolant leak rate.

Activity that originates in the RCS is released to the secondary coolant by means of the primary-to-secondary coolant leak rate and the break flow in the ruptured SG. The total primary-to-secondary coolant leak rate is assumed to be 1 gpm. After the accident, 0.35 gpm of the primary-to-secondary coolant leakage is assumed to occur in the ruptured SG and 0.65 gpm in the intact SGs. This leakage continues for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

The flashed portion of the ruptured tube break flow in the ruptured SG conservatively goes directly to the environment, without mixing with any secondary coolant. Therefore, no partitioning of any nuclides is expected to occur in this release pathway. The unflashed portion of the break flow and the 0.35 gpm normal primary-to-secondary leakage are assumed to mix in the bulk water of the SG.

NOC-AE-07002127 Attachment .1 Page 166 of 219 4.6 STEAM GENERATOR TUBE RUPTURE The intact SGs do not experience tube bundle uncovery. Therefore, primary-to-secondary coolant leakage into the intact SGs mixes with the bulk water in the SG and no flashing to the environment is assumed to occur.

For all post-accident releases through the PORVs of the intact SG loops, the mechanism for release to the environment is steaming of the secondary coolant. Because of this release dynamic, Regulatory Guide 1.183 allows for a reduction in the amount of activity released to the environment based on partitioning of nuclides between the liquid and gas states of water. For iodine, Regulatory Guide 1.183 allows a partition coefficient of 100 for all iodines. However, organic iodines are assumed to be released directly to the environment. Reviewing the specified AST release fractions, it is concluded that the only nuclides other than iodines to be released from the core source term are noble gas nuclides. Because of their volatility, 100% of the noble gases are assumed to be released.

The methodology used to model steaming of activity through PORVs following the postulated SGTR event assumes an average cumulative release rate through the SG PORVs. The partition factors are applied to these release rates. This data was then converted using the assumption of cooled liquid conditions (i.e., 62.4 lbm/ft3), as specified by the applicable guidance of Regulatory Guide 1.183. The steaming release and primary-to-secondary coolant leakage is postulated to end at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, when the RCS and secondary loop have reached equilibrium. Steam release through the above seat main steam line orifices is assumed to conservatively continue for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Per Appendix E of Regulatory Guide 1.183, the chemical form of radionuclide released from the fuel is assumed to be 95% cesium iodine (CsI), 4.85% elemental iodine, and 0.15% organic iodine. This analysis assumes that iodine released from the steam generators to the environment is 4.2% elemental, 13.1% organic, and 82.7% particulate (see Section 4.2.5).

Different forms of radionuclide have different transport behaviors. Based on Regulatory Guide 1.183, the particulate form of radioiodine (CsI) is not released from the steam generator to the environment. However, for conservatism, particulate iodine released from the intact steam generators is assumed to have the same partition coefficient as elemental iodine.

All releases from the SG PORVs (i.e., from the intact SGs) and the ruptured SG are considered ground releases from the nearest PORV to the Control Room HVAC intake.

4.6.4.1 Thermal/Hydraulic Analysis of the SGTR The sequence of events for the thermal hydraulics model is provided in Table 4.6-2.

NOC-AE-07002127 Attachment 1 Page 167 of 219 4.6 STEAM GENERA TOR TUBE RUPTURE Table 4.6-2 Thermal Hydraulics Analysis Sequence of Events (seconds)

Event Time SG Tube Rupture 0 Reactor Trip 66.5 SI Actuation 544.5 Ruptured SG Isolated 607 Ruptured SG PORV Fails Open 611 Ruptured SG PORV Block Valve Closed 1507 RCS Cool-down Initiated 1800 Two Charging Pumps Started 1800 Break Flow Stops Flashing 2027 RCS Cool-down Terminated 2610.6 RCS Depressurization Initiated 3150.6 RCS Depressurization Terminated 3305.8 SI Terminated 3425.8 Excess Charging Flow Eliminated 3425.8 Break Flow Terminated 5128 From this detailed timeline for the accident, the dose analysis uses the nine events in Table 4.6-3 in the model. The last three entries in the table below are taken from the description of the transient. These times are also used in the CLB analysis.

NOC-AE-07002127 Attachment 1 Page 168 of 219 4.6 STEAM GENERA TOR TUBE R UPTURE Table 4.6-3 Thermal Hydraulics Analysis Time Points used in the SGTR Dose Analysis (seconds)

Event T&H Time SG Tube Rupture 0 Reactor Trip 66.5 Ruptured SG Isolated 607 Ruptured SG PORV Block Valve Closed 1507 RCS Flashing in Faulted SG ends 2027 Break Flow Terminated 5128 Dose Model ParameterChanges 7200 RHR Entry 28800 End of Orifice Releases 129600 The mass flows to and from the SGs for both the CLB and the AST analyses are presented in Tables 4.6-4 through 4.6-7.

Table 4.6-4 Thermal Hydraulics Analysis Total Break Flow Flow During Period Time Period (sec) (lbm) 0 to 66.5, 3941 66.5 to 607 25368 607 to 1507 50333 1507 to 2027 26649 2027 to 5128 97094 5128 to 7200 0 7200 to 28800 0 28800 to 1.296E+5 0

NOC-AE-07002127 Attachment 1 Page 169 of 219 4.6 STEAM GENERA TOR TUBE R UPTURE Table 4.6-5 Thermal Hydraulics Analysis Flashed Break Flow Flow During Period Time Period (see) (Ibm) 0 to 66.5 617

.66.5 to 607 1696 607 to 1507 6900 1507 to 2027 2208 2027 to 5128 0 5128 to 7200 0 7200 to 28800 0 28800 to 1.296E+5 0 Table 4.6-6 Thermal Hydraulics Analysis Total Intact SG Steam Flow to Atmosphere Flow During Period Time Period (sec) (Ibm) 0 to 66.5 240000 66.5 to 607 37085 607 to 1507 7369 1507 .to 2027 120353 2027 to 5128 269795 5128 to 7200 227041 7200 to 28800 1158465 28800 to 1.296E+5 0 The iodine partition factor of 100 for liquid and steam phase for steam generator releases is modeled by reducing the flow out of the SG by a factor of 100. The steam released from the condenser (during 0-66.5 seconds) has a total iodine DF of 10,000 from changing phase in the SG and exiting through the condenser. After 66.5 seconds, the condenser is longer available due to the assumed loss of offsite power.

NOC-AE-07002127 Attachment 1 Page 170 of 219 4.6 STEAM GENERA TOR TUBE R UPTURE Table 4.6-7 Thermal Hydraulics Analysis Total Ruptured SG Steam Flow to Atmosphere Total Ruptured SG Flow Time Period (sec) During Period (lbm) 0 to 66.5 84000 66.5 to 607 10066 607 to 1507 131365 1507 to 2027 0 2027 to 5128 0 5128 to 7200 0 7200 to 28800 50040 28800 to 1.296E+5 0 4.6.4.2 Modification of the Thermal/Hydraulic Data for Dose Analysis The time intervals determined by the thermal/hydraulic analyses were arbitrarily increased to provide additional margin in the dose analyses. The adjusted times listed below are based on the data in Table 4.6-3.

Table 4.6-8 Modified Time Sequence of Events for SGTR Dose Analysis Event Time (sec)

SG Tube Rupture 0 Reactor Trip 66.5 Ruptured SG Isolated 607 Ruptured SG PORV Block Valve Closed 1507 RCS Flashing in Ruptured SG ends 2087 Break Flow Terminated 5248 Dose Model ParameterChanges 7380 RHR Entry 28980 End of Orifice Releases 129800 Flow rates are calculated using the time periods taken from Table 4.6-3. These flow rates are arbitrarily increased by 40% and the increased flow rates are assumed to exist during the longer phase intervals with the adjusted times in Table 4.6-9. This results in a larger integrated mass release. The mass releases used in the SGTR are presented in Tables 4.6-9 through 4.6-12. In actual use in RADTRAD these mass releases are converted to cfin using a cold water density of 8.33 lbm/gal.

Tables 4.6-9 through 4.6-12 report a total mass released during each time period. Although the analyses use a volumetric flow rate, the total flow values are presented to illustrate the

NOC-AE-07002127 Attachment 1 Page 171 of 219 4.6 STEAM GENERA TOR TUBE R UPTURE conservatisms built into the analyses. The values for the "Dose Analysis" columns are calculated as:

T & HFlow,Table 4.6 - 5 Dose Analysis Value =TT & &HFlo, Table 4.6-

• 1.4

• Dose At from Table 4.6-9 HAt, Table4.6-4 Table 4.6-9 Total Break Flow used in SGTR Dose Analysis Total Flow During Period (Ibm)

Ending Time T&H Dose Phase (sec) Analysis Analysis  % Difference SG Tube Rupture 0 - --

Reactor Trip 66.5 3941 5517 40%

Ruptured SG Isolated 607 25368 35515 40%

Ruptured SG PORV Block Valve Closed 1507 50333 70466 40%

RCS Flashing in Faulted SG Stops 2087 26649 41613 56%

Break Flow Terminated 5248 97094 138562 43%

Dose Model ParameterChanges 7380 0 0 0%

RHR Entry 28980 0 0 0%

End of Orifice Releases 129800 0 0 0%

Table 4.6-10 Total Flashed Break Flow used in SGTR Dose Analysis Total Flow During Period (lbm)

Ending Time T&H Dose Phase (sec) Analysis Analysis  % Difference SG Tube Rupture 0 -

Reactor Trip 66.5 617 864 40%

Ruptured SG Isolated 607 1696 2374 40%

Ruptured SG PORV Block Valve Closed 1507 6900 9660 40%

RCS Flashing in Faulted SG Stops 2087 2208 3448 56%

Break Flow Terminated 5248 0 0 0%

Dose Model ParameterChanges 7380 0 0 0%

RHR Entry 28980 0 0 0%

End of Orifice Releases 129800 0 0 0%

NOC-AE-07002127 Attachment 1 Page 172 of 219 4.6 STEAM GENERA TOR TUBE RUPTURE Table 4.6-11 Total Intact SG Flow to Atmosphere Used in SGTR Dose.Analysis Total Flow During Period (Ibm)

Time T&H Dose Phase (sec) Analysis Analysis  % Difference SG Tube Rupture 0 - -

Reactor Trip 66.5 240000 336000 40%

Ruptured SG Isolated 607 37085 51919 40%

Ruptured SG PORV Block Valve Closed 1507 7369 10317 40%

RCS Flashing in Faulted SG Stops 2087 120353 187936 56%

Break Flow Terminated 5248 269795 385021 43%

Dose Model ParameterChanges 7380 227041 327062 44%

RHR Entry 28980 1158465 1621851 40%

End of Orifice Releases 129800 0 0 0%

Table 4.6-12 Total Ruptured SG Flow to Atmosphere Used in SGTR Dose Analysis Total Flow During Period (lbm)

Time T&H Dose Phase (sec) Analysis Analysis  % Difference SG Tube Rupture 0 - -

Reactor Trip 66.5 84000 117600 40%

Ruptured SG Isolated 607 10066 14092 40%

Ruptured SG PORV Block Valve Closed 1507 131365 183911 40%

RCS Flashing in Faulted SG Stops .2087 0 0 0%

Break Flow Terminated 5248 0 0 0%

Dose Model ParameterChanges 7380 0 0 0%

RHR Entry 28980 50040 70056 40%

End of Orifice Releases 129800 0 0 0%

NOC-AE-07002127 Attachment 1 Page 173 of 219 4.6 STEAM GENERA TOR TUBE RUPTURE 4.6.5 Assumptions and Inputs The following inputs and assumptions are used in the SGTR analysis.

1. The source term is based upon a power level of 4100 MW thermal, 5 w/o enrichment, and a three region core with equilibrium cycle core at end of life. The three regions have operated at a specific power of 39.3 MW/MTU for 509, 1018, and 1527 EFPD, respectively. The assumed power level is greater than the Rated Thermal Power of 3853 MWth plus a 0.6% measurement uncertainty.

2. The equilibrium secondary activity before the accident is based upon a pre-incident primary-to-secondary leakage of 1 gpm. This is conservative since the Technical Specifications limits the pre-accident leakage to 150 gpd per steam generator or 600 gpd (0.42 gpm) total. The secondary coolant activity is based on 0.1 gCi/gm of dose equivalent 1-131. Noble gas activity in the secondary coolant is based on 1% failed fuel.

3. No fuel failures are assumed to be caused by the SGTR.

4. Total primary-to-secondary leakage through the steam generator tubes prior to the accident and during the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the transient is 1 gpm. Eight hours after the accident, the residual heat removal system starts and primary-to-secondary leakage is stopped.

Primary-to-secondary leakage is conservatively modeled at 0.65 gpm for the three intact steam generators and at 0.35 gpm for the ruptured steam generator.

5. The intact SGs do not experience tube bundle uncovery. Therefore, primary-to-secondary coolant leakage into the intact SGs mixes with the bulk water in the SG and no flashing to the environment is assumed to occur.

6. For a Pre-accident iodine spike, the activity in the reactor coolant is based upon an iodine spike which has raised the reactor coolant concentration to 60 pCi/gm of dose equivalent I-131. Noble gas activity is based on 1% failed fuel.

7. For an Accident-induced iodine spike, the accident initiates an iodine spike in the RCS which increases the iodine release rate from the fuel to a value 335 times greater than the release rate corresponding to a RCS concentration of 1 tCi/grn dose equivalent 1-131.

lodines, Cs, and Rb are assumed to be released at this rate for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (Cs and Rb are released at the rate of 1-131). The iodine activity released from the fuel to the RCS is conservatively assumed to mix instantaneously and uniformly in the RCS. Since Regulatory Guide 1.183 specifies that the chemical form of particulate iodine is (CsI), the spike is also assumed to relatively increase the Alkali metal (Cs and Rb) in the RCS.

Noble gas activity is conservatively based on 1% failed fuel.

8. The activity released from the fuel gap is assumed to be instantaneously mixed with the reactor coolant within the pressure vessel per Regulatory Guide 1.183.

9. A partition coefficient of 100 is assumed for elemental iodine released from the steam generators. (Regulator Guide 1.183, Appendix E, Section 5.6) Organic iodine is not partitioned. Organic iodine is assumed to migrate directly to the steam space and become immediately available for release.

NOC-AE-07002127 Attachment 1 Page 174 of 219 4.6 STEAM GENERA TOR TUBE R UPTURE

10. Operators are assumed to identify the ruptured steam generator and attempt to close the PORV on the ruptured steam generator in 10 minutes. However, the PORV is assumed to fail open (the single failure for this accident scenario) at that time. It is assumed that the failed PORV is isolated by manually closing the PORV block valve within 15 minutes of the PORV failure. Therefore, the steam release via the ruptured steam generator's PORV is assumed to continue for a total of 25 minutes.

11. Eight hours after the accident, the residual heat removal system is in operation and no further steam containing radionuclides is released from steam generators to the environment except the leakage through the MSIV above-seat drain orifices. The release through the orifices continues until 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after the start of the accident. (These orifice releases occur in the Isolation Valve Cubicle next to the PORVs. Therefore, the PORV-to-Control Room X/Qs are used for the Control Room and TSC dose analyses.) This is conservative since all releases would terminate in less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when the RHR system is in operation.

12. The SG releases are Via the PORVs and ruptured SG safety valves. The above-seat drain releases occur in the Isolation Valve Cubicle next to the PORVs. Therefore, the PORV-to-Control Room X/Qs are used for the Control Room and TSC dose analyses.

13. Offsite Power is lost. After 66.5 seconds, the condensers are unavailable for steam dump.

14. The Control Room ventilation system automatically transfers to the emergency mode of operation after the initiation of safety injection. This is assumed to happen at t-0 instead of upon reactor trip at 66.5 seconds. Since the mass releases are increased by about 40%, the time difference is negligible.

15. All activity is released to the environment with no consideration given to cloud depletion by ground deposition during transport to the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ).

16. Reactor coolant density is 8.33 lbs/gal.

Input parameters used for the SGTR analysis are given in Table 4.6-13. Conformance with Regulatory Guide 1.183 guidance addressing the SGTR analysis is provided in Attachment 6, Tables A and F.

NOC-AE-07002127 Attachment 1 Page 175 of 219 4.6 STEAM GENERATOR TUBE R UPTURE Table 4.6-13 Inputs for SGTR Analysis Parameter CLB AST Core power (for radiological source terms) 410OMWt Core power level 3876 MWt (for steam releases) (3853MWt + 0.6%)

RCS & Secondary density 8.33 Ibm/gallon RCS Mass 2.658E+8 gm SG Node Volume Intact 5.94E+4 gal 5.937E+4 gal Ruptured 1.98E+4 gal 1.979E+4 gal Secondary Mass 659,412 ibm Primary-to-Secondary Leakage Intact 0.42 gpm 0.65 gpm Ruptured 0 gpm 0.35 gpm Accident Time line Table 4.6-8 Operator Action Times diagnose SGTR @10 minutes close PORV block valve on ruptured SG @25 minutes Total Break Flow Table 4.6-9 Total Flashed Break Flow Table 4.6-10 Total Intact SG Flow to Atmosphere Table 4.6-11 Total Ruptured SG Flow to Atmosphere Table 4.6-12 Release from Above (MSIV) Seat Drains Intact SGs 5.79 lbm/sec Ruptured SG 1.93 lbm/sec Steam Flow rate 1.574E+7 lbm/hr DF in condenser (before LOOP) 10,000 Iodine appearance rate into the RCS for the 500 335 accident-induced spike Iodine Species Released from the RCS (%) 91/4/5 4.85/0.15/95 (elemental/organic/particulate)

Iodine Species for Flashed RCS Break Flow to the environment (%) 91/4/5 4.85/0.15/95 (elemental/organic/particulate)

Iodine Partition Factors for Releases from the Secondary Side 100/100/100 100/1/100 (elemental/organic/particulate)

NOC-AE-07002127 Attachment 1 Page 176 of 219 4.6 STEAM GENERATOR TUBE RUPTURE Table 4.6-13 Inputs for SGTR Analysis Parameter CLB AST Resulting Iodine Species Released from the Secondary Side to Environment 91/4/5 4.2/13.1/82.749

% (elemental/organic/particulate)

Dose Conversion Factors Table 4.2-6 Decay Constants and Decay Daughter Table 4.2-7 Fractions Table_4.2-7 Offsite breathing rates Table 4.2-1 Offsite X/Q's Table 4.1-24 Control Room HVAC Parameters Table 4.2-3 Control Room HVAC Flow Rates Table 4.2-2 TSC HVAC Parameters Table 4.2-5 TSC HVAC Flow Rates Table 4.2-4 Control Room and TSC x/Q's Table 4.1-37 49 See Section 4.2.5

NOC-AE-07002127 Attachment 1 Page 177 of 219 4.6 STEAM GENERATOR TUBE RUPTURE 4.6.6 Summary and Conclusions Table 4.6-14 below provides the results for the SGTR scenarios.

Table 4.6-14 SGTR Dose Results (rem TEDE)

Pre-Existing Iodine Spike I Accident-induced Iodine Spike Receptor Result Limit Result Limit EAB (worst 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) 2.37 25ýu 1.08 2.551 2,551 LPZ 0.92 2550 0.44 Control Room 2.15 550 1.00 550 TSC 2.09 5 0.98 5 For the cases analyzed in this calculation, it is shown that a SGTR that involves a pre-accident 60 jiCi/gm iodine spike, which instantaneously releases activity into the RCS prior to initiating SGTR releases, would be the bounding SGTR accident scenario. All doses are well below their respective acceptance criteria; therefore, this design-basis SGTR accident is sufficiently mitigated.

50 10CFR50.67 51 10CFR50.67 as modified by Regulatory Guide 1.183 in Table 6 on Page 1.183-20.

NOC-AE-07002127 Attachment 1 Page 178 of 219 4.7 CONTROL ROD EJECTION ACCIDENT 4.7 Control Rod Ejection Radiological Assessment The Control Rod Ejection Accident (CREA) is defined as the mechanical failure of a control rod mechanism pressure housing resulting in the ejection of an RCCA and drive shaft. The consequence of this mechanical failure is a rapid positive reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel rod damage.

4.7.1 Methodology Overview An analysis of the effects of a postulated rod ejection accident is performed using the assumptions of Regulatory Guide (RG) 1.183. For the analysis, it is assumed that prior to the postulated accident, the plant is operating at an equilibrium level of radioactivity in the primary and secondary systems as a result of coincident fuel defects and SG tube leakage. Following a postulated rod ejection accident, two activity release paths contribute to the total radiological consequences of the accident. The first release path is via Containment leakage resulting from release of activity from the primary coolant to the Containment. The second path is the contribution of steam in the secondary system dumped through the SG PORVs and safety valves since offsite power is assumed to be lost.

4.7.1.1 Comparison of Modeling with Regulatory Guide 1.183, Appendix E Regulatory Guide 1.183 Appendix E, Position 5.5. 1:

A portion of the primary-to-secondaryleakage willflash to vapor, based on the thermodynamic conditions in the reactorand secondary coolant.

Duringperiods of steam generatordryout, all of the primary-to-secondary leakage is assumed to flash to vapor and be releasedto the environment with no mitigation.

• With regard to the unaffected steam generatorsusedforplant cooldown, the primary-to-secondaryleakage can be assumed to mix with the secondary water withoutflashing duringperiods of total tube submergence.

Treatment for the CREA analysis:

The SGs do not experience tube uncovery. The primary-to-secondary leakage is assumed to mix with the secondary water without flashing.

Regulatory Guide 1.183 Appendix E, Position 5.5.2:

The leakage that immediatelyflashes to vapor will rise through the bulk water of the steam generatorand enter the steam space. Credit may be taken for scrubbing in the generator,using the models in NUREG-0409, "Iodine Behavior in a PWR Cooling System Followinga PostulatedSteam GeneratorTube Rupture Accident" (Ref E-2), duringperiods of total submergence of the tubes.

NOC-AE-07002127 Attachment 1 Page 179 of219 4.7 CONTROL ROD EJECTION ACCIDENT Treatment for the CREA analysis:

This assumption is not used. It is assumed that the primary-to-secondary leakage does not flash in the SGs. The nuclides in the primary-to-secondary leakage are added to the bulk fluid in the SGs.

Regulatory Guide 1.183 Appendix E, Position 5.5.3:

The leakage that does not immediatelyflash is assumed to mix with the bulk water.

Treatment for the CREA analysis:

It is assumed that the primary-to-secondary leakage does not flash in the SGs. The nuclides in the primary-to-secondary leakage are added to the bulk fluid in the SGs.

Regulatory Guide 1.183 Appendix E, Position 5.5.4:

The radioactivityin the bulk water is assumed to become vapor at a rate that is the function of the steaming rate and the partitioncoefficient. A partitioncoefficient for iodine of 100 may be assumed. The retention ofparticulateradionuclides in the steam generatorsis limited by the moisture carryoverfrom the steam generators.

Treatment for the CREA analysis:

A partition coefficient of 100 is assumed for elemental iodine released from the steam generators. Organic iodine is not partitioned. Organic iodine is assumed to migrate directly to the steam space and become immediately available for release.

Regulatory Guide 1.183 Appendix E, Position 5.6:

Operating experience and analyses have shown thatfor some steam generatordesigns, tube uncovery may occurfor a shortperiodfollowing any reactor trip (Ref E-3). The potential impact of tube uncovery on the transportmodel parameters (e.g., flashfraction, scrubbingcredit) needs to be considered. The impact of emergency operatingprocedure restorationstrategies on steam generatorwater levels should be evaluated.

Treatment for the CREA analysis:

Tube uncovery does not occur following this event and the subsequent reactor trip.

4.7.2 Analytical Model It is assumed that prior to the accident the plant has been operating with simultaneous fuel defects and SG tube leakage for a period of time sufficient to establish equilibrium levels of activity in the primary and secondary coolant. The model for the activity available for leakage from the Containment assumes that the activity in the fuel pellet-clad gap and the activity released due to fuel melting are instantaneously mixed in the Containment and available for release. All of the gap activity of the fuel rods failed by accident is assumed released to the Containment. Of the fuel melted, 100 percent of the noble gases and 25 percent of the iodines are

NOC-AE-07002127 Attachment 1 Page 180 of 219 4.7 CONTROL ROD EJECTION ACCIDENT assumed available for leakage from the Containment. The only removal processes considered for the Containment are radioactive decay and leakage.

The model for the activity available for release to the atmosphere from the safety valves assumes that the release consists of the activity in the secondary coolant prior to the accident plus that activity leaking from the primary coolant through the SG tubes following the accident. The primary coolant activity after the accident is assumed to be composed of the equilibrium activity prior to the accident, plus 100 percent of the noble gases and iodines released by fuel failed during the accident, plus 100 percent of the noble gases and 50 percent of the iodines released by fuel melted by the accident. The leakage of primary coolant to the secondary side of the SG is assumed to continue at its initial rate, assumed to be the same rate as the leakage prior to the accident, until the pressures in the primary and secondary systems are equalized. No mass transfer from the primary system to the secondary system is assumed thereafter.

Since a coincident loss of offsite power is assumed, activity is assumed to be released to the atmosphere through the SG PORVs and not the main condenser.

The two release pathways are shown below.

Figure 4.7-1: CREA Analysis Model 10% FF Gap Release Containment 100% MeltedN. Gas A Leakage 25% Melted lodines otalltflent 10% FF Gap Release 1 gpm 100% Melted N. Gas R Leakage PORVs 50% Melted lodines R 1n 4.7.3 Radiological Source Term The sudden rod ejection and localized temperature spike associated with the CREA results in 10% core damage. Only 2.5% of the damaged core releases melted fuel activity (i.e., 0.25% of the total core melts). Therefore for both cases, the source term available for release is associated

-with this fraction of melted fuel and the fraction of core activity existing in the gap.

Release fractions and transport fractions conform to Regulatory Guide 1.183, Appendix H and Table 3. To conform to this regulatory guidance, 10% of the core inventory of iodine and noble

NOC-AE-07002127 Attachment 1 Page 181 of 219 4.7 CONTROL ROD EJECTION ACCIDENT gas is assumed to be in the fuel-clad gap. Additionally, Table 3 of Regulatory Guide 1.183 shows that 12% of the core cesium and rubidium should be assumed to be in the fuel-clad gap and should be released in its entirety from the damaged 10% of the total core.

With regard to the fraction released from melted fuel, it is assumed that 90% of the core inventory of iodine and noble gas, and 88% of the core cesium and rubidium remain available for release due to melting (i.e., these are the remaining fractions of activity that are not in the fuel-clad gap).

One hundred percent of the noble gases and iodines in the clad gaps of the fuel rods experiencing clad damage (assumed to be 10 percent of the rods in the core) is assumed released. The accident evaluation conservatively assumes this activity to be released twice: to the Containment for leakage to the atmosphere and to the primary coolant for leakage to the secondary system.

The fraction of fuel melting is assumed to be 0.25 percent of the core as determined by the following method:

1. A conservative upper limit of 50 percent of rods experiencing clad damage may experience centerline melting (a total of 5 percent of the core).

2. Of the rods experiencing centerline melting, only a conservative maximum of the innermost 10 percent of the volume actually melts (0.5 percent of the core could experience melting).

3. A conservative maximum of 50 percent of the axial length of the rod would experience melting due to the power distribution (0.5 of the 0.5 percent of the core = 0.25 percent of the core).

NOC-AE-07002127 Attachment 1 Page 182 of 219 4.7 CONTROL ROD EJECTION ACCIDENT 4.7.3.1 Reactor. Core Releases 4.7.3.1.1 Release from Cladding Failures Table 4.7-1 provides the radionuclides released from the gap of the 10% failed fuel.

Release fractions and transport fractions are consistent with Regulatory Guide 1.183, Appendix G and Table 3. To conform with this regulatory guidance, 5% of the core inventory of iodine and noble gas is assumed to be in the fuel-clad gap, excluding 1-131 and Kr-85, where 8% and 10%

are assumed, respectively. Additionally, Table 3 of Regulatory Guide 1.183 shows that 12% of the core cesium and rubidium should be assumed to be in the fuel-clad gap.

Table 4.7-1 10% Failed Fuel Gap Release Source (Ci)

Isotope CLB AST  % Difference 1-131 1.1OE+06 8.50E+05 -22.7%

1-132 1.60E+06 7.60E+05 -52.5%

1-133 2.30E+06 1.1OE+06 -52.2%

1-134 2.50E+06 1.20E+06 -52.0%

1-135 2.1OE+06 1.OOE+06 -52.4%

Kr-83m 1.40E+05 7.OOE+04 -50.0%

Kr-85m 3.00E+05 1.50E+05 -50.0%

Kr-85 3.70E+04 1.20E+04 -67.6%

Kr-87 5.50E+05 2.80E+05 -49.1%

Kr-88 7.90E+05 3.90E+05 -50.6%

Kr-89 9.70E+05 4.80E+05 -50.5%

Rb-88 - 9.50E+05 Rb-89 - 1.20E+06 Xe-131m 7.70E+03 5.50E+03 -28.6%

Xe-133m 3.30E+05 3.40E+04 -89.7%

Xe-133 2.30E+06 1.1OE+06 -52.2%

Xe-135m 4.60E+05 2.1OE+05 -54.3%

Xe-135 6.50E+05 2.80E+05 -56.9%

Xe-137 2.OOE+06 9.50E+05 -52.5%

Xe-138 1.90E+06 9.OOE+05 -52.6%

Cs-134 - 2.6E+05 Cs-136 - 7.6E+04 Cs-137 - 1.6E+05 Cs-138 - 2.4E+06

NOC-AE-07002127 Attachment 1 Page 183 of 219 4.7 CONTROL ROD EJECTION ACCIDENT 4.7.3.1.2 Release from Fuel Melt The material released as a result of the fuel melt is presented in Table 4.7-2. These are based upon the inventory in Table 4.2-9.

Table 4.7-2 0.25% Core Melt Source Isotope CLB AST  % Difference 1-131 2.75E+05 2.8E+05 1.8%

1-132 4.OOE+05 3.8E+05 -5.0%

1-133 5.75E+05 5.5E+05 -4.3%

1-134 6.25E+05 6.OE+05 -4.0%

1-135 5.25E+05 5.OE+05 -4.8%

Kr-83m 3.50E+04 3.5E+04 0.0%

Kr-85m 7.50E+04 7.3E+04 -2.7%

Kr-85 3.08E+03 3.OE+03 -2.6%

Kr-87 1.38E+05 1.4E+05 1.4%

Kr-88 1.98E+05 2.OE+05 1.0%

Kr-89 2.43E+05 2.4E+05 -1.2%

Rb-88 - 2.OE+05 Rb-89 - 2.5E+05 Xe-131m 1.93E+03 2.8E+03 45.1%

Xe-133m 8.25E+04 1.7E+04 -79.4%

Xe-133 5.75E+05 5.5E+05 -4.3%

Xe-135m 1.15E+05 1.1E+05 -4.3%

Xe-135 1.63E+05 1.4E+05 -14.1%

Xe-137 5.OOE+05 4.8E+05 -4.0%

Xe-138 4.75E+05 4.5E+05 -5.3%

Cs-134 - 5.5E+04 Cs-136 - 1.6E+04 Cs-137 - 3.3E+04 Cs-138 - 5.OE+05 4.7.3.2 Reactor Coolant System Source Terms 4.7.3.2.1 RCS Iodine Concentrations The initial RCS concentrations are assumed to be at a pre-existing iodine spike level of 60 pCi/gm as shown in Table 4.2-17.

NOC-AE-07002127 Attachment 1 Page 184 of 219 4.7 CONTROL ROD EJECTIONACCIDENT 4.7.3.2.2 RCS Noble Gas Concentrations The initial RCS noble gas concentrations corresponding to 1%failed fuel are given in Table 4.2-14.

4.7.3.2.3 RCS Cesium and Rubidium Concentrations The RCS cesium and rubidium concentrations corresponding to a 1% failed fuel (Table 4.2-14).

The Cs and Rb is assumed not to spike along with the iodines. Since the Cs and Rb are bound into particulate iodines, and since the iodines do not leave the water in the SGs or are appreciably from the RCB, the impact of this assumption is negligible.

4.7.3.3 Secondary System Source Terms Releases from the secondary systems are only modeled for the Release from Secondary Systems scenario - not for the Release from Containment Building scenario.

4.7.3.3.1 Secondary System Iodine Concentrations The initial secondary systems concentrations are assumed to be at the Technical Specification limit for the secondary side of 0.1 pCi/gmn as shown in Table 4.2-19.

4.7.3.3.2 Secondary System Noble Gas Concentrations The secondary systems noble gas concentrations corresponding to 1.0% failed fuel are given in Table 4.2-20.

4.7.3.3.3 Secondary System Cesium and Rubidium Concentrations Cesium and rubidium are assumed to be bound with iodines as particulates. Therefore, there is no release of Cs or Rb from water in the steam generators.

4.7.4 Radiological Releases In a CRE accident, nuclides released to the RCS from the fuel would be available for release to the environment through two pathways: into the containment and subsequent leakage to the environment; or, leakage to the environment via primary-to-secondary leakage and then steaming from the SGs. In order to bound the resultant doses from this accident, two cases are considered when analyzing the radioactive release:

Scenario 1: RCB Leakage For this scenario, the ejected control rod is assumed to breach the reactor pressure vessel (RPV), effectively causing the equivalent of a small break loss of coolant accident. In this case, all activity from damaged fuel that has been mixed with the

NOC-AE-07002127 Attachment 1 Page 185 of 219 4.7 CONTROL ROD EJECTION A CCIDENT primary coolant of the reactor coolant system (RCS) leaks directly to the containment volume. This flashed release is assumed to instantaneously and homogeneously mix with the containment atmosphere and subsequently be available for release to the environment via an assumed containment leak rate limit.

Credit for mitigation of the release by containment spray is not taken.

Scenario 2: Steam Generator PORV Release All of the activity from damaged fuel is mixed with the RCS. The combined RCS activity then leaks to the secondary side through the steam generator (SG) tubes at a conservative rate of 1.0 gpm total leakage. The activity is then available for release to the environment by steaming of the SG PORVs and safeties.

This methodology maximizes the release of activity released to the environment. Therefore, certain secondary aspects of each scenario, which would be modeled if the scenarios were stand-alone methodologies, are not modeled. Specifically, the dose contribution of primary-to-secondary leakage and subsequent release to the environment via SG releases is not modeled for the Containment Building leakage scenario. Similarly, for the release through the Secondary Side scenario, the dose contribution of any release into the containment building and subsequent leakage to the environment is not modeled.

In reality, the release path would probably be a combination of these two release pathways, but the radiological consequences would be limited by the total doses determined using the independent scenarios.

4.7.4.1 Release from Containment Building Scenario For this scenario, the ejected control rod is assumed to breach the reactor pressure vessel (RPV),

effectively causing the equivalent of a small break loss of coolant accident. In this case, all activity from damaged fuel that has been mixed with the primary coolant of the reactor coolant system (RCS) leaks directly to the containment volume. This flashed release is assumed to instantaneously and homogeneously mix with the containment atmosphere and subsequently be available for release to the environment via an assumed containment leak rate limit.

The nuclides released to the containment are a mix of the core source term (Tables 4.7-1 and 4.7-

2) and the curies contained in the RCS fluid released from the reactor pressure vessel. The total activity released to the containment and available for release is presented in Table 4.7-3.

No releases from the secondary side are assumed in this scenario.

NOC-AE-07002127 Attachment 1 Page 186 of 219 4.7 CONTROL ROD EJECTIONACCIDENT Table 4.7-3 Release From the RCB Scenario:

Total Activity Released into the RCB (Ci)

Isotope CLB AST  % Difference 1-131 1.18E+06 9.3E+05 -21.2%

1-132 1.71E+06 8.8E+05 -48.5%

1-133 2.46E+06 1.2E+06 -51.2%

1-134 2.66E+06 1.4E+06 -47.4%

1-135 2.24E+06 1.2E+06 -46.4%

Kr-83m 1.75E+05 1.1E+05 -37.1%

Kr-85m 3.75E+05 2.2E+05 -41.3%

Kr-85 4.21E+04 1.7E+04 -59.6%

Kr-87 6.88E+05 4.2E+05 -39.0%

Kr-88 9.89E+05 5.9E+05 -40.3%

Kr-89 1.12E+06 7.2E+05 -35.7%

Rb-86 - 4.5E+00 Rb-88 - 1.2E+06 Rb-89 - 1.5E+06 Xe-131m 1.01E+04 9.0E+03 -10.9%

Xe-133m 4.17E+05 5.2E+04 -87.5%

Xe-133 2.94E+06 1.8E+06 -38.8%

Xe-135m 5.75E+05 3.2E+05 -44.3%

Xe-135 8.15E+05 4.2E+05 -48.5%

Xe-137 2.50E+06 1.4E+06 -44.0%

Xe-138 2.38E+06 1.4E+06 -41.2%

Cs-134 - 3.2E+05 Cs-136 - 9.3E+04 Cs-137 - 1.9E+05 Cs-138 - 2.9E+06 4.7.4.2 Release via Secondary Side Scenario Activity that originates in the RCS is released to the secondary coolant by means of the primary-to-secondary coolant leak rate. This design basis leak rate value is 1.0 gpm for all SGs.

Releases to the environment are associated with the secondary coolant steaming from the SGs.

Because of the release dynamic of the activity from the SG PORVs, Regulatory Guide 1.183 allows for a reduction in the amount of activity released to the environment based on partitioning of nuclides between the liquid and gas states of water for this release path. For iodine, the partition factor of 100 was taken directly from the suggested guidance. No particulates are

NOC-AE-07002127 Attachment 1 Page 187 of 219 4.7 CONTROL ROD EJECTION ACCIDENT assumed to be released. Because of their volatility, 100% of the noble gases are assumed to be released. The total activity released to the steam generators is presented in Table 4.7-4.

Table 4.7-4 Release From the Secondary Side Scenario:

Total Activity in the Steam Generators (RCS+SG)

(Ci)

Isotope CLB AST  % Difference 1-131 1.25E+06 1.OE+06 -20.0%

1-132 1.81E+06 9.7E+05 -46.4%

1-133 2.61E+06 1.4E+06 -46.4%

1-134 2.82E+06 1.5E+06 -46.8%

1-135 2.37E+06 1.4E+06 -40.9%

Kr-83m 1.75E+05 1.I E+05 -37.1%

Kr-85m 3.75E+05 2.2E+05 -41.3%

Kr-85 4.21E+04 1.7E+04 -59.6%

Kr-87 6.88E+05 4.2E+05 -39.0%

Kr-88 9.89E+05 5.9E+05 -40.3%

Kr-89 1.21E+06 7.2E+05 -40.5%

Rb-86 - 4.5E+00 -

Rb-88 - 1.2E+06 -

Rb-89 - 1.5E+06 -

Xe-131m 1.01E+04 9.OE+03 -10.9%

Xe-133m 4.17E+05 5.2E+04 -87.5%

Xe-133 2.94E+06 1.8E+06 -38.8%

Xe-135m 5.75E+05 3.2E+05 -44.3%

Xe-135 8.15E+05 4.2E+05 -20.0%

Xe-137 2.50E+06 1.4E+06 -46.4%

Xe-138 2.38E+06 1.4E+06 -46.4%

Cs-134 - 3.2E+05 -

Cs-136 - 9.3E+04 -

Cs-137 - 1.9E+05 -

Cs-138 - 2.9E+06 -

The methodology used to model steaming of activity through intact SG PORVs following the postulated CREA event assumes an average cumulative release rate through the SG PORVs that, for simplicity and conservatism, is reduced in steps. The steaming release from the PORVs and primary-to-secondary coolant leakage is postulated to end at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, when the RCS and secondary loop have reached equilibrium. Leakage via the MSIV above-seat drain orifices is assumed to continue for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Table 4.7-5 shows the time steps and associated release rates.

NOC-AE-07002127 Attachment 1 Page 188 of 219 4.7 CONTROL ROD EJECTIONACCIDENT Table 4.7-5 Steam Released to the Environment (ibm)

CLB AST Time RAbove Seat  %

(Hours) PORV PORV Drains Total Difference 0- 1.25 15,526,178 15,535,885 34,740 15,570,625 0.3%

1.25- 36 0 0 965,772 965,772 36-720 0 0 0 0 Total 15,526,178 15,535,885 1,000,512 16,536,397 6.5%

A constant 7.72 ibm/sec leakage from the MSIV above seat drain orifices is assumed in the revised analyses (0 lbm/sec was assumed in the CLB).

4.7.5 Assumptions and Inputs The following inputs and assumptions were used in the CREA analysis.

Assumptions Applicable to both Scenarios

1. The source term is based upon a power level of 4100 MW thermal, 5 w/o enrichment, and a 3 region core with equilibrium cycle core at end of life. The three regions have operated at a specific power of 39.3 MW/MTU for 509, 1018, and 1527 EFPD, respectively. The assumed power level is greater than the Rated Thermal Power of 3853 MWth plus a 0.6%

measurement uncertainty.

2. The clad of 10% of the fuel is damaged during the initiation of this accident, and is assumed to have failed. Therefore, 10% of the core inventory of noble gases and iodines are released from the fuel gap (Regulatory Guide 1.183, Appendix H). Release fractions of other nuclide groups contained in the fuel gap are detailed in Table 3 of Regulatory Guide 1.183.

3. The amount of fuel melt is 0.25%. The 0.25% of the core is determined by the following method: a) A conservative upper limit of 50% of rods experiencing clad damage may experience centerline melting (a total of 5% of the core); b) Of the rods experiencing centerline melting, only a conservative maximum of the innermost 10 percent of the volume actually melts (0.5% of the core could experienced melting); and c) A conservative maximum of 50% of the axial length of the rod would experience melting due to the power distribution (half of the 0.5% of the core = 0.25% of the core)..

4. The initial RCS iodine concentrations are based on a pre-existing iodine spike to the Technical Specification limit of 60 gCi/gm and the initial Secondary system concentrations are based on a pre-existing iodine spike to the Technical Specification limit of 0.1 gtCi/gm.

Noble Gas concentrations are based on 1% failed fuel.

5. The Control Room ventilation system is assumed to transfer to the emergency mode of operation immediately upon the receipt of the safety injection signal (at t=0).

6. All releases to the atmosphere are assumed to be at ground level.

NOC-AE-07002127 Attachment 1 Page 189 of 219 4.7 CONTROL ROD EJECTION ACCIDENT

7. The RCS density is 8.33 ibm/gal.

Assumptions Specific to the Release via Containment Leakage Scenario

8. One hundred percent of the noble gases and iodines in the gap of the fuel failed by the accident, plus 100% of noble gases and 25% of the iodines contained in the melted fuel fraction are assumed to be released to the containment in accordance with Appendix H of Regulatory Guide 1.183.

9. The containment free volume is 3.41E+6 ft3 (+0.1% / -0.85%) or 3.38E+6 ft3 to 3.41E+6 ft3 .

A value of3.38E+6 ft3 is utilized for the dilution volume in containment and 3.41E+6 ft3 is used for the leakage determination. Utilizing the minimum containment free volume conservatively maximizes the radioactive concentration in containment and using the maximum value for determining the containment leakage conservatively maximizes the containment leakage.

10. The activity released to the containment through the rupture in the reactor vessel head is assumed to mix instantaneously throughout the containment. No credit is assumed for removal of iodine in the containment due to containment sprays.

11. For the containment leakage case, all leakage is assumed to be at the Technical Specification limit of 0.3 percent per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 0.15% per day thereafter.

12. lodines released to the containment (from the fuel and RCS) are assumed to be 95%

particulate, 4.85% elemental, and 0.15% organic (Regulatory Guide 1.183, Appendix H,

.position 4).

Assumptions Specific to the Release via the Secondary Side Scenario

13. One hundred percent of the noble gases and iodines in the gap of the fuel failed by the accident, plus 100% of noble gases and 50% of the iodines contained in the melted fuel fraction are assumed to be released to the reactor coolant in accordance with Appendix H of Regulatory Guide 1.183. Fractions of other nuclides released from the melted fuel are used from Table 2 of Regulatory Guide 1.183. Though these are described as LOCA values for fuel melt release, they are conservatively used for the other nuclide groups.

14. The activity released from the fuel from either the gap or from fuel pellets is assumed to be instantaneously mixed with the reactor coolant within the pressure vessel.

15. Primary-to-secondary leakage is conservatively modeled at a total of 1 gpm for all steam generators. Primary-to secondary leakage stops at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when the RCS and SG pressures are equalized.

.16. lodines released to the Secondary side (from the fuel and RCS) are assumed to be 95%

particulate, 4.85% elemental, and 0.15% organic (Regulatory Guide 1.183, Appendix H, position 4).

17. This analysis assumes that iodine released from the steam generators to the environment is 4.2% elemental, 13.1% organic, and 82.7% particulate (see Section 4.2.5).

18. A partition coefficient of 100 is assumed for iodine, cesium, and rubidium released from the steam generators. (Regulatory Guide 1.183, Appendix G, Section 5.6) Organic iodine is not partitioned. Organic iodine is assumed to migrate directly to the steam space and become immediately available for release.

NOC-AE-07002127 Attachment 1 Page 190 of219' 4.7 CONTROL ROD EJECTIONACCIDENT

19. Upon loss of offsite power, a total of 1.56 x 107 pounds of steam is discharged from the secondary system through the safety valves or PORVs for 4500 seconds following the accident. Steam release is terminated after this time. The minimum time to release the initial steam generator mass is 191 seconds. The rate of release necessary to release the total steam generator mass of 659,412 pounds in 191 seconds is 207,000 lbm/min. Assuming this flow rate is constant for 4500 seconds yields a total mass release of 1.56 x 107 pounds. Note that the total mass released is very conservative in relation to the initial SG mass.

20. Steam continues to be released from the orifices that replaced the MSIV above-seat isolation valves until 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

21. All releases are via the PORVs or safeties and the above-seat drains. These releases occur in the Isolation Valve Cubicle next to the PORVs. Therefore, the PORV-to-Control Room X/Qs are used for the Control Room and TSC dose analyses.

Input parameters used for the CREA analysis are given in Tables 4.7-6 and 4.7-7. Conformance with Regulatory Guide 1.183 guidance addressing CREA analysis is provided in Attachment 6, Tables A and F.

Table 4.7-6 Inputs for CREA Analysis Release from the RCB Scenario Parameter CLB AST Core power (for radiological source terms) 4100 MWt Core power (for steam releases) 3876 MWt (3853MWt_++/-0.6%)

RCS density 8.33 lbm/gallon RCS Volume 2.6E+8 gm F 2.658E+8 gm Initial RCS Activities Iodines Pre-existing spike to Tech Spec limit of 60 gCi/gm Noble Gases 1% Failed Fuel Initial Secondary Side Activities lodines Pre-existing spike to Tech Spec limit of 0.10 gCi/gm Noble Gases 1% Failed Fuel Fuel Melted by Accident 0.25% of core Fuel Clad Damage 10% of core

NOC-AE-07002127 Attachment 1 Page 191 of 219 4.7 CONTROL ROD EJECTION ACCIDENT Table 4.7-6 Inputs for CREA Analysis Release from the RCB Scenario Parameter CLB AST Iodine Species Released to Containment 91/4/5 4.85/0.15/95 (elemental/organic/particulate)

Iodine Species Released From Containment'2 91/4/5 4.85/0.15/95 (elemental/organic/particulate)

Containment Free Volume For dilution of radionuclides 3.38E+6 ft3 3.38E+6 ft3 For leakage rate 3.41E+6 ft3 3.41E+6 ft3 Containment Leak Rate 0-24 hrs 0.3%/day 24hrs - 30 days 0.15%/day Dose Conversion Factors Table 4.2-6 Decay Constants and Decay Daughter Table 4.2-7 Fractions Offsite breathing rates Table 4.2-1 Offsite x/Q's Table 4.1-24 Control Room HVAC Parameters Table 4.2-3 Control Room HVAC Flow Rates Table 4.2-2 TSC HVAC Parameters Table 4.2-5 TSC HVAC Flow Rates Table 4.2-4 Control Room and TSC X/Q's Table 4.1-37 52 Containment sprays are not used in this analysis.

NOC-AE-07002127 Attachment 1 Page 192 of 219 4.7 CONTROL ROD EJECTIONACCIDENT Table 4.7-7 Inputs for CREA Analysis Release from the Secondary Side Scenario Parameter CLB AST Core power (for radiological source terms) 4100 MWt Core power level (for steam releases) 3876 MWt (3853MWt + 0.6%)

RCS density 8.33 Ibm/gallon RCS Mass 2.6E+8 gm 2.658E+8 gm SG Mass 6.59E+5 lbm 659,412 lbm Initial RCS Activities lodines Pre-existing spike to Tech Spec limit of 60 gCi/grn Noble Gases 1% Failed Fuel Initial Secondary Side Activities lodines Pre-existing spike to Tech Spec limit of 0.10 gCi/grn Noble Gases 1% Failed Fuel Primary-to-Secondary Leakage I gpm Fuel Melted by Accident 0.25% of core Fuel Clad Damage 10% of core Minimum time to release initial SG mass 191 seconds Steam Flow Rate to release initial SG mass 2.07E+5 lbm/min Maximum time for primary to secondary side 4500 seconds pressure equilibrium Steam Releases Table 4.7-5 Iodine Species Released to RCS 91%/4%/5% 4.85%/0.15%/95%

(elemental/organic/particulate)

Iodine Partition Factors for Releases from the 100/100/100 100/1/100 Secondary Side (elemental/organic/particulate)

Resulting lodifie Species Released from the 91/4/5 4.2/13.1/82.7"3 Secondary Side to Environment

% (elemental/organic/particulate)

Steam Flow rate 1.574E+7 lbm/hr Dose Conversion Factors Table 4.2-6 53 See Section 4.2.5

NOC-AE-07002127 Attachment 1 Page 193 of 219 4.7 CONTROL ROD EJECTION ACCIDENT Table 4.7-7 Inputs for CREA Analysis Release from the Secondary Side Scenario Parameter CLB AST Decay Constants and Decay Daughter Table 4.2-7 Fractions Offsite breathing rates Table 4.2-1 Offsite x/Q's Table 4.1-24 Control Room HVAC Parameters Table 4.2-3 Control Room HVAC Flow Rates Table 4.2-2 TSC HVAC Parameters Table 4.2-5 TSC HVAC Flow Rates Table 4.2-4 Control Room and TSC X/Q's Table 4.1-37 4.7.6 Summary and Conclusions Tables 4.7-8 through 4.7-10 below provides the analysis results. Per Standard Review Plan 15.4.8 (Reference 42), doses resulting from both release pathways are provided.

Table 4.7-8 CREA Doses from Containment Leakage (rem TEDE)

Receptor Dose EAB (worst 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) 0.86 LPZ 1.7 Control Room 2.4 TSC 2.3 Table 4.7-9 CREA Doses from Secondary Side Release (rem TEDE)

Receptor Dose EAB (worst 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) 0.55 LPZ 0.20 Control Room 0.41 TSC 0.40

NOC-AE-07002127 Attachment 1 Page 194 of 219 4.7 CONTROL ROD EJECTION ACCIDENT Table 4.7-10 Total CREA Dose Results (rem TEDE)

Receptor Dose Limits EAB (worst 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) 1.4 6.3 LPZ 1.9 6.3 Control Room 2.8 5 TSC 2.7 5 The actual doses for the CREA would be a composite of the doses computed for the independent release paths via the containment building and through the secondary system releases. The primary-to-secondary leakage used in the CREA analyses bound the primary-to-secondary leakage limit in the Technical Specifications. Since the doses resulting from the secondary side release path are below the acceptance criteria for the CREA, the Technical Specification limit on primary-to-secondary leakage is acceptable.

Also, the CREA analyses do not take credit for containment sprays to mitigate the release of radionuclides from the containment building. Since the doses resulting from the containment leakage pathway are below the acceptance criteria for the CREA, a reduction of the pressure setpoint for actuation of the containment sprays is not necessary to obtain credit for spray removal of fission products.

Radiological doses resulting from a design basis CREA for a Control Room operator and a person located at the EAB or LPZ are less than the regulatory dose limits as given in 10CFR50.67 for the Control Room and TSC and in 10CFR50.67, as modified by Regulatory Guide 1.183 in Table 6 on Page 1.183-20, for the EAB and LPZ.

NOC-AE-07002127 Attachment 1 Page 195 of 219 4.8 LOCKED ROTOR ACCIDENT 4.8 Locked Rotor Accident Radiological Assessment 4.8.1 Methodology Overview The Locked Rotor Accident (LRA) analysis postulates the instantaneous seizure of a reactor coolant pump (RCP) rotor, where the reactor is tripped on the subsequent low flow signal.

Following the trip, heat stored in fuel rods continues to pass into the reactor coolant, causing the coolant to expand. At the same time, heat transfer to the shell side of the SG is reduced, first because the reduced flow results in a decreased tube side film coefficient, and then because the reactor coolant in the tubes cools down while the shell side temperature increases (turbine steam flow is reduced to zero upon plant trip). The rapid expansion of the. coolant in the reactor core, combined with the reduced heat transfer in the SGs, causes an insurgence of coolant into the pressurizer and a pressure increase throughout the RCS. This insurgence into the pressurizer causes a pressure increase, which in turn actuates the automatic spray system, opens the pressurizer PORVs, and also opens the pressurizer safety valves.

The pressurizer PORVs are designed for reliable operation and would be expected to function properly during the accident. However, for conservatism, their pressure reducing effect and the pressure reducing effect of the spray is not included in the analysis.

This evaluation of the radiological consequences of a postulated seizure of a RCP rotor, i.e., an LRA, assumes that the reactor has been operating with a small percent of defective fuel (i.e., 1%)

and leaking SG tubes (1.0 gpm total). Tube uncovery, due to failure of a feedwater isolation valve, is assumed for the steam generator in the loop with the locked rotor with a primary-to-secondary leak rate of 0.35 gpm of the 1.0 gpm total. The reactor is assumed to have been operating in this condition for sufficient time to establish equilibrium concentrations of radionuclides in the reactor coolant and secondary coolant.

It is conservatively assumed that, as a result of the postulated LRA, 10% of the fuel rods in the core undergo sufficient clad damage to result in the release of their gap activity.

As a result of this accident, radionuclides carried by the primary coolant to the SGs, via leaking SG tubes, are released to the environment via SG PORVs. A failure of the feedwater system is assumed to occur which causes tube uncovery in one SG.

The LRA dose assessment is modeled to calculate the doses due to the activity that was instantaneously released into the RCS from the postulated damaged fuel fraction, and the activity resulting from a pre-accident 60 4Ci/gm DE 1-131 spike. Leakage and steaming rates through the SG PORVs are used to model the transport of activity from the RCS to the environment. Prior to the accident, a secondary coolant specific activity equal to the Technical Specification limit of 0.1 pCi/gm DE 1-131 equilibrium activity is assumed.

NOC-AE-07002127 Attachment 1 Page 196 of 219 4.8 LOCKED ROTOR ACCIDENT 4.8.1.1 Comparison of Modeling with Regulatory Guide 1.183, Appendix E Regulatory Guide 1.183 Appendix E, Position 5.5.1:

A portion of the primary-to-secondaryleakage willflash to vapor, based on the thermodynamic conditions in the reactorand secondary coolant.

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

• With regardto the unaffected steam generatorsusedforplantcooldown, the primary-to-secondaryleakage can be assumed to mix with the secondary water withoutflashing duringperiods of total tube submergence.

Treatment for the LRA analysis:

In one SG assumed to experience tube uncovery, all of the primary-to-secondary leakage is assumed to flash to vapor and be released to the environment with no mitigation. In the other SGs, the primary-to-secondary leakage is assumed to mix with the secondary water without flashing.

Regulatory Guide 1.183 Appendix E, Position 5.5.2:

The leakage that immediatelyflashes to vapor will rise through the bulk water of the steam generatorand enter the steam space. Credit may be taken for scrubbing in the generator,using the models in NUREG-0409, "Iodine Behavior in a PWR Cooling System Followinga PostulatedSteam GeneratorTube Rupture Accident" (Ref E-2), duringperiods of total submergence of the tubes.

Treatment for the LRA analysis:

This assumption is not used. It is assumed that the primary-to-secondary leakage does not flash in the SGs that do not experience tube uncovery. The nuclides in the primary-to-secondary leakage are added to the bulk fluid in these SGs.

Regulatory Guide 1.183 Appendix E, Position 5.5.3:

The leakage that does not immediately flash is assumed to mix with the bulk water.

Treatment for the LRA analysis:

It is assumed that the primary-to-secondary leakage does not flash in the SGs that do not experience tube uncovery. The nuclides in the primary-to-secondary leakage are added to the bulk fluid in the intact SGs.

NOC-AE-07002127 Attachment 1 Page 197 of 219 4.8 LOCKED ROTOR ACCIDENT Regulatory Guide 1.183 Appendix E, Position 5.5.4:

The radioactivityin the bulk water is assumed to become vapor at a rate that is the function of the steaming rate and the partitioncoefficient. A partition coefficient for iodine of 100 may be assumed. The retention ofparticulateradionuclidesin the steam generatorsis limited by the moisture carryoverfrom the steam generators.

Treatment for the LRA analysis:

A partition coefficient of 100 is assumed for elemental iodine released from the bulk fluid in the steam generators. Organic iodine is not partitioned. Organic iodine is assumed to migrate directly to the steam space and become immediately available for release.

Regulatory Guide 1.183 Appendix E, Position 5.6:

Operatingexperience and analyses have shown thatfor some steam generatordesigns, tube uncovery may occurfor a shortperiodfollowing any reactortrip (Ref E-3). The potential impact of tube uncovery on the transportmodel parameters(e.g., flashfraction, scrubbing credit) needs to be considered. The impact of emergency operatingprocedure restorationstrategieson steam generatorwater levels should be evaluated.

Treatment for the LRA analysis:

Tube uncovery does not occur following this event and the subsequent reactor trip.

However, a single failure in the feedwater system that isolated the feedwater entering one SG is assumed to occur. Tube uncovery is assumed for the steam generator in the loop with the feedwater isolation valve malfunction. Primary-to-secondary leakage from this steam generator is assumed to be 0.35 gpm of the total 1.0 gpm. This leakage is assumed to' flash and be immediately released to the environment.

4.8.2 Analytical Model The RADTRAD computer code is used to calculate the offsite, Control Room, and TSC doses.

• The first model (A, in Figure 4.8-1) calculates the dose resulting from secondary side steam releases due to primary-to-secondary leakage to the three steam generators that have covered tubes. The second model (B) calculates the dose due to primary-to-secondary leakage in the steam generator with uncovered tubes. Section 5.5.1 in Appendix E Regulatory Guide 1.183 states that during periods of steam generator dryout; all primary-to-secondary leakage will flash and be released directly to the environment. The release paths are shown below.

NOC-AE-07002127 Attachment 1 Page 198 of 219 4.8 LOCKED ROTOR ACCIDENT Figure 4.8-1: LRA RADTRAD Model 10% of Total Gap Activity + 60 iCi/gm 0.65 gpm A. i Reactor >Cs.DEI-13l + RCS and Rb Kr, Xe, R 1e0 Leakage PORVs 10% of Total Gap Activity + 60 jtCi/gm 0.35 gpm B. DEI-131 + RCS Kr, Xe, Racr b6a Leakage or Cs, and Rb 4.8.3 Radiological Source Term For conservatism, the LRA core source terms are those associated with a DBA power level of 4100 MWth, which is greater than the RTP of 3853 MWth plus a 0.6% measurement uncertainty.

The instantaneous seizure of the RCP rotor associated with the LRA results in a small percentage of fuel damage. The dose analysis for this event conservatively assumes 10% fuel damage. The design basis of this accident assumes that no fuel melt is postulated to occur. Therefore, the source term available for release is associated with this fraction of damaged fuel and the fraction of core activity existing in the gap, plus the iodine in the RCS due to a design basis pre-accident 60 pCi /grn DE 1-131 spike, and the noble gas activity associated with assumed 1% fuel defects.

Release fractions and transport fractions are consistent with Regulatory Guide 1.183, Appendix G and Table 3. To conform with this regulatory guidance, 5% of the core inventory of iodine and noble gas is assumed to be in the fuel-clad gap, excluding 1-131 and Kr-85, where 8% and 10%

are assumed, respectively. Additionally, Table 3 of Regulatory Guide 1.183 shows that 12% of the core cesium and rubidium should be assumed to be in the fuel-clad gap.

The source term model also consists of the 0.1 ptCi/grn DE 1-131 equilibrium secondary coolant activity concentration, consistent with the TS requirements.

NOC-AE-07002127 Attachment 1 Page 199 of 219 4.8 LOCKED ROTOR ACCIDENT 4.8.3.1 Reactor Coolant System Source Term 4.8.3.1.1 RCS Iodine Concentrations The RCS iodine concentrations for a pre-existing iodine spike to 60 pCi/gm are given in Table 4.2-17.

4.8.3.1.2 RCS Noble Gas Concentrations The RCS noble gas concentrations for 1% failed fuel are given in Table 4.2-14.

4.8.3.1.3 RCS Cesium and Rubidium Concentrations No spiking of Cs or Rb is assumed. The RCS Cs and Rb concentrations corresponding to 1%

failed fuel are used (Table 4.2-20).

4.8.3.2 Secondary System Source Terms 4.8.3.2.1 Secondary System Iodine Concentrations The secondary systems iodine concentrations corresponding to the Technical Specification limit of 0.10 gCi/gm are given in Table 4.2-19.

4.8.3.2.2 Secondary System Noble Gas Concentrations The secondary systems noble gas concentrations corresponding to 1.0% failed fuel are given in Table 4.2-20.

4.8.3.2.3 Secondary System Cesium and Rubidium Concentrations The secondary system Cs and Rb concentrations corresponding to 1% failed fuel are used (Table 4.2-20). Cesium and rubidium are assumed to be bound with iodines as particulates. Therefore, there is no release of Cs or Rb from water in the steam generators.

4.8.3.3 Fuel Pin Gap Source The accident release inventory is derived from the core isotopic inventory. This inventory is corrected to the total gap inventory in order to calculate the release from a failure of 10% of the fuel rods. The gap fractions are from Regulatory Guide 1.183, Table 3. The gap release source is presented in Table 4.8-1.

NOC-AE-07002127 Attachment 1 Page 200 of 219 4.8 LOCKED ROTOR ACCIDENT Table 4.8-1 10% Gap Release Source (Ci)

Isotope CLB AST  % Difference 1-131 1.1OE+06 8.5E+05 -22.7%

1-132 1.60E+06 7.6E+05 -52.5%

1-133 2.30E+06 1.1E+06 -52.2%

1-134 2.50E+06 1.2E+06 -52.0%

1-135 2.1OE+06 1.OE+06 -52.4%

Kr-83m 1.40E+05 7.OE+04 -50.0%

Kr-85m 3.OOE+05 1.5E+05 -50.0%

Kr-85 3.70E+04 1 2E+04 -67.6%

Kr-87 5.50E+05 2.8E+05 -49.1%

Kr-88 7.89E+05 3.9E+05 -50.6%

Kr-89 9.70E+05 4.8E+05 -50.5%

Rb-86 0 Rb-88 9.5E+05 Rb-89 1.2E+06 Xe-131m 7.70E+03 5.5E+03 -28.6%

Xe-133m 3.30E+05 3.4E+04 -89.7%

Xe-133 2.30E+06 1.1 E+06 -52.2%

Xe-135m 4.60E+05 2.1E+05 -54.3%

Xe-135 6.50E+05 2.8E+05 -56.9%

Xe-137 2.OOE+06 9.5E+05 -52.5%

Xe-138 1.90E+06 9.OE+05 -52.6%

Cs-134 2.6E+05 Cs-136 7.6E+04 Cs-137 1.6E+05 Cs-138 2.4E+06 The values in Table 4.8-1 are based on the reactor core sources in Table 4.2-9. The AST values use the gap fractions from Regulatory Guide 1.183, Table 3.

NOC-AE-07002127 Attachment 1 Page 201 of 219 4.8 LOCKED ROTOR ACCIDENT 4.8.3.4 Total Source Available for Release The total source available for release is presented in Table 4.8-2.

Table 4.8-2 Total Source Available for Release (RCS+Sec)

(Ci)

Isotope CLB AST  % Difference 1-131 1.11E+06 8.6E+05 -22.5%

1-132 1.61E+06 7.8E+05 -51.6%

1-133 2.32E+06 1.1E+06 -52.6%

1-134 2.50E+06 1.2E+06 -52.0%

1-135 2.11E+06 1.1E+06 -47.9%

Kr-83m 1.40E+05 7.0E+04 -50.0%

Kr-85m 3.OOE+05 1.5E+05 -50.0%

Kr-85 3.91E+04 1.4E+04 -64.2%

Kr-87 5.50E+05 2.8E+05 -49.1%

Kr-88 7.91E+05 3.9E+05 -50.7%

Kr-89 9.70E+05 4.8E+05 -50.5%

Rb-86 4.5E+00 Rb-88 9.5E+05 Rb-89 1.2E+06 Xe-131m 8.21E+03 6.2E+03 -24.5%

Xe-133m 3.34E+05 3.5E+04 -89.5%

Xe-133 2.36E+06 1.2E+06 -49.2%

Xe-135m 4.60E+05 2.1E+05 -54.3%

Xe-135 6.52E+05 2.8E+05 -57.1%

Xe-137 2.OOE+06 9.5E+05 -52.5%

Xe-138 1.90E+06 9.OE+05 -52.6%

Cs-134 2.6E+05 Cs-136 7.7E+04 Cs-137 1.6E+05 Cs-138 2.4E+06 The chemical form of the iodine in the RCS is 95% CsI, 4.85% elemental, and 0.15% organic.

The chemical form of the iodine released from the secondary side is 4.2% elemental, 13.1%

organic, and 82.7% particulate (Section 4.2.5).

NOC-AE-07002127 Attachment 1 Page 202 of 219 4.8 LOCKED ROTOR ACCIDENT 4.8.4 Radiological Releases Activity that originates in the RCS is released to the secondary coolant by means of the primary-to-secondary coolant leak rate. This design basis leak rate value is 1.0 gpm for all SGs. For the SG on the loop with the locked RCP rotor, tube uncovery is assumed due to a feedwater isolation valve malfunction. Primary-to-secondary leakage from this steam generator is assumed to be 0.35 gpm of the total.

Releases to the environment are associated with the secondary coolant steaming from the SGs.

Because of the release dynamic of the activity from the SG PORVs, Regulatory Guide 1.183 allows for a reduction in the amount of activity released to the environment based on partitioning of nuclides between the liquid and gas states of water for this release path. For iodine, the partition factor of 100 was taken directly from the suggested guidance. No particulates are assumed to be released. Because of their volatility, 100% of the noble gases are assumed to be released.

The methodology used to model steaming of activity through intact SG PORVs following the postulated LRA event assumes an average cumulative release rate through the SG PORVs that, for simplicity and conservatism, is reduced in steps. The steaming release from the PORVs and primary-to-secondary coolant leakage is postulated to end at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, when the RCS and secondary loop have reached equilibrium. Leakage via the MSIV above-seat drain orifices is assumed to continue for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Table 4.8-3 below shows the time steps and associated release rates.

Table 4.8-3 Steam Released to the Environment (Ibm)

CLB AST Time T Above Seat (Hours) PORV PORV Drains Total  % Difference 0-2 455,047 640,000 55,584 695,584 52.9%

2- 8 1,137,757 1,120,000 166,752 1,286,752 13.1%

8-12 0 0 111,168 111,168 -

12-36 0 0 667,008 667,008 -

36-720 0 0 0 0 -

A constant 7.72 lbm/sec leakage from the MSIV above seat drain orifices is assumed in the revised analyses (0 lbm/sec was assumed in the CLB).

NOC-AE-07002127 Attachment 1 Page 203 of 219 4.8 LOCKED ROTOR ACCIDENT 4.8.5 Assumptions and Inputs The following inputs and assumptions were used in the LRA analysis.

1. The source term is based upon a power level of 4100 MW thermal, 5 w/o enrichment, and a three region core with equilibrium cycle core at end of life. The three regions have operated at a specific power of 39.3 MW/MTU for 509, 1018, and 1527 EFPD, respectively. The assumed power level is greater than the Rated Thermal Power of 3853 MWth plus a 0.6%

measurement uncertainty.

2. The initial activity in the reactor coolant is based upon an iodine spike which has raised the reactor coolant concentration to 60 [tCi/grn of dose equivalent 1-131. Noble gas activity is based on 1% failed fuel.

3. Prior to the accident, the secondary coolant specific activity is equal to the Technical Specification limit of 0.10 ptCi/gm dose equivalent 1-131. This DEI activity is given in Table 4.2-19.

4. Ten percent (10%) fuel failure is assumed to occur. The activity released from the pellet-to-clad gap of the failed fuel is assumed to be instantaneously mixed with the reactor coolant system, per Regulatory Guide 1.183. No fuel melting occurs.

5. A feedwater system malfunction caused by the closure of a feedwater isolation valve is postulated, resulting in tube uncovery in that SG.

6. Primary-to-secondary leakage through the steam generator tubes prior to the accident and during the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the transient is 1 gpm. Eight hours after the accident, the residual heat removal system starts and primary-to-secondary leakage is stopped. Primary-to-secondary leakage is conservatively modeled at 0.65 gpm for the three steam generators with covered tubes and at 0.35 gpm for the steam generator with uncovered tubes.

7. A partition coefficient of 100 is assumed for elemental iodine released from the steam generators. (Regulatory Guide 1.183, Appendix G, Position 5.5.4) Organic iodine is not partitioned. Organic iodine is assumed to migrate directly to the steam space and become immediately available for release.

8. The Control Room and TSC ventilation systems are assumed to transfer to the emergency mode of operation immediately after the initiation of this accident. This assumption is countered by the assumption of an additional (second) single failure of a train of the Control Room Emergency HVAC system, specifically the clean-up (recirculation) filters.

9. Offsite power is lost; Main Steam condensers are not available for steam dump.

10. Eight hours after the accident, the residual heat removal system is in operation and no further steam containing radionuclides are released from steam generators to the environment except the leakage through the MSIV above seat drain orifices. The release through the orifices continues until 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after the start of the accident.

11. All releases occur via the PORVs or safeties and the above-seat drain orifices in the Isolation Valve Cubicle next to the PORVs. Therefore, the PORV-to-Control Room X/Qs are used for the Control Room and TSC dose analyses.

12. The reactor coolant density is 8.33 lbm/gal (14.7 psia, 70 'F).

NOC-AE-07002127 Attachment 1 Page 204 of 219 4.8 LOCKED ROTOR ACCIDENT Input parameters used for the LRA analysis are given in Table 4.8-4. Conformance with Regulatory Guide 1.183 guidance addressing LRA analysis is provided in Attachment 6, Tables A and G.

Table 4.8-4 Inputs for LRA Analysis Parameter CLB AST Core power (for radiological source terms) 410OMWt Core power (for steamreleases) 3876 MWt (3853MWt + 0.6%)

RCS density 8.33 Ibm/gallon RCS Mass 2.658E+8 gm SG Mass 659,412 Ibm 2.991E+08 gm Primary-to-Secondary Leakage SGs w/o tube uncovery 1.0 gpm 0.65 gpm SG w/tube uncovery N/A 0.35 gpm Release from SGs Table 4.8-3 Release from Above (MSIV) Seat Drains SGs w/o tube uncovery N/A 5.79 lbm/sec SG w/tube uncovery N/A 1.93 lbm/sec Steam Flow rate 1.574E+7 Ibm/hr Iodine Partition Factors for Releases from the Secondary Side (elemental/organic/particulate)

Resulting Iodine Species Released from the Secondary Side to Environment 91/4/5 4.2/13.1/82.754

% (elemental/organic/particulate) I Dose Conversion Factors Table 4.2-6 Decay Constants and Decay Daughter Table 4.2-7 Fractions Offsite breathing rates Table 4.2-1 Offsite x/Q's Table 4.1-24 Control Room HVAC Parameters Table 4.2-3 Control Room HVAC Flow Rates Table 4.2-2 54 See Section 4.2.5

NOC-AE-07002127 Attachment 1 Page 205 of 219 4.8 LOCKED ROTOR ACCIDENT Table 4.8-4 Inputs for LRA Analysis Parameter CLB AST TSC HVAC Parameters Table 4.2-5 TSC HVAC Flow Rates Table 4.2-4 Control Room and TSC X/Q's Table 4.1-37 4.8.6 Summary and Conclusions Radiological doses resulting from a design basis LRA for a Control Room operator and a person located at the EAB or LPZ are to be less than the regulatory dose limits as given in 10CFR50.67 for the Control Room and TSC and in 10CFR50.67, as modified by Regulatory Guide 1.183 in Table 6 on Page 1.183-20, for the EAB and LPZ.

Table 4.8-5 provides the results for the LRA analysis.

Table 4.8-5 LRA Dose Results (rem TEDE)

Receptor Dose Limits EAB (worst 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) 1.9 2.5 LPZ 1.5 2.5 Control Room 3.9 5 TSC 3.7 5 These calculated doses above are well below their respective acceptance criteria, so it is verified that the LRA is sufficiently mitigated.

NOC-AE-07002127 Attachment 1 Page 206 of 219 4.9 NUREG-0737 Evaluations As part of the DBA LOCA analysis, radiation levels from contained sources (containment structure and Control Room and Technical Support Center (TSC) filters) were evaluated. These evaluations were used to determine if an impact on the following areas covered by NUREG-0737 would occur as a result of an increase in the associated radiation levels:

• CLB radiological dose analyses for post-accident vital area access and post-accident sampling (NUREG-0737, Item II.B.2 and Item II.B.3),

" CLB radiological dose analyses for the post-accident containment high range radiation monitors (NUREG-0737, Item II.F.1), and

• CLB control room post-accident radiological dose analyses for emergency support facility upgrades and control room habitability (NUREG-0737, Items III.A.1.2 and III.D.3.4).

Evaluations Post Accident Vital Area Access and Sampling - Post-accident personnel missions resulting in mission doses (including post-accident sampling) have been previously identified. The implementation of the AST methodology does not result in any new operator missions. Plant calculations used in support of plant post-accident vital area access (prepared in accordance with NUREG-0737, Items II.B.2 and II.B.3) were judged to be unaffected based on an assessment of AST vs. TID-14844 contained sources.

The results of the assessment of post-accident shine due to contained sources in various plant locations is that the current calculated doses (based on TID-14844 source terms) bound the corresponding doses that would be calculated based on the AST. This conclusion is reached on the basis of (1) a comparison of the post-LOCA containment airborne source terms (MeV/sec as a function of time for different photon energy groups) using STPEGS-specific airborne activity removal rates, (2) a general comparison of the potential for post-LOCA waterborne source terms (total MeV/sec as a function of time as well as MeV/sec as functions of time for photons with energies greater than 1.5 MeV),

and (3) a comparison of the post-SGTR source term (MeV/sec for different photon energy groups) for the location and time where post-SGTR access would be required.

" Post Accident Sampling System - The requirements of NUREG 0737 for Post Accident Sampling System (PASS) were deleted as part of Amendment No. 133 to Facility Operating License No. NPF-76 and Amendment No. 122 to Facility Operating License No. NPF-80 issued November 7, 2001 via Document ST-AE-NOC-01000894 South Texas Projects, Units 1 and 2 - Issuance of Amendments on the Elimination of Requirements for Post Accident Sampling (TAC Nos. MB2900 and MB2904).

• Post-Accident Radiation Monitor - The CLB analysis for the containment high range radiation monitors used to monitor post-accident primary containment radiation levels

NOC-AE-07002127 Attachment 1 Page 207 of 219 use a source term different from either TID-14844 or the AST. Therefore, there is no impact of AST implementation on the containment high range monitor evaluation.

(NUREG-0737, Item II.F.1)

" Control Room Radiation Protection - The doses to Control Room operators were specifically calculated using AST for the Design Basis Accidents described in this submittal. Results are presented with each respective accident description. (NUREG-0737, Item III.D.3.4).

" Technical Support Center Radiation Protection - The doses to TSC personnel were specifically calculated using AST for the Design Basis Accidents described in this submittal. Results are presented with each respective accident description. (NUREG-0737, Item III.A. 1.2).

" Radioactive Sources Outside the Primary Containment - The DBA LOCA Control Room/TSC dose analysis, as well as that for offsite doses, considers the effects of ESF leakage outside the primary containment and (for the Control Room and TSC dose analyses only) the shine contribution from the containment and other source term bearing systems and/or components (NUREG - 0737, Item III.D. 1.1).

4.10 Conclusion The proposed changes provide a source term for STP that will result in a more accurate assessment of the DBA radiological doses. The revised radiological dose to the control room operator allows for. a revised air unfiltered in-leakage assumption that provides a conservative margin over that determined by air in-leakage testing. Changes related to the applicability requirements during movement of irradiated fuel assemblies use insights from TSTF-51.

Adequate defense-in-depth is maintained by the requirements for radioactive decay and water level Technical Specification requirements for systems needed for decay heat removal, or to mitigate potential reactor vessel drain down events, or the requirements to maintain high water levels over irradiated fuel are not impacted by the proposed amendment.

The proposed amendment also deletes the APPLICABILITY requirements of CORE ALTERATIONS for selected TS, since the only accident postulated to occur during CORE ALTERATIONS that results in radioactive release is the fuel handling accident.

Shutdown safety controls are provided during periods when fuel is being handled. These controls address (1) procedures to assess the impact of removing systems from service during shutdown conditions, (2) the ability to implement prompt methods to close both the Reactor Containment Building and/or the Fuel Handling Building(s) in the event of a FHA, and (3) controls to avoid unmonitored releases.

Implementation of the AST as the plant radiological consequence analyses licensing basis requires a licensing amendment request pursuant to the requirements of 10 CFR 50.67.

Radiological dose analyses were performed for the DBA LOCA, FHA, MSLB, SGTR, CREA, and LRA using conservative assumptions. Doses calculated with the AST for accidents

NOC-AE-07002127 Attachment 1 Page 208 of 219 involving damaged fuel reflect delayed and/or reduced activity releases (relative to those of the TID-14844-based CLB) to the containment and to the FHB, as applicable. Offsite, Control Room, and TSC doses remain well below regulatory requirements.

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration 5.1.1 Overview On December 23, 1999, the NRC issued the Final Rule on "Use of Alternate Source Terms at Operating Reactors." The Final Rule, issued under 10 CFR 50.67, "Accident Source Term",

allows holders of operating licenses issued prior to January 10, 1997, to voluntarily replace the traditional source term used in design basis accident analyses with alternative source terms. This action would allow interested licensees to pursue cost beneficial licensing actions to reduce unnecessary regulatory burden without compromising the margin of safety of the facility.

Based on the above rule and in accordance with 10 CFR 50.67 and 10 CFR 50.90, "Application for amendment of license or construction permit." STPNOC is requesting an amendment to Appendix A, Technical Specifications (TS), of Facility Operating License Nos. NPF-76 and NPF-80 for STP, Units 1 and 2. The proposed changes are requested to support application of an alternative source term (AST) methodology, with the exception that Technical Information Document (TID) 14844, "Calculation of Distance Factors for Power and Test Reactor Sites," will continue to be used as the radiation dose basis for equipment qualification. The proposed AST methodology conforms to the guidance in Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,"

dated July 2000, except where alternate methods for complying with the specified portions of the NRC's regulations have been used as allowed by RG 1.183. The AST analyses were also performed in accordance with the guidance in Standard Review Plan Section 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms."

In support of a full-scope implementation of the AST methodology, STP has performed radiological consequence analyses for the following six design basis accidents (DBAs) that result in control room and offsite exposure as specified in RG 1.183.

" Loss of Coolant Accident (LOCA)

• Fuel Handling Accident (FHA)

" Main Steam Line Break (MSLB)

" Steam Generator Tube Rupture (SGTR)

• Control Rod Ejection Accident (CREA)

• Locked Rotor Accident (LRA)

NOC-AE-07002127 Attachment 1 Page 209 of 219 The proposed changes related to the applicability requirements during movement of irradiated fuel assemblies are based on insights from Technical Specification Task Force Traveler (TSTF)-

51, "Revise Containment Requirements During Handling of Irradiated Fuel and Core Alterations," Revision 2. The NRC approved TSTF-51 on July 31, 2003. TSTF-51 changes the TS operability requirements for engineered safety features such that they are not required to be operable after sufficient radioactive decay has occurred to ensure that offsite doses remain within limits.

Proposed changes to the current licensing basis, justified by the AST analyses, include the following items:

• The use of updated meteorological data to calculate onsite and offsite atmospheric dispersion

" Relies on less filtration

> No credit taken for Fuel Handling Building Exhaust Air Ventilation filtration

> No credit taken for Control Room Ventilation makeup filtration

> No credit taken for. either Control Room Ventilation makeup or recirculation cleanup filtration for the Fuel Handling Accident 0 Containment isolation capability is no longer required to mitigate a FHA

• Analysis of only a single limiting FHA rather than one analysis for an FHA inside containment and a second analysis for an FHA in the fuel handling building (FHB)

• Revised control room unfiltered in-leakage assumption.

5.1.2 Criteria According to 10 CFR 50.92, "Issuance of amendment," paragraph (c), a proposed amendment to an operating license 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 a margin of safety.

In support of this determination, an evaluation of each of the three criteria set forth in 10 CFR 50.92 is provided below regarding the proposed license amendment.

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

The implementation of AST assumptions has been evaluated in revisions to the analyses of the following limiting DBAs.

" Loss-of-Coolant Accident

" Fuel Handling Accident

" Control Rod Ejection Accident

NOC-AE-07002127 Attachment 1 Page 210 of 219

• Locked Rotor Accident

• Main Steam Line Break Accident

• Steam Generator Tube Rupture Accident Based upon the results of these analyses and evaluations, it has been demonstrated that, with the requested changes, the dose consequences of these limiting events satisfies the dose limits in 10 CFR 50.67 and are within the regulatory guidance provided by the NRC for use with the AST methodology. The AST is an input to calculations used to evaluate the consequences of an accident and does not affect the plant response or the actual pathway of the activity released from the fuel. Therefore, it is concluded that AST does not involve a significant increase in the consequences of an accident previously evaluated.

Implementation of AST provides for elimination of the Fuel Handling Building ventilation system filtration TS requirements and elimination of Control Room ventilation filtration TS requirements in Modes 5 or 6. It also eliminates containment integrity TS requirements while handling irradiated fuel and during core alterations. The equipment affected by the proposed changes is mitigative in nature and relied upon after an accident has been initiated. The affected systems are not accident initiators; and application of the AST methodology is not an initiator of a design basis accident.

Elimination of the requirement to suspend operations involving positive reactivity additions that could result in loss of required SHUTDOWN MARGIN or required boron concentration if the control room ventilation system is inoperable in Modes 5 or 6 does not increase the probability of an accident because the proposed change does not affect the design and operational controls to prevent dilution events. These same design and operational controls prevent a loss of SHUTDOWN MARGIN or a boron dilution event so that radiological consequences from these events are precluded.

The proposed changes do not involve physical modifications to plant equipment and do not change the operational methods or procedures used for moving irradiated fuel assemblies. The proposed changes do not affect any of the parameters or conditions that could contribute to the initiation of any accidents. Relaxation of operability requirements during the specified conditions will not significantly increase the probability of occurrence of an accident previously analyzed. Since design basis accident initiators are not being altered by adoption of the AST, the probability of an accident previously evaluated is not affected.

Administrative changes to delete a footnote from Technical Specification surveillance requirement 4.7.7.e.3) and a note from ACTION 20 of Technical Specification Table 3.3-3, in which the provisions of the notes have expired, does not impact the probability or consequences of an accident previously evaluated.

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

NOC-AE-07002127 Attachment 1 Page 211 of 219

2. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated The proposed changes do not involve a physical change. The change will allow the automatic start feature of systems no longer credited in the accident analyses for mitigation to be disabled through the STPNOC modification process. Implementation of AST provides increased operating margins for filtration system efficiencies. Application of AST provides for relaxation of certain Control Room ventilation system filtration requirements. The Fuel Handling Building filtration and holdup is no longer credited in the AST analyses. Therefore, the Fuel Handling Building Exhaust Air Ventilation system is no longer required in the Technical Specifications. It also relaxes containment integrity requirements while handling irradiated fuel and during core alterations.

Elimination of the requirement to suspend operations involving positive reactivity additions that could result in loss of required SHUTDOWN MARGIN or required boron concentration if the control room ventilation system is inoperable in Mode 5 or Mode 6 does not create the possibility of a new or different kind of accident because these events have already been analyzed in the safety analysis with a conclusion that adequate measures exist to prevent these events.

Similarly, the proposed changes do not require any physical changes to any structures, systems or components involved in the mitigation of any accidents. Therefore, no new initiators or precursors of a new or different kind of accident are created. New equipment or personnel failure modes that might initiate a new type of accident are not created as a result of the proposed changes.

Administrative changes to delete a footnote from Technical Specification surveillance requirement 4.7.7.e.3) and a note from ACTION 20 of Technical Specification Table 3.3-3, in which the provisions of the notes have expired, does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

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

Approval of a change from the original source term methodology (i.e., TID 14844) to an AST methodology, consistent with the guidance in RG 1.183, will not result in a significant reduction in the margin of safety. The safety margins and analytical conservatisms associated with the AST methodology have been evaluated and were found acceptable. The results of the revised DBA analyses, performed in support of the proposed changes, are subject to specific acceptance criteria as specified in RG 1.183.

The dose consequences of these DBAs remain within the acceptance criteria presented in 10 CFR 50.67 and RG 1.183.

NOC-AE-07002127 Attachment. 1 Page 212 of 219 Elimination of the requirement to suspend operations involving positive reactivity additions that could result in loss of required SHUTDOWN MARGIN or required boron concentration if the control room ventilation system is inoperable in Mode 5 or Mode 6 does not result in a reduction in a margin to safety because adequate measures exist to preclude radiological consequences from these events.

The proposed changes continue to ensure that the doses at the exclusion area boundary (EAB) and low population zone boundary (LPZ), as well as the Control Room and Technical Support Center, are within the specified regulatory limits.

Administrative changes to delete a footnote from Technical Specification surveillance requirement 4.7.7.e.3) and a note from ACTION 20 of Technical Specification Table 3.3-3, in which the provisions of the notes have expired, does not impact the margin of safety.

Therefore, based on the above discussion, the proposed changes do not involve a significant reduction in a margin of safety.

Conclusion Based on the above discussion, it has been determined that the requested TS changes do not involve a significant increase in the probability or consequences of an accident previously evaluated; or create the possibility of a new or different kind of accident from any accident previously evaluated; or involve a significant reduction in a margin of safety. Therefore, the requested license amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c).

5.2 Applicable Regulatory Requirements/Criteria The NRC's traditional methods for calculating the radiological consequences of design basis accidents (i.e., prior to adopting the AST methodology) are described in a series of Regulatory Guides (RGs) and Standard Review Plan (SRP) chapters. That guidance was developed to be consistent with the TID-14844 source term and the whole body and thyroid dose guidelines stated in 10 CFR 100.11. Many of those analysis assumptions and methods are inconsistent with the AST methodology and with the Total Effective Dose Equivalent (TEDE) criteria provided in 10 CFR 50.67. RG 1.183 provides assumptions and methods that are acceptable to the NRC staff for performing design basis radiological analyses using an AST approach. This guidance supersedes corresponding radiological analysis assumptions provided in the previous Regulatory Guides and SRP chapters when used in conjunction with an approved AST methodology and the TEDE criteria provided in 10 CFR 50.67.

Due to the comprehensive nature of RG 1.183, Attachment 6, "Regulatory Guide Conformance Tables," were developed to show how each section of the RG 1.183 guidance is being addressed.

NOC-AE-07002127 Attachment 1 Page 213 of 219 The NRC also published a new SRP section to address AST; i.e., SRP Section 15.0.1, Revision 0, "Radiological Consequence Analyses Using Alternative Source Terms." This SRP section is consistent with the guidance found in RG 1.183. The plant-specific information provided in this license amendment request is also consistent with the guidance found in SRP 15.0.1.

10 CFR 50.36, "Technical Specifications 10 CFR 50.36 specifies the items that should be included in the Technical Specifications.

Specifically, Part 50.36(c)(2)(ii) states that a technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

(a) Criterion1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

The systems proposed for removal from the Technical Specifications are not used to detect degradation of the reactor coolant pressure boundary. The systems are not an initial condition of a design basis accident or transient analysis that either assumes the failure of, or presents a challenge to the integrity of a fission product barrier.

(b) Criterion2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The systems proposed for removal from the Technical Specifications are not process variables, design features, or pose any operating restrictions that are an initial condition of a design basis accident or transient analysis that either assumes the failure of, or presents a challenge to the integrity of a fission product barrier.

(c) Criterion3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The results of the revised accident analyses based on the alternative source term no longer credits the items proposed for removal from the Technical Specifications as accident mitigation features. The items proposed for removal from the Technical Specifications do not present a challenge to the integrity of a fission product barrier. These systems are not primary success path for mitigation of the DBA.

NOC-AE-07002127 Attachment 1 Page 214 of 219 (d) Criterion4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The requirements proposed for relocation from the TS, in the Modes specified, do not contribute to the conditional probability of core damage or conditional probability of a large release. The requirements being relocated do not contain constraints of prime importance in limiting the likelihood or severity of the accident sequences that are commonly found to dominate risk. The operability of the system is not risk significant.

Safety Margins and Defense in Depth Regulatory Guide 1.183 states that proposed uses of an AST and the associated proposed facility modifications and changes to procedures should be evaluated to determine whether the proposed changes are consistent with the principle that sufficient safety margins are maintained, including a margin to account for analysis uncertainties. Specific values and limits contained in the technical specifications and the response times for the safety system assumed in the accident analyses are not changed. Caution has been taken to ensure that the dose analyses have not been "tuned" to a specific set of accident progression assumptions so that the assumptions remain conservative for the accident sequences considered. The dose consequence results of the accident analyses remain well below regulatory limits.

Regulatory Guide 1.183 states that proposed uses of an AST and the associated proposed facility modifications and changes to procedures should be evaluated to determine whether the proposed changes are consistent with the principle that adequate defense in depth is maintained to compensate for uncertainties in accident progression and analysis data. System redundancy, independence, and diversity features -are not changed for those safety systems credited in the accident analyses. No new programmatic compensatory activities or reliance on manual operator actions is required to implement this change. For those systems that are no longer credited in the accident analyses for mitigation, programmatic controls will be used to provide an additional layer of defense-in-depth to align these systems to ensure that any release from a fuel handling accident will be filtered (not required to be met to meet the dose consequence results) and monitored.

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

NOC-AE-07002127 Attachment 1 Page 215 of 219

6.0 ENVIRONMENTAL CONSIDERATION

On December 23, 1999, the NRC issued the Final Rule on "Use of Alternate Source Terms at Operating Reactors." The Final Rule, issued under 10 CFR 50.67, "Accident source term, "allows holders of operating licenses issued prior to January 10, 1997, to voluntarily replace the traditional source term used in design basis accident analyses with alternative source terms. This action would allow interested licensees to pursue cost beneficial licensing actions to reduce unnecessary regulatory burden without compromising the margin of safety of the facility.

Based on the above rule and in accordance with 10 CFR 50.67 and 10 CFR 50.90, "Application for amendment of license or construction permit," STPNOC is requesting an amendment to Appendix A, Technical Specifications (TS), of Facility Operating License Nos. NPF-76 and NPF-80 for South Texas Project, Units 1 and 2. The proposed changes are requested to support application of an alternative source term (AST) methodology, with the exception that Technical Information Document (TID) 14844, "Calculation of Distance Factors for Power and Test Reactor Sites," will continue to be used as the radiation dose basis for equipment qualification.

The proposed AST methodology conforms to the guidance in Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," dated July 2000, except where alternate methods for complying with the specified portions of the NRC's regulations have been used as allowed by RG 1.183. The AST analyses were also performed in accordance with the guidance in Standard Review Plan Section 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms."

STPNOC has evaluated the proposed changes against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21, "Criteria for and identification of licensing and regulatory actions requiring environmental assessments." STPNOC has determined that the proposed changes meet the criteria for a categorical exclusion as set forth in 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review," paragraph (c)(9), and as such, has determined that no irreversible consequences exist in accordance with 10 CFR 50.92, "Issuance of amendment,"

paragraph (b). This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10. CFR 50, "Domestic Licensing of Production and Utilization Facilities," which changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20," Standards for Protection Against Radiation," or that changes an inspection or a surveillance requirement, and the amendment meets the following specific criteria.

(i) The amendment involves no significant hazards consideration.

As demonstrated in Section 5.1 above, the proposed changes do not involve a significant hazards consideration.

NOC-AE-07002127 Attachment 1 Page 216 of 219 (ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

STPNOC meets the radiological criteria described in 10 CFR 50.67 for the exclusion area boundary (EAB) and the low population zone (LPZ).

Adoption of the AST methodology and TS changes which implement certain conservative assumptions in the AST analyses will not result in physical changes to the plant that could significantly alter the type or amounts of effluents that may be released offsite. Changes to operational parameters that could affect effluent releases have been demonstrated through analysis to satisfy regulatory requirements.

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

STPNOC meets the radiological criteria described in 10 CFR 50.67 for the Control Room and Technical Support Center. Control Room and Technical Support Center exposure to operators is less than the five rem total effective dose equivalent (TEDE) over 30 days for all accidents.

The implementation of the AST methodology has been evaluated in revisions to the analyses of the limiting design basis accidents at the South Texas Project, Units 1 and 2.

These accidents include the loss of coolant accident, the fuel handling accident, the control rod ejection accident, locked rotor accident, main steam line break accident, and steam generator tube rupture accident. Based upon the results of these analyses, it has been demonstrated that, with the proposed changes, the dose consequences of these limiting events are within the regulatory guidance provided by the NRC for use with the alternative source term approach (i.e., 10 CFR 50.67 and RG 1.183). Thus, there will be no significant increase in either individual or cumulative occupational radiation exposure.

NOC-AE-07002127 Attachment 1 Page 217 of 219

7.0 REFERENCES

1. Technical Information Document (TID) - 14844, "Calculation of Distance Factors for Power And Test Reactor Sites," U.S. Atomic Energy Commission, March 23, 1962.

2. NUREG-0800, Standard Review Plan, SRP-15.0.1, Rev. 0, "Radiological Consequence Analyses Using Alternative Source Terms," USNRC.

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

4. [Not Used]

5. Technical Specification Task Force Traveler, TSTF-5 1, Revision 2, dated July 31, 2003, "Revise Containment Requirements During Handling of Irradiated Fuel and Core Alterations."

6. Nuclear Utilities Management and Resources Council (NUMARC) 93-01, "Industry Guidelines for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants,"

Revision 3.

7. South Texas Project, Units 1 and 2 - Issuance of Amendments on Relocation of Various Technical Specifications (TSs) to the Technical Specification Requirements Manual (TRM), dated December 17, 2002 (TAC NOS. MB3588 and MB3592)

8. South Texas Project, Units 1 and 2 - Issuance of Amendments Re: Allowed Outage Time for Control Room Envelope and the Fuel Handling Building Ventilation Systems, dated September 26, 2000, (TAC NOS. MA3849 and MA3850).

9. NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants",

USNRC.

10. Vermont Yankee Nuclear Power Station License No. DPR-28 (Docket No. 50.27 1)

Technical Specification Proposed Change No. 262, "Alternative Source Term," dated July 31, 2003 (BVY 03-70)

11. Surry Units 1 and 2 - Issuance of Amendments Re: Alternative Source Term (TAC NOS. MA 8649 and MA 8650), dated March 8, 2002.

12. Salem Nuclear Generating Station, Units 1 and 2, Issuance of Amendments Re:

Request of Relaxation of Technical Specification Requirements Applicable During Movement of Irradiated Fuel (TAC NOS. MA 5710 and MA 5711), dated September 16, 2004.

13. NUREG-0737, "Clarification of TMI Action Plan Requirements", November 1980, USNRC.

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

15. Bander, T.J., PAVAN, An Atmospheric DispersionProgramfor EvaluatingDesign Basis Accidental Releases of Radioactive Materialsfrom NuclearPower Stations, NUREG/CR-2858, PNL-4413, Pacific Northwest National Laboratory, Richland, WA, 1982.

16. NRC Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," Revision 1, USNRC, November 1982.

17. NRC Safety Guide 23, "Onsite Meteorological Programs," USNRC, February 17, 1972.

NOC-AE-07002127 Attachment 1 Page 218 of 219

18. NRC NUREG - 6331, "Atmospheric Dispersion Relative Concentrations in Building Wakes", Revision 1, May 1997, ARCON 96, RSICC Computer Code Collection No.

CCC-664.

19. Humphreys, S.L., et. al., RADTRAD, "A Simplified Model for Radionuclide Transport and Removal and Dose Estimation", NUREG/CR-6604 Including Supplements 1 and 2 (RADTRAD version 3.03, USNRC, October 2002).

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

EPA 520/1-88-020, Environmental Protection Agency, Washington, D.C., 1988.

21. "Federal Guidance Report No. 12, External Exposure to Radionuclides in Air, Water, and Soil," EPA 420-r-93-081, Environmental Protection Agency, Washington, D.C.,

1993.

22. Murphy, K.G., and Campe, K.M., "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Design Criteria 19", Proceedings of the 13th AEC Air Cleaning Conference, CONF-740807, U.S. Atomic Energy Commission, Washington, D.C., 1974.

23. American Society for Testing and Materials, "Standard Test Method for Determining Air Change in a Single Zone by Means of a Tracer Gas Dilution", ASTM E741-00, 2000

24. Letter from T. J. Jordan, STPNOC, to the NRC Document Control Desk, dated August 5, 2004 (NOC-AE-04001758)

25. International Commission on Radiation Protection (ICRP), "Limits for Intakes of Radionuclides by Workers," ICRP Publication 30, Annals of the ICRP Volume 2, 1979.

26. Nuclides and Isotopes, 1 4 th Edition, GE Nuclear Energy, 1989.

27. Westinghouse SIP Volume 3-1, RadiationAnalysis Design Manual, Rev. 4, August 1992.

28. Bell, M.J., "ORIGEN - The ORNL Isotope Generation and Depletion Code," Oak Ridge National Laboratory, May 1973.

29. Westinghouse SIP Volume 3-1, RadiationAnalysis DesignManual, Rev. 5, May 1997.

30. ORNL RSICC CCC-371, ORIGEN2, V2.1, Isotope Generation and Depletion Code -

Matrix Exponential Method, August 1991.

31. Liu, Y. S., et al., "ANC: A Westinghouse Advanced Nodal Code", WCAP-10965-A (Proprietary) and WCAP-10966-A (Non -Proprietary), September 1986.

32. NRC Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance With 10 CFR Part 50, Appendix I," Revision 1, October 1997.

33. International Commission on Radiation Protection (ICRP), "Report to Committee II on Permissible Dose for Internal Radiation," International Commission on Radiation Protection (ICRP), Publication 2, 1959.

34. STARDOSE Model report, Polestar Applied Technology, Inc., January 31, 1997.

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

36. STARpH, "A Code for Evaluating Containment Water Pool pH during Accidents,"

Version 1.04, February 2000 and Version 1.05E, December 2005, Polestar Applied Technology, Inc.

NOC-AE-07002127 Attachment 1 Page 219 of 219

37. Wren, J.C., et al, "The Interaction of Iodine with Organic Material in Containment,"

CSNI Workshop on the Chemistry of Iodine in Reactor Safety, Wiirenlingen, Switzerland, June 10-12, 1996.

38. Dean, J.A., "Lange's Handbook of Chemistry," 14 th Edition, McGraw-Hill, 1992.

39. Weber, C.F., et al, "Models of Iodine Behavior in Reactor Containments," ORNL/TM-12202, October 1992.

40. "MicroShield," Version 5, Grove Engineering Inc, Rockville, Maryland, 1996.

41. NRC Regulatory Guide 1.4, "Assumptions Used For Evaluating The Potential Radiological Consequences Of A Loss Of Coolant Accident For Pressurized Water Reactors," Revision 2, USNRC, June 1974.

42. NUREG-0800, Standard Review Plan 15.4.8, "Radiological Consequences of a Control Rod Ejection Accident (PWR), Appendix A," USNRC, Draft Rev 2, 1996.

43. NRC Regulatory Issue Summary 2006-04, "Experience with Implementation of Alternative Source Terms," dated March 7, 2006.

NOC-AE-07002 127 Attachment 2 Markup of Technical Specification pages Page 1-2 Page 3/4 3-20 Page 3/4 3-25 Page 3/4 3-26 Page 3/4 3-27 Page 3/4 3-28 Page 3/4 3-34 Page 3/4 3-35 Page 3/4 3-44 Page 3/4 3-48 Page 3/4 3-49 Page 3/4 7-16 Page 3/4 7/17 (no changes)

Page 3/4 7-18 Page 3/4 7-19 Page 3/4 7-20 Page 3/4 8-9 Page 3/4 8-9a Page 3/4 8-13 Page 3/4 8-16 Page 3/4 9-4 Page 3/4 9-10 Page 3/4 9-14 Page 3/4 9-15 Page 3/4 9-16

NOC-AE-07002127 DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a. All penetrations required to be closed during accident conditions are either:

1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or

2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Specification 3.6.3.

b. All equipment hatches are closed and sealed,

c. Each air lock is in compliance with the requirements of Specification 3.6.1.3,

d. The containment leakage rates are within the limits of Specification 3.6.1 2, and

e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or O-rings) is OPERABLE.

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATIONS 1.9 CORE ALTERATIONS shall be the movement of any fuel, sources, or reactivity control components

[excluding rod cluster control assemblies (RCCAs) locked out in the integrated head package] within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS REPORT 1.9a The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.6. Plant operation within these core operating limits is addressed within the individual Specifications.

DIGITAL CHANNEL OPERATIONAL TEST 1-10 A DIGITAL CHANNEL OPERATIONAL TEST shall consist of injecting simulated process data where available or exercising the digital computer hardware using data base manipulation to verify OPERABILITY of alarm, interlock, and/or trip functions.

DOSE EQUIVALENT 1-131 1.11 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microCurie/gram) which alone would produce the same Comitteid EffectiveDosEie 'EgiRfen, dose as the quantity and isotopic mixture of 1-131. 1-132,1-133, 1-134, and 1-135 actually present. The Cormitted Effetive Dos*ýEgiialen dose conversion factors used for this calculation shall be those listed in Federal GuidanceReport 1.lmt VIalues of Radi~onucfide Int a-keand Air Co-ncentration and Dos EConversion Fac tors for,Inhalation4

,Submersion and: Ingestion,7'1988;1 (Table 2A,. Expos ure-to- Dose Conve rsio'n Fa-ctors for In'hallationl Td-hip F=-7 of~ NIPC 2 {~Q ~ ~ 17 SOUTH TEXAS - UNIT2 1 & 2 1-2 Unit 1 - Amendment No: A122-10 Unit 2- Amendment No. 14llion

NOC-AE-07002127 TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 0 MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE M FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION x

3. Containment Isolation (Continued) z

b. Containment Ventilation C,)

Isolation

1) Automatic Actuation Logic 2 1 2 1,2,3,4 18

2) Actuation Relays*** 3 2 3 1,2,3,4 18

3) Safety Injection *** See Item 1. above for all Safety Injection initiating functions and requirements.

4) RCB Purge 2 1 2 1,2,3, 4, j 18 90 Radioactivity- High

5) Containment Spray- See Item 2. above for Containment Spray manual initiating functions and requirements.

Manual Initiation

6) Phase "A" Isolation- See Item 3.a. above for Phase "A" Isolation manual initiating functions and Manual Isolation requirements.

c. Phase "B" Isolation

1) Automatic Actuation Logic 2 1 2 1,2,3,4 14

2) Actuation Relays 3 2 3 1,2,3,4 14

3) Containment Pressure -- 4 2 3 1,2,3 17 High-3

4) Containment Spray-- See Item 2. above for Containment Spray manual initiating functions and requirements.

Manual Initiation C d. RCP Seal Injection Isolation

1) Automatic Actuation Logic 1 1 1 1,2,3,4 16

=3 and Actuation Relays CD 0-CD z

P

NOC-AE-07002127 TABLE 3.3-3 (Continued)

C,,

0 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION C:

MINIMUM m TOTAL NO. CHANNELS CHANNELS APPLICABLE x FUNCTIONAL UNIT TO TRIP OPERABLE MODES ACTION OF CHANNELS Cn

10. Control Room Ventilation z

C') a. Manual Initiation 3 (1/train) 2 (1/train) 3 (1/train) ýWY2 3 4 27 903 b. Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements.

c. Automatic Actuation Logic 3 2 3 1 22374 27 and Actuation Relays

d. Control Room Intake Air 2 1 2 28 Radioactivity - High

e. Loss of Power See Item 8. above for all Loss of Power initiating functions and requirements.

3 CD CD

NOC-AE-07002127 TABLE 3.3-3 (Continued)

TABLE NOTATIONS

• • • Function is actuated by either actuation train A or actuation train B. Actuation train C is not used for this function.

• • • • Automatic switchover to containment sump is accomplished for each train using the corresponding RWST level transmitter.

1. Trip function may be blocked in this MODE below the P-1 1 (Pressurizer Pressure Interlock) Setpoint.

1. 1. 1. Trip function automatically blocked above P-1 1 and may be blocked below P-1 1 when Low Compensated Steamline Pressure Protection is not blocked.

ACTION STATEMENTS ACTION 14 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />; however, one channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE.

ACTION 15 - (Not Used)

ACTION 16 - With the Charging Header Pressure channel inoperable:

a) Place the Charging Header Pressure channel in the tripped condition within one hour and b) Restore the Charging Header Pressure channel to operable status within 7 days or be in at least Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

ACTION 17 - With the number of OPERABLE channels one less than the Total Number of Channels, place the inoperable channel in the bypassed condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. One additional channel may be bypassed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for surveillance testing per Specification 4.3.2.1.

ACTION 18 - a) With less than the Minimum Channels OPERABLE requirement for Automatic Actuation Logic or Actuation Relays, operation may continue provided the containment purge supply and exhaust valves are maintained closed.

b) MODE 1, 2, 3,U4*

1. With one less than the Minimum Channels OPERABLE requirement for RCB Purge Radioactivity-High, within 30 days restore the inoperable channel or maintain the containment purge supply and exhaust valves closed.

NOTE:

MODE 1, 2, 3, or 4: Supplementary containment purge supply and isolation valves may be open during the allowed outage time for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at a time for required purge operation provided the valves are under administrative control.

__Ives,4ay be-open dur~ing the allowed outage timnefor upt.hirat at S ITH X - 2 _ _ _ _ _ _ _ _ _ _

SOUTH TEXAS - UNITS 1 & 2 3/4 3-26 Unit 1 - Amendment No. 4-,4ý-1-3$Vd Unit 2 - Amendment No.,2-41

NOC-AE-07002127 TABLE 3.3-3 (Continued)

ACTION STATEMENTS (Continued)

2. With two less than the Minimum Channels OPERABLE requirement for RCB Purge Radioactivity-High, operation may continue provided the containment purge supply and exhaust valves are maintained closed.

NOTE:

I- d-- e-_h - Ina~ emy~e4lu4i6 leý r s a4-avimen-fogig oprU IUrdmran aw-ioe o diitai~~irl V ACTION 19: With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

ACTION 20: With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. For Functional Units with installed bypass test capability, the inoperable channel may be placed in bypass, and must be placed in the tripped condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Note: A channel may be bypassed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for surveillance testing per Specification 4.3.2.1, provided no more than one channel is in bypass at any time.

b. For Functional Units with no installed bypass test capability,

1. The inoperable channel is placed in the tripped condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and

2. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for surveillance testing of other channels per Specification 4.3.2.1.

b of ACPERNBLE ~hann'ol&

With thur~q rpm:is thTta punoro lan, ACTION 21: With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> determine by observation of the associated permissive annunciator window(s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

ACTION 22: With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE.

ACTION 23 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SOUTH TEXAS - UNITS 1 & 2 3/4 3-27 Unit 1 - Amendment No. '4-)j 1-gO.ia4*

Unit 2 -Amendment No. °-5§

NOC-AE-07002127 TABLE 3.3-3 (Continued)

ACTION STATEMENTS (Continued)

ACTION 24 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or declare the associated valve inoperable and take the ACTION required by Specification 3.7.1.5.

ACTION 25 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE.

ACTION 26 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, declare the affected Auxiliary Feedwater Pump inoperable and take ACTION required by Specification 3.7.1.2.

ACTION 27 - For an inoperable channel, declare its associated ventilation train inoperable and apply the actions of Specification 3.7.7.

ACTION 28 - a. With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, within 7 days initiate and maintain operation of the Control Room Makeup and Cleanup Filtration System (at 100% capacity) in the recirculation and makeup filtration mode.

b. With the number of OPERABLE channels two less than the Minimum Channels OPERABLE requirement, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> initiate and maintain operation of the Control Room Makeup and Cleanup Filtration System (at 100% capacity) in the recirculation and makeup filtration mode7bA

~J-~:*--I-----~ -------

FFf-laveoprain it oc poANwthn, 12 h o u r initiat6-and imaantain operation of the Gontro, Room Makeup and Cle.anup Filtration .Syete- Lit 100O%oapaoity)in the ciricudation and makeup~ fitration- mde.-,R- LTERATIOWS nioenn-firamtdfe asemliesan c r an-ope4r tionsytith lnh;roa ~e rit~~herspant fu~elpool are p~errmitted d-rion aiono thVe- Conier aker and G4I~nu~p

~ inthorebrcuatin I~!!fitationn dG

c. With required ACTION 28a. or 28b. not met in MODE 1,2,3, or 4, neda L-M =,  ;ý be in MODE 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 in the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

K~ithequw r~d-ACT-IOhJ 2Ft. or-28b. not me-t inAN MODE :5-or 6&,-4imei~ately uspend CORE-Al TERA14TI US,mvernen t 1it.I wit loadk~

ýFor-apinoperable channel -decfare-ý_ t then actoans t~ atiop-3-. 8-PPTIONqý-With fuel E-Ghannefiý_"_

than the M~inimum Channels -OPERiABLE requirement, fuel movement within the spent fue-po~r- n-o e wýithloads-ovrho ct~fuel ~pool ma';pfoceaed provided s*se-s innirain !A-#§.gt t ~rf i iyi sroprto n ihrjhg~te SOUTH TEXAS - UNIT 1 & 2 3/4 3-28 Unit 1 - Amendment No. _

Unit 2 - Amendment No.;,,

C/) NOC-AE-07002127 0

C: TABLE 3.3-4 (Continued)

--i ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS C,)

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE C

z 9. Engineered Safety Features Actuation System Interlocks

a. Pressurizer Pressure, P-i 1 <1985 psig <__1995 psig cU)

b. Low-Low TAVG, P-1 2 > 563°F > 560.7°F

c. Reactor Trip, P-4 N.A. N.A.

10. Control Room Ventilation CA)

a. Manual Initiation N.A. N.A.

b. Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values

c. Automatic Actuation Logic N.A. N.A.

C: C and Actuation Relays

d. Control Room Intake Air

<6.1x10"5 _<7.8x10-5

> > Radioactivity - High pCi/cc pCi/cc 3 3 CL e. Loss of Power See Item 8. above for all Loss of Power Trip Setpoints 3 CL 3

and Allowable Values (D

Z =':F':1 1Ti* qOA zzoZ "

NOC-AE-07002127 TABLE 3.3-4 (Continued)

C,,

0 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS r

--I FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE m

z S~. ... c ~A~>%*,t ..hk..~...

C',

90 d~SphT~uj~~IE~haet <~ 5.xi 0K276- 1 C33 CDCD 3 3 CD (D zz

NOC-AE-07002127 TABLE 4.3-2 (Continued)

Cn 0 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS m DIGITAL OR TRIP ANALOG ACTUATING MODES C/) CHANNEL DEVICE MASTER SLAVE FOR WHICH C: CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL AC.TUATION RELAY RELAY SURVEILLANCE z FUNCTIONAL UNIT CHECK CALIBRATION TEST (7) TEST LC)GIC TEST TEST TEST IS REQUIRED

3. Containment Isolation (Continued)

3) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

1 ,2, 3, 4, 9;:

4) RCB Purge S R Q N.A. N. A. N.A. N.A.

C,,

Radioactivity-H igh

5) Containment Spray See Item 2. above for Containment Spray manual initiation Surveillan ce Requirements.

- Manual Initiation

6) Phase "A" Isolation- See Item 3. a. above for Phase "A" Isolation manual initiation Surveill.ance Requirements.

Manual Initiation

c. Phase "B" Isolation

1) Automatic Actuation N.A. N.A. N.A. N.A. Q( 1) N.A. N.A. 1,2,3,4 Logic

2) Actuation Relays N.A. N.A. N.A. N.A. N. A. Q(6) Q(8) 1,2,3,4

3) Containment S R Q N.A. N. A. N.A. N.A. 1,2,3 C:C:

Pressure--High-3

4) Containment See Item 2. above for Containment Spray manual initiation Surveillan ce Requirements.

Spray- Manual Initiation

d. RCP Seal Injection CD CD Isolation

1) Automatic N.A. N.A. N.A. N.A. N. A. Q Q(8) 1,2,3,4 z z Actuation Logic and Actuation Relays

2) Charging Header S R Q N.A. N. A. N.A. N.A. 1,2,3,4 Pressure - Low Coincident with See Item 3.a. above for Phase "A" surveillance requirements.

Phase "A" Isolation

NOC-AE-07002127 (I,

0 TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION m SURVEILLANCE REQUIREMENTS DIGITAL OR TRIP ANALOG ACTUATING MODES z CHANNEL CHANNEL DEVICE ACTUATION MASTER SLAVE FOR WHICH C', CHANNEL CHANNEL CALIBRATIO OPERATIONAL OPERATIONAL LOGIC RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK N TEST (7) TEST TEST TEST TEST IS REQUIRED r%) 10. Control Room Ventilation (Continued)

b. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

c. Automatic Actuation N.A. N.A. N.A. N.A. Q(6) N.A. N.A.

Logic and Actuation Relays

d. Control Room Intake N.A. _1 -3 S R Q N.A. N.A. N.A.

Air Radioactivity-High

e. Loss of Power See Item 8. above for all Loss of Power Surveillance Requirements.

C C

3 3

=3 :3 CD (D 3 3 ii a

z

=3 z

NOC-AE-07002127 (I) 0 TABLE 4.3-2 (Continued) r

--I ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION m SURVEILLANCE REQUIREMENTS x

Cl)

C DIGITAL OR TRIP z ANALOG ACTUATING MODES Cl) CHANNEL DEVICE ACTUATION MASTER SLAVE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL LOGIC RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST (7) TEST TEST TEST TEST IS REQUIRED TABLE NOTATION (1) Each train shall be tested at least every 92 days on a STAGGERED TEST BASIS.

(2) Deleted (3) Deleted (4) Deleted C C (5) Deleted (6) Each actuation train shall be tested at least every 92 days on a STAGGERED TEST BASIS. Testing of each actuation train shall include r3 3 master relay testing of both logic trains. If an ESFAS instrumentation channel is inoperable due to failure of the Actuation Logic Test and/or CD C Master Relay Test, increase the surveillance frequency such that each train is tested at least every 62 days on a STAGGERED TEST BASIS Q-0 unless the failure can be determined by performance of an engineering evaluation to be a single random failure.

3 3 (7) For channels with bypass test instrumentation, input relays are tested on an 18-month (R) frequency.

z z 0 0 (8) The test interval is R for Potter & Brumfield MDR Series slave relays.

,n ALT at. .Ar .hnG~amil gr~-> f~ia z<tLP

NOC-AE-07002127 PLANT SYSTEMS 3/4.7.7 CONTROL ROOM MAKEUP AND CLEANUP FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.7 Three independent Control Room Makeup and Cleanup Filtration Systems shall be OPERABLE.

APPLICABILITY: R MODES 23 and-4 ACTION:

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

b. With two Control Room Makeup and Cleanup Filtration Systems inoperable, restore at least two systems to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. With three Control Room Makeup and Cleanup Filtration Systems inoperable, 6

_____*.L__.pQQ ... , restore at least one system to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

F- .:y7:;**i696 '*[:,A* ":*

of !h9 rmining (,1,E .RALECon-Atrol- Room Makeup and GoanUP ,i~tration 9Sycomc 41n1cos-frdquire-d-SHLTD0WN MARGIN or requirod boronp ccnrton, movementP spont typ,_Fjp prpnn Ganti atnnon lAflth loadsdtever

,thep ontf SURVEILLANCE REQUIREMENTS 4.7.7 Each Control Room Makeup and Cleanup Filtration System shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the control room air temperature is less than or equal to 780 F;

b. At least once per 92 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers of the makeup and cleanup air filter units and verifying that the system operates for at least 10 continuous hours with the makeup filter unit heaters operating; SOUTH TEXAS - UNITS 1 & 2 3/4 7-16 Unit 1 - Amendment No. ,128 Unit 2 - Amendment No.

I No Changes this Page NOC-AF--07002127 I PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

c. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system by:

1) Verifying that the makeup and cleanup systems satisfy the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% for HEPA filter banks and 0.10% for charcoal adsorber banks and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 6000 cfm + 10% for the cleanup units and 1000 cfm + 10% for the makeup units;

2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of ASTM D3803-1989, "Standard Test Method for Nuclear-Grade Activated Carbon," for a methyl iodide penetration of less than 1.0% when tested at a temperature of 300C and a relative humidity of 70%; and

3) Verifying a system flow rate of 6000 cfm 10% for the cleanup units and 1000 cfm

+ 10% for the makeup units during system operation when tested in accordance with ANSI N510-1980.

d. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of ASTM D3803-1989 for a methyl iodide penetration of less than 1.0% when tested at a temperature of 300C and a relative humidity of 70%.

e. At least once per 18 months by:

1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6.1 inches Water Gauge for the makeup units and 6.0 inches Water Gauge for the cleanup units while operating the system at a flow rate of 6000 cfrn + 10% for the cleanup units and 1000 cfm & 10% for the makeup units.

2) Verifying that on a control room emergency ventilation test signal (High Radiation and/or Safety Injection test signal), the system automatically switches into a recirculation and makeup air filtration mode of operation with flow through the HEPA filters and charcoal adsorber banks of the cleanup and makeup units; SOUTH TEXAS - UNITS 1 & 2 3/4 7-17 Unit 1 -Amendment No. 127 Unit 2 - Amendment No. 116

NOC-AE-07002127 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

3) Verifying that the system maintains the control room envelope at a positive pressure of greater than or equal to 1/8 inch Water Gauge at less than or equal to a pressurization flow of 2000 cfm relative to adjacent areas during system operation j; and I

4) Verifying that the makeup filter unit heaters dissipate 4.5 + 0.45 kW when tested in accordance with ANSI N510-1980.

f. After each complete or partial replacement of a HEPA filter bank, by verifying that the HEPA filter bank satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a DOP test aerosol while operating the system at a flow rate of 6000 cfm + 10% for the cleanup units and 1000 cfm + 10% for the makeup units; and

g. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the charcoal adsorber bank satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.10% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 6000 cfm + 10% for the cleanup units and 1000 cfm + 10% for the makeup units.

SOUTH TEXAS - UNITS 1 & 2 3/4 7-18 Unit 1 - Amendment No.

Unit 2 - Amendment No.

NOC-AE-07002127 PLANT SYSTEMS M4.7.6- PUEL R4NDLING BUILUIWG (hhbatYFMAUý;i A... i4EM

. ThliS speciication is not used)

IMI~ITINGv G0C9M0NITF'QM CI ('DATfr)hl P

-7 L ~ ~ -.

C) 13, PFIF) ayS t GM GMPrjSt9H G+ TnG be OPER.", LE.

-  ;,- -j I iew. tuL c!.

b. Three exhai sit ventflp ton:trains-af'k APPLGABILTY MOES1

___ __ __ __ r ___

.e r __ __

egiaie __ esoP. F4r

'OndRABn SHUTDOW7 cinyt~oreOwng 30es HOTurs.Y~hd~hG4tS-8tOLD fl,~ ~~~~~~~~~~~~~~~~~ gtartltrt-n npaL.-a4nn tlllrlfl ntrn, ýp.t n nn- in^ enrarI k t.Itrlr traint OPEABE taus-wthin 1`2 ho~urs or h4 in At 104st HOTjSTANDBY ulithimnthe noxt 6 hourws

%AAA, on -~xast vontilation 4jraiP'inoperab~q,,,rertoro 4h6 inQpý6rabo oxau ventilation train to OPERABLE statwrs with in 7,days or be i~ es O TNB ti hofe hour- --d in~COLD S-HUTDOWN injth~o fol1lowng430 hours.

1. With more thaR one-FB exhaust Venti1a-tiQn Ir~i inpral) Id6tF9att toexha ust ontilatio ~ OEAL s.*tatus 'Mithin 1:a "our Or be in at least HOLT STADB Within the next 6 haiurs and in COLD SHUTDO)(WN~ in the follAown ý30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> S&uR-VEILLANQCE REQUIREMENTR 14.7-.8 -The Fue! Marfdling Building ExhausYAirSs, r halb dmnsrtd PRAL:

- At Iat; onco 'por 31 days on a STAGGERED TEST BSSby initiating, from the c~trol I.has LA*--

hogih E~flesand charcoal ads~orborsa4VGiyig

?omlw thand-9 opwarte& .fr- ý1 Gfl6s c&tin6uG h(Aurs With the hea r rI,,yitgha twoofthe~ thr betrands~ ,n Of the thhree main exas fn prating to m~aintain adoguIate air flow rate-  ;

At 1-+oas i49 18oo mo8

nbth~anld'(!)atrn srcirlm t~c thi. HERPA filtoK OFchaF~eal adsEdrbor hosn~or (2),folloyving paintinq, ire, or chempical releasein aR~

[,o tilatio GE) Gm~nr ý tho system by:.I

4) --, the cleanup system satisfigs the in place, penetration and bypeass

-eiyigtat, 140kg ' " testig

...... aeceptancoc6ritoria

"  %' " , of loss < thapn . 0.05% .df~r HEPA jilterban.san

• n '*.e"!R*

crcaca dobrbnsaduo hts r~d~~~ d h iOt%

r'i rdan~

S. -1 70 .an, +ke Slfc+rM floW ratE 1is29,000 cfM r!19%j VeFjfxi-,-:a VVjthjR Ql An o nftýr r omoval, that a labortatoý analysis ofa f 4pTleseentatve TcII uI samII p~U.l- J 44U (A inJx11,L IlI~l~IL~~IJIl SOUTH TEXAS - UNITS 1 & 2 3/4 7-19 Unit 1 - Amendment No.

(Next page is Unit 2 - Amendment No. 4ii:"1-0 3/4 7/21

NOC-AE-07002127 PLANT SY&T-EF-S W4, 7. 8 FUEL HANDLIN~GI BIDIi (F=HB) EXH~LAUSTQ- AIR SY-SýA Reision 2, March 1978, meets the laborator taSting c~rteria of AkSTMi Fo26i) 1989 Ilenotr at! Gof less;;q thanR 1.00,4a he Ates atatemperatd~ nf o 51nD -a-d a relativ reovn~al, that a Iabqrat9ry ahalysis of a Fapr~esentative carbGR sample obtained 4n# R 2G~xnFu:Rireu VVItri ReguuuItiery Wr'rITIiuR1 k-n. of Rectxulatery 6..' - hI-nr FqerI'.I[:Irýlrl r(_44 eets t he abor~atoerj teotting Griteria of,/STM D3903 10989 fer a methyl odd iii8,

~ontraionofles thn ..... thest~ed at a temperature of 30 0C and a relative I d At loasIt onncer-pr 18months b"ý 1,)V~fyig that theo pres- we dro acroSs the combined HEPA filtors and charcoala 4 hdcorbRbnkG is IeGG than 6 inc~hos Water Gauge while eperating tlho system at a L24 kllrifying that thocGystem strt op9High Radiation; and Safot' Injection test signals

~nd~irotc lowthFGUgh tho- HEEPA filters and charcoal adsorbors, -

X/CýF~fthat

-1 _Y .

j . . .t .

(Z.. .

Y .lit u- n5Z8It*I E 1 I af IWP nn o rfivi rnJt.ýf cir f o tr fhrn ý t

o.

'.AF,6ao -bpmlet66otprial; rplac emog ot AHERA filter bank, b+, verifying that they, HEAfilter bank satisfies theý iq :piacc plcntratiG Rand: bypass leakage testing aGcdptan Ge

,,ýperafing the Gystm atafof 1% al:d~

I+ým I f ,-

+f I,

ý pI 14 r v rt; t *rý ru 1

ctF

+

- -. ~h ,r~~i u c cc-*

A 6"r*

V , k ýn 1+~

Hr'-***

.4~k

, y-, *ý=r* rig chaFrcoa -dGsrbgr bank -satissfies- the ;n place penetration and bypa~ss akgtetn 0 9000cfm;9- r ., A>f~1Q. tha -. 50490fra .lI10'aer-RewthA ha~Gae hYIGabRFfi~atts a hl p~t~ h ytm_: loýrt9 SOUTH TEXAS - UNITS 1 & 2 3/4 7-20 Unit 1 - Amendment No.

Unit 2- Amendment No. i 14

NOC-AE-07002127 ELECTRICAL POWER SYSTEMS A.C. SOURCES SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. One circuit between the offsite transmission network and the Onsite Class 1 E Distribution System, and

b. Two1 standby diesel generators each with a separate fuel tank containing a minimum volume of 60,500 gallons of fuel.

APPLICABILITY: MODE 5 and MODE 6 with water level in the refueling cavity < 23 ft above the reactor pressure vessel flange.

ACTION:

With less than the above minimum required A.C. electrical power sources OPERABLE, immediately suspend all operations involving CORE ALTERATIONS, operations involving positive reactivity additions that could result in loss of required SHUTDOWN MARGIN or required boron concentration, .,,V&.... .... ....... f ,ul operations with a potential for draining the reactor vessel - Fa~e Immediately initiate actions to restore the inoperable A.C. electrical power source to OPERABLE status.

SURVEILLANCE REOUIREMENTS 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the requirements of Specifications 4.8.1.1, 4.8.1.1.2 (except for Specification 4.8.1.1 .2a.3), and 4.8.1.1.3.

4.8.1.2.1 The alternate onsite emergency power source shall be demonstrated functional by:

a. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of taking credit for the onsite emergency power source as a standby diesel generator, verify it starts and achieves steady state voltage (+/-10%) and frequency L+2%) in 5 minutes.

b. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of taking credit for the onsite emergency power source as a standby diesel generator and every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter, verify the emergency power source is capable of being aligned to the required ESF bus by performing a breaker alignment check.

'An alternate onsite emergency power source, capable of supplying power for one train of shutdown cooling may be substituted for one of the required diesels for 14 consecutive days (SR 4.8.1.2.1 is the only requirement applicable).

SOUTH TEXAS - UNITS 1 & 2 3/4 8-9 Unit 1 -Amendment',-

Unit 2 - Amendment

NOC-AE-07002127 ELECTRICAL POVER SYSTEMS A.C. SOURCES SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.3 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. One circuit between the offsite transmission network and the Onsite Class 1 E Distribution System, and

b. One standby diesel generator with a separate fuel tank containing a minimum volume of 60,500 gallons of fuel.

APPLICABILITY: MODE 6 with water level in the refueling cavity > 23 ft above the reactor pressure vessel flange.

ACTION:

With less than the above minimum required A.C. electrical power sources OPERABLE, immediately suspend all operations involving CORE ALTERATIONS, operations involving positive reactivity additions that could result in loss of required SHUTDOWN MARGIN or required boron concentration, f f-' f4i*, operations with a potential for draining the reactor vessel or.crane oper t*q j .witb

-, Immediately initiate actions to restore the inoperable A.C. electrical power source to OPERABLE status.

SURVEILLANCE REOUIREMENTS 4.8.1.3 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the requirements of Specifications 4.8.1.1.1, 4.8.1.1.2 (except for Specification 4.8.1.1.2a.3), and 4.8.1.1.3.

SOUTH TEXAS - UNITS 1 & 2 3/4 8-9a Unit 1 - Amendment Unit 2 - Amendment 0-7

NOC-AE-07002127 ELECTRICAL POWER SYSTEMS D.C. SOURCES SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 DC electrical power subsystem shall be OPERABLE to support the DC electrical power distribution subsystem(s) required by LCO 3.8.3.2, "Onsite Power Distribution -

Shutdown."

APPLICABILITY: MODES 5 and 6.

ACTION:

With one or more required DC electrical power subsystems inoperable, immediately declare affected required feature(s) inoperable OR immediately initiate action to suspend operations with a potential for draining the reactor vessel, suspend all operations involving CORE ALTERATIONS, r operations involving positive reactivity additions that could result in loss of required SHUTDOWN MARGIN or required boron concentration" ' ...... nt ofi....o

• j; initiate corrective action to restore the required DC electrical power subsystems to OPERABLE status as soon as possible.

SURVEILLANCE REOUIREMENT 4.8.2.2 The required DC sources shall be demonstrated OPERABLE in accordance with Specification 4.8.2.1.

SOUTH TEXAS - UNITS 1 & 2 3/4 8-13 Unit 1 - Amendment 1 Unit 2 - Amendment

NOC-AE-07002127 ELECTRICAL POWER SYSTEMS ONSITE POWER DISTRIBUTION SHUTDOWN LIMITING CONDITION FQR OPERATION 3.8.3.2 The necessary portion of AC, DC, and AC vital bus electrical power distribution subsystems shall be OPERABLE to support equipment required to be OPERABLE.

APPLICABILITY: MODES 5 and 6.

ACTION:

With one or more required AC, DC, or AC vital bus electrical power distribution subsystems inoperable, immediately declare associated supported required feature(s) inoperable OR immediately initiate action to suspend operations with a potential for draining the reactor vessel, suspend all operations involving CORE ALTERATIONS, operations involving positive reactivity additions that could result in loss of required SHUTDOWN MARGIN or required boron concentration, ,,v..cnt of ir,....t*6d and immediately initiate corrective action to restore required AC, DC, and AC vital bus electrical power distribution subsystems to OPERABLE status and declare associated required residual heat removal subsystem(s) inoperable and not in operation.

SURVEILLANCE REQUIREMENT 4.8.3.2 Verify correct breaker alignment and voltage to required AC, DC, and AC vital bus electrical power distribution subsystems at least once per 7 days.

SOUTH TEXAS - UNITS 1 & 2 3/4 8-16 Unit 1 -Amendment Unit 2 -Amendment 117

REFUELING OPERATIONS REFUE IONSNOC-AE-070021 ING PERA NOC-AE-07002127 27 3.9.4 The conainimentbuilding p netre ations shall be in thG fo~llevvi g~tatus:. H

a. Th oaoquo rn ont c and aa Aco~ h ol1d in.rla c- a ,minimum G o ,ffour b ol ts 1.>The Roactor has been su 7iia -eý 16 hur,6\!

A'ppni tfhe 9guipnionfhatc!1;iS capablo of being clesod,

b. 1) Aminimum ofon dear in teýOconfihMbnt Auxiliary Ailck (AAL i lon-edi7-AN D . i .~ L X r A

-. M ~4~) !W~W IDA el tAFl-Gp d--)Is 4lse 4.

The.;ter * .23 level s 'feet a. ioVo Me reactor vessel nango.

ý - 11 d- -1 hip ~ ~~ ~iia ~ o ~ 5h~ ~ cbt'~

h~iiaio ~5h (fe u theto ofeau~

ue1lhandling acGcidont insido continMent) -WithIn

.a. 30, minutes, if the reactoer h-as been- Subcr~itical <165. hours

~4'4l~

4; SrL. SGG. aMui u~ m x xPOSiiri ~E*"I u~ x VVRHI. f- HO U ff t in :[ea t r e iqasg.. noun 4 riut 1 Al 4n aGcess fromn the bentainmc )nt atmosf:phqere t the

~outsido atmesphero ehllbt-ethe;-:

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4: 4444' fV~~ A I -~ .W.-&~A $.

Q I IE M 71I I Ars Mrý ~ ID A N T 4:9 A'>.Each of- Pda b ev6 requiredc rontainm ont uaildingp cnctratkbh  !'qha;lb d e e m n d blih ri its'required 4ohdifieno r,'capableo Qfbeingt losed srequired i n p Iificatie 3 .9.1w ithin 100 h ours pI r4t heci tA thos on a Ofnand a t l oa t GRG Sta~ nc crjr7p dlc avsd>urin:'4' 'C O RE A TE RAT A 9 E) r' movomon..P- offrridii attodluolItIiAi 44.

in the 44.

Atinniz pep in -ýhmor' rý-Lcl, hrindifinn SOUTH TEXAS - UNITS 1 & 2 3/4 9-4 Unit 1 - Amendment No. 43 Unit 2 - Amendment No. ý

NOC-AE-07002127 REFUELING OPERATIONS

~t of irradiatod fuol within j M.^_+;_ý 0 r) "ll- -- f -- -

SOUTH TEXAS - UNITS 1 & 2 3/4 9-10 Unit 1 - Amendment 5 Unit 2 - Amendment

NOC-AE-07002127 REFUELING OPERATIONS 3/4.9.12' (-hiSspecificationis not'dsed) FUEL HANLINGBUILDING,EXHAUST- AIR r STEN Iiir TI fD r~Drrk 3.9.12 Tho FHB Exhaust ~ir System 4 comprisod~of tho following compononts shill be~.

I A.:Nve_ oxh-Auvt -Airfilter trains,

• b.Tvtn)exohaustvontilatin tra nG 1AP-PLlCA4B[LlTY'( Whcnevor-irradiated fuel is in the spent fuel pool a' ith 'bcSS th-;r4 thos ;afb~ol FHB Exh au'st Air ýSystem components OPERABLE but with at least ono FHB oxhaust air filteFrrain,oenoR FHB exhaustyontilationR,

~tramR, and asSO~iate~d dampers OPERABLE, fuel mo-mnt wAithinR the6 spent

'fuol~e poo orianc operatiOR With I6ads ove,r I.- spent fuel pool may, prcod Sprovide 1_the-COPERABL-E-FHB, Exhau~stAir ~se on r aal of-be&ingC poDWood froCm an OPERABLE emroc pwer Source and aro in' operation and discharging tehrog t !east oe raie fHP it~ and G~~harcoal ads64erber 14iith no FHR- exhus ai filte'r trin OPEABLE, suspend all~ epertiens~~

', loads evorth Pie t fuel 'pool;ii

' < The pro~visinc)sof Spocificat!on 3.0.3 ~ntailcbo iSURVEILLANCE REQUIREMENTS P. 2 ~The above e4gurrP0 FH ~Ehaust AirSy toms shall be domonstrated

'a_ AfoSrt A! oncoý per 31 (lays onI a.STAGGEýREIDI TES'T' BASISIG by initiating, jroml th-o- )trol rooIm, flow threugh tho HEP4A fbiters~a~ Gh~caIGA Ad~br-andý veflyi_6 that tho .ystor opqo'aos fbr at least. 1 GCitinlious ho)4&withfho, 1heaters-op-orating with the oper-ablo exhaust boposcter fa~s and the-ope-rable

>1 At least e p F=HB oxhaust-air'filter tramn, one FHB, botR 77as, and oe ~llB mnain oxhauft'fan-. arGapable of being powered fromn Ph CPERABLEEdnsite omergency, iiaes 49- 5 anc 3/4'9-16 have beeL-n SOUTH TEXAS - UNITS 1 & 2 3/4 9-14 Unit 1 - Amendment No. 71,Ld Unit 2 -Amendment No. 60.1!

NOC-AE-07002127 rbcf:ifflllffl Mra nm:ohýn

~44-444- -44~4~ ,-~ -~~- - -'K eand-4-"Y 6 Ht -14fera~

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,ve4alR_ zone GcoJmmuR: Gt~_ thteS~tMb-MJx A -

toirutot-o in-pew pu uun iuuyy oPakage toct;ng aG~ptan&&c-Grtria of locc th-ap 0.05Z%for, HERA, flilter-bani- -4 oi Q%4 for. Ghar~oal aadsorbor banks -anpd ucoc tho tot rduroe~daci

(:ý 1 bf P4-1 I G I li Pliamh 1978a sstm fl ow rate is 29,000*~rn -1:1 00'c-

,GratGFy aRalysis b i d ith R l I

4~~4444' n.

4 oTnfc~traTxir Guide 1i.52 , Reivision -- ~ 2., M~arcnh - - i - m19:78nrto'

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, meetp- the I.AbGratery Fade AGfi testiRgdi Griteria C: -.- h o f- FaI F Ar D38 "Staiýdard Test Methodi fn-methyl iedide &M444444~4444atiRc4 44 4 e4 4 4 ha .446 nte4Meat 0G t

- ~-- 4~ " -/:4~Jff~~-

p 4relatv 444dt-f and~

of the th(ý ebAý ahct'.eo oxas f thc th6main exhaust fans' oprt9

ýý.hen tested, in R ccordancoewit 1 T-ý

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- 1 ~-1.S4- N 544 1980 jAf1obiron 4

14 I -4 77 of6 Iw 4Jhu

-ko vry, 720 hew of prber praonb i':it~i 1dy fo

-n - -- - I..44L remGval laberatery aRalyGis ef iia reprersentativ barwpn sampi! eb!'pui aGGord ih R W P ii G. Gf R l i , *

'4-9-7. moots h loaborator,,, testing criteria or A T-M W80U3 -14UU-for a methylidd

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that .the prýsG494GP aC;r()SS ýýe t

Ile F" i*4 r hAnký ir, l6ss thap 6 ipr-hqp Watpý, G;;- ga wkle kgý jeA-ý of 2 9 000 GfPTýAqýý

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• grin,. flrTno-i- -nfl - -, 'i-4 ar fi LJ~ rhi.,tinrV44tnr.t4C. inn ni t1~~r. C-, ,p-inrn p. itnrnatit'., tilt nitfltp t.~i rl Ir4C. p that on a Hi h KaGla;19R ies (69!&r,

,already operatdýgyjýqd ci rp(,-tq its exhý'wpt flpvý -_HEPA filters a qýyqgýj SOUTH TEXAS - UNITS 1 & 2 3/4 9-15 Unit 1- Amendment No.

Unit 2 - Amendment No. 4

NOC-AE-07002127 YI'~ tha AnfPFMD'mant theI~ Syt th pn ulECFg GIae ta!

eý1lt / ag a i _ l~

"quýn ppyLsq r _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

thp rh k-dreebr bank-saf isies the in-Place penetr~afi6n-n and bypass eakagotoctindý 4)t%iRaccodRlGG nc ith'A'Nsl N51 0 19RQOf~

SOUTH TEXAS - UNITS 1 & 2 3/4 9-16 Unit 1 - Amendment No. M

NOC-AE-07002127 Attachment 3 Attachment 3 Technical Specification Bases pages (information only)

Page B 3/4 3-2b Page B 3/4 3-2c Page B 3/4 3-2c. 1 (temp)

Page B 3/4 3-2d Page B 3/4 3-3 Page B 3/4 7-4 Page B 3/4 7-5 Page B 3/4 7-6 Page B 3/4 7-7 Page B 3/4 8-14 (no changes)

Page B 3/4 8-15 Page B 3/4 8-19 Page B 3/4 8-20 (no changes)

Page B 3/4 9-1b Page B 3/4 9-2 Page B 3/4 9-3 Page B 3/4 9-3a

NOC-AE-07002127 Attachment 3 INSTRUMENTATION BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)

When control rods are at the top or above the active fuel region LŽ 2 step 259), they are no longer capable of adding positive reactivity to the core, and as such, they are not capable of rod withdrawal as intended by MODE 5*. Therefore, ACTION 10 on Table 3.3-1 is not applicable in this region. This allows the Reactor Trip Breakers to be closed, without meeting the requirements of MODE 5%, while unlocking and stepping the control rods to a position no lower than 259. (CR 97-908-17)

Several ACTIONS in Tables 3.3-1 and 3.3-3 have been revised to change the allowed outage times and bypass test times in accordance with WCAP-10271 and WCAP-1 4333.

Additionally, some ACTIONS have been divided such that only certain requirements apply depending on whether the Functional Units have been modified with installed bypass test capability.

Regardless of whether the Functional Units have installed bypass test capability, it should be noted that in certain situations, the ACTIONS permit continued operation (for limited periods of time) with less than the minimum number of channels specified in Tables 3.3-1 and 3.3-3. For example, Table 3.3-1 Functional Unit 11 (Pressurizer Pressure - High) requires a minimum of 3 channels operable. However, since continued operation with an inoperable channel is permitted beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, provided the inoperable channel is placed in trip, and since periodic surveillance testing of the other channels must continue to be performed, ACTION 6 permits a channel to be placed in bypass for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to permit testing. Thus, for a limited period of time (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />), 2 channels, or one less than the minimum, would be permitted to be inoperable.

Actuation relays consist of slave relays, including the relay contacts for actuating the ESF equipment. If a slave relay becomes inoperable for a particular component(s), then the associated component(s) LCO Required Action should be entered. If an entire train of slave relays for a functional unit becomes inoperable, then the Required Action for the functional unit actuation train should be entered. (CR 00-1 3604-7)

During a plant shutdown for refueling, the Normal Containment Purge System is in operation. The Supplementary Containment Purge System may be used during normal plant operation. Redundant Class 1 E radiation monitors (i.e., the Reactor Containment Building [RCB]

Purge Isolation) monitor the radiation in these purge lines. Upon either monitor sensing radiation above a preset limit, a signal is sent to the ESFAS logic trains, and the Containment ventilation isolation signal is actuated. In a LOCA, both Normal and Supplementary purge lines are isolated by a Safety Injection (SI) signal. Actuation of the purge isolation by these radiation monitors is not credited in the LOCA accident analyses, and is only a backup function for this event.

SOUTH TEXAS - UNITS 1 & 2 B 3/4 3-2b Unit 1 - Amendment No.

Unit 2 - Amendment No. 01 -16

NOC-AE-07002127 Attachment 3 INSTRUMENTATION BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)

ACTION 18.a. applies when the actuation logic for RCB Purge Radioactivity - High is inoperable because it affects both channels. The required action is to maintain the isolation valves closed. Loss of power supply to the output ESF relays of either channel of these monitors will be considered inoperable actuation logic and the isolation valves will be maintained closed in accordance with proposed ACTION 18.a. This is because this failure mode will result in the inability of the other actuation signals to close the purge valves ifthe initial signal is reset.

In MODE 1, 2, 3, W4, 6 when one of the two required channels of RCB Purge Radioactivity - High is inoperable, ACTION 18.b.1 requires restoration within 30 days. The allowed outage time is a reasonable time for easily accessible non-risk-significant instrumentation. The required action is modified by a note that allows the supplementary purge valves to be opened in MODE 1-4 under administrative control during the 30-day allowed outage time to permit operation of the supplementary purge system for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at a time for the evolutions permitted by the Technical Specifications (containment pressure control, ALARA and respirable air quality needed for personnel entry into containment and for surveillance tests that required the valves to be open). The 2-hour allowance is adequate time for the routine pressure control purge operations during power operation.li not~ e alI n_%vs he , rmIi or-supplemeontary' purge suppi'ndchiale t;u-to6 our a-tme~r-~-MODE a4~i 5ff for equired purge oppratkibn- .The 6hu s ustfid ecuseth dsinasiseve~nt in this MAODS IAUldc be expe-cted to bo~a durlin

ýIwe-4er dvlphingevntad pu e oelrationg iin Quppoto euln ciiicaetp~l muhlnger than thqse done at p Opening the valves for purge operations is not permitted after the 30-day allowed outage time has expired.

In MODE 1 - 4, the safety analysis credits only the SI signal for actuation of CVI. As a backup, the operable radiation monitoring channel would still be available to actuate containment isolation. InMD n 6 rdbp -'_ vn adte s Administrative control during purge evolutions with an inoperable radiation monitoring channel would include the operator ability to manually initiate CVI from the control room handswitch and typically include an assessment of plant conditions for potential actuation precursors, monitoring containment radiation and limiting purge duration.

ACTION 18.b.2 applies in MODE 1,2, 3, when both channels of RCB Purge Radioactivity - High are inoperable. The action requires the purge isolation valves to be maintained closed and there is no provision for purge operation under administrative control.

SOUTH

n. &e 2 6 3/4 3-2cnit nit21irradiated

_-of fuel o .rN° ALTF-R,

~ ~~

TeATOdiet ~ ~ he ~ ~ t-pyh-requiremrents-of-T,%-3t9.-G4

~~t noeal otimn elt~~eainSse during Refueling. With oneIn channel~~~~~

RBPreRdacityHigh nl tnprb he action includes a provisio tha ure oer~on f~up-o--howsat a im,&.Th4 baifrh6-hou 4rdurioR-f toe allws SOUTH TEXAS - UNITS 1 & 2 B 3/4 3-2c: Unit 1 - Amendment No. ~ i Unit 2 - Amendment No. ___Q-__-

NOC-AE-07002127 Attachment 3 voifage or rmore th'an Gne degradedvcoltage Ghannol por buc are inet Th n t~

potiox gcluir 8 roe oring All but one r-hannol por bpc to QPERAIBLE#sttu.Th! j Tame allw aipý tieto rpair n-le~t faihuros an a_

probability of an event Fequiring an LOP sta4 GGcrrng~ during thiS infetoral.' If t~he ichanno~l r correRHCP @n 9ýim 4a-sý after approval'qf,;bt6.licenc amondrnept thant SQU-TH:TEAS UJNIT-S 1 & 2 B 3/4 3c1 eý nf Andhien

, O 481

NOC-AE-07002127 Attachment 3 INSTRUMENTATION BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)

ACTION 27 for an inoperable channel of control room ventilation requires the associated train of control room ventilation to be declared inoperable and the appropriate action take in accordance with Specification 3.7.7. Each control room ventilation system (train) is actuated by its own instrumentation channel. Consequently an inoperable channel of ventilation actuation instrumentation renders that system/train of ventilation inoperable and Specification 3.7.7 prescribes the appropriate action.

ACTION 28.a. provides 7 days to place the Control Room ventilation in the recirculation and make-up filtration mode of operation at 100% capacity (any two of the three trains of control room makeup and cleanup filtration meet the 100% capacity requirement) when one the two radioactivity high actuation channels is inoperable. This time is acceptable because there is still an operable channel that will function to realign the control room envelope on a high radiation signal unless the failure mode is due to the output power supply. However, in that case, the operator can manually initiate the function. The 7 day allowed outage time is based on the low probability of a Design Basis Accident (DBA) occurring during this time period, and ability of the remaining train to provide the required capability.

ACTION 28.b. applies when both channels of control room ventilation radioactivity-high are inoperable and requires the ventilation system to be placed in recirculation and make-up filtration within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. f* " t ......

61 t e ntroiul doom A.koup-aand-Cea Filtration System tb tho rpnirni 4ation nrao The additionalt restriction provides assurance that potential radiation releases from design basis accidents inside and outside containment have been considered 1k1_ for this configuration.

bai accienqsiresithe otside pand onacdinmn hav bOee cosiered for tehnisa Spcnfiguation.

Unit"2 - Amendme._._-6Pnt ACTION 28.c. applies for MODEs 1, 2, 3, & 4. It-------------------- No. l does not apply.

SOUTH TEXAS - UNITS 1 & 2 B 3/4 3-2d niUnit 21 _-AmnmnAmendment No.N°

NOC-AE-07002127 Attachment 3 INSTRUMENTATION BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) adioactk'ityjigat'io romý C3toi a~i~titia The Engineered Safety Features Actuation System interlocks perform the following functions:

P-4 Reactor tripped - Actuates Turbine trip via P-1 6, closes main feedwater valves on Tavg below Setpoint, prevents the opening of the main feedwater valves which were closed by a Safety Injection or High Steam Generator Water Level and allows Safety Injection block so that components can be reset or tripped. Reactor tripped with the source range blocked provides a non-protective function that closes the Steam Generator Blowdown isolation valves and allows reopening the valves after the source range block is reset.

Reactor not tripped - prevents manual block of Safety Injection.

P-1 1 On increasing pressurizer pressure, P-11 automatically reinstates Safety Injection actuation on low pressurizer pressure or low compensated steamline pressure signals, reinstates steamline isolation on low compensated steamline pressure signals, and opens the accumulator discharge isolation valves. On decreasing pressure, P-1 1 allows the manual block of Safety Injection actuation on low pressurizer pressure or low compensated steamline pressure signals, allows the manual block of steamline isolation on low compensated steamline pressure signals, and enables steam line isolation on high negative steam line pressure rate (when steamline pressure is manually blocked).

P-1 2 On increasing reactor coolant loop temperature, P-1 2 automatically provides an arming signal to the Steam Dump System. On decreasing reactor coolant loop temperature, P-12 automatically removes the arming signal from the Steam Dump System.

P-1 4 On increasing steam generator water level, P-14 automatically trips the turbine and the main feedwater pumps, and closes all feedwater isolation valves and feedwater control valves.

For Table 4.3-1 Notations 3 and 6, the term "incore" applies to either a PDMS measurement OR a movable incore detector system measurement, because both methods represent a measurement of the reactor core power distribution.

3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 (NOT USED)

SOUTH TEXAS - UNITS 1 & 2 B 3/4 3-3 Unit 1 - Amendment No.

N° 34.IC Unit 2 -Amendment No. ~~4

NOC-AE-07002127 Attachment 3 PLANT SYSTEMS BASES The limitations on minimum water level and maximum temperature are based on providing a 30-day cooling water supply to safety-related equipment without exceeding its design basis temperature and is consistent with the recommendations of Regulatory Guide 1.27, "Ultimate Heat Sink for Nuclear Plants," March 1974.

3/4.7.6 (NOT USED) 3/4.7.7 CONTROL ROOM MAKEUP AND CLEANUP FILTRATION SYSTEM The Control Room Makeup and Filtration System is comprised of three 50-percent redundant systems (trains) that share a common intake plenum and exhaust plenum. Each system/train is comprised of a makeup fan, a makeup filtration unit, a cleanup filtration unit, a cleanup fan, a control room air handling unit, a supply fan, a return fan, and associated ductwork and dampers.

Two of the three 50% design capacity trains are required to be operable during the following modes of operation: shutdown, hot standby, normal operation, postulated accident condition, and loss of offsite power. The toilet kitchen exhaust, heating, and computer room HVAC Subsystem associated with the Control Room Makeup and Filtration System are nonsafety-related and not required for operability.

The OPERABILITY of the Control Room Makeup and Cleanup Filtration System ensures that:

(1) the ambient air temperature does not exceed the allowable temperature for continuous-duty rating for the equipment and instrumentation cooled by this system, and (2) the control room will remain habitable for operations personnel during and following all credible accident conditions.

Operation of the system with the heaters operating for at least 10 continuous hours in a 92-day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem total effective osePuivalent TEDE) - . This limitation is consistent with the requirements of General Design Criterion 19 of Appendix A, 10 CFR Part 50. ANSI N510-1980 will be used as a procedural guide for surveillance testing.

ffh u--_m.4 A_ýrI0LWQ__peifid s--a" ithles t 4~r4Ai;J* re*a~__d_"+=

satnudafe operatio-n- Introdctionf nnof-Goolant-irvenfor -mus bý-fo-.u~st bor-nconeh-raio gAýI ater wht-wukfo ereguiredi the RC~S for minimumI fedhon inRSbrn-~cnrton butpoairovie aceptable margiqn to mportujc inca 4

o11e peratingwith apositi moderator temperature coefficicent-,must alo bp -evaft t

,"'ihdrawn-4t least gon stepý, How esne the~ GonG re rosir -abeie th SOUTH TEXAS -UNITS 1 & 2 B 3/47-4 Unit 1 -Amendment No. 5ý _

Unit 2 - Amendment No. 510ý

NOC-AE-07002127 Attachment 3 PLANT SYSTEMS BASES Th ea6~idenits Dpstula te d t~o- ccur _d'dRing cor-e -ilt-eraifbpsin-~ a~ddifti on -to' fi --iu-e-l _fianriliin6 accident, are: inadvertent 'criticality (due to a control rodl removal error or continuous rod Withdrawal error duringrefuelingior boron dilution)-and the inadvertent loading of, and subsequent operation with, fuiel assembly ih'an imoroperý location. These events are~ not ipostulated to result,in4 del Tcladding integrity damage. ý--Since the o,nly accident to occur, kurinq, CORE AlTERATIONSAthat'results in a significant radioactive release isthe*fuel handling accident and the accident m tqtoýfeturso theControl Room Makeup anid Cleanup Filtration Syste ,iare not credited in the accident analysis'for a fuel hiandlingi accidenm* there are -OPERABILITY:requirements for this system in-MODES 5 and 61 The time limits associated with the ACTIONS to restore an inoperable train to OPERABLE status are consistent with the redundancy and capability of the system and the low probability of a design basis accident while the affected train(s) is out of service. A limited allowed outage time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is allowed for all three trains to be out of service simultaneously in recognition of the fact that there are common plenums and some maintenance or testing activities required opening or entry into these common plenums. This time is reasonable to diagnose, plan, and possibly repair problems with the boundary or the ventilation system. This is acceptable based on the low probability of a design basis event in that brief allowed outage time and because administrative controls impose compensatory actions that reduce the already small risk associated with being in the ACTION. The compensatory actions are consistent with the intent of GDC 19 to protect plant personnel from potential hazards such as radioactive contamination, smoke, and temperature, etc. Pre-planned measures should be available to address these concerns for intentional and unintentional entry into the condition. The compensatory actions include:

" Procedures will preclude intentionally removing multiple trains of Control Room Envelope HVAC from service if Containment Spray is not functional or intentionally making a train of Containment Spray unavailable when multiple trains of Control Room Envelope HVAC are out of service. For purposes of this compensatory action, Containment Spray is considered functional if at least one train can be manually or automatically initiated.

• The plant will not make planned simultaneous entries into TS 3.7.7 ACTION c. for MODES 1,2, 3 a nd4and 9ACT10N-a-** .

The compensatory action may include placing fans in pull-to-lock as necessary to preclude there being a motive force to transport contaminated air to a clean environment in the event of an accident. These compensatory actions also include administrative controls on opening plenums or other openings such that appropriate communication is established with the control room to assure timely closing of the system if necessary. Since the Control Room Envelope boundary integrity also affects operability of the overall system, entry and exit is administratively controlled. Administrative control of entry and exit through doors is performed by the person(s) entering or exiting the area. Extended opening of the boundary is coordinated with the control room with appropriate plans for closure and communication.

Surveillance Requirement 4.7.7.e.3 verifies the integrity of the control room enclosure, and the assumed inleakage rates of the potentially contaminated air. The control room positive pressure, with respect to potentially contaminated adjacent areas, is periodically tested to verify proper functioning of the Control Room HVAC. During the emergency mode of operation, the SOUTH TEXAS - UNITS 1 & 2 B 3/4 7-5 Unit 1 - Amendment No.

Unit 2 - Amendment No. 0

NOC-AE-07002127 Attachment 3 PLANT SYSTEMS BASES Control Room HVAC is designed to pressurize the control room to at least 1/8 inch water gauge (in-wg) positive pressure with respect to adjacent areas in order to prevent unfiltered inleakage. The Control Room HVAC is designed to maintain this positive pressure with two trains at a makeup flow rate of 2000 cfm. The frequency of 18 months is consistent with the guidance provided in NUREG-0800. If the surveillance results are less than 1/8 in-wg and the pressure differential is not positive, the surveillance requirement is considered not met and the appropriate action of TS 3.7.7 must be applied.

,314.7.8 (Not used) FUEL HANDLIWO BUILJDING!EXH,51UST RSSE 4hee-'a he -exhaust- ester fans6.....share a .......

• .l . pn. An, .PERABLE voftlatiGý

ýlhaus4..ItriR-een its f~" 0PFRAEBl~rai R oxhau-st Rh~tb6t# an"OEAL OJ4.PE=RABIL1ITY of the.rFL16HaRGWdlrBidn ExastAr-36 nubsta fýilterp'd priorl m-a *hftg-the erlierel nmiRO h yt~.wt~p P eea 1hp 1P~qt 40confinu~ous hours i 1dyPrn qF~_~ hbJdp6,:R1tf ifl tI4AdsorbFso and HE-PA\ itrs,. The eperatiOR fc i ytma~ h eut4ý efiite .16cado calculationoh-S 1118,w 1b se 1npoao~,or a~ cpMD1Ratir oan oDoonst ','niioatlon-trc-4ýand an inoporadDio dxhausfi~raton Y~i 7Avc.With more than one, noper hip-train of; ofithor FHB oxhaus itrfo6FH-B-oxhalidfventtiafon or with _4mbk~t s involi'in moreothan E0eloprlý 6flthe~ ho oxhau'st -oft~hAnortho exh~aust f~itratifrV the-allbwo'd 6oulag-- +ino-ofdi

.4,w~ d hours. ou tiof 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is aflowod,ýD im formulipl ýains obo týM ioepoi- simaou~ 4n recognitiqp of tho fqP~t thaqt thoroý arecommoe43pto-nums_ andim s esnal~ dýýg-oeo plan, and possibly ropýairprobloms with the boundapy or thc SOUTH TEXAS - UNITS 1 & 2 B 3/4 7-6 Unit 1 - Amendment No.

Unit 2 - Amendment No.

NOC-AE-07002127 Attachment 3 PLANT SYSTEMS BASES Th G nc0Rator" acin rGonsise~to With the intent of GD 49 GD 6an Pr rs~nolandthopubic t e-ptential haardSc~ SWG41 adiati ve Gtam~

nd tmpoatur, oc. Po ~~nno mosurch solid be av.ailable to-addre~sq thp

ýGIGPRG8RS for intontiona1'and unintPR~ti Ra'!tr into the Gonditfioe Tho 6cor iR!d lacing tooka nlbsryý~r 6h~ aan miiv f~ll_

~ranporcontrn~atdai tb 'a cleanoviomoti the dvdent of an aGcidont'Th1 iýG Go 0.,.i aanGR PI,~ st-I n~ nanteal anon G at.~ anH -ý Pt

'heng~ syi~stem if'nocosrary., Sin& the Fuel Had~~giidingounda-ry pt' r tGoporability, of theo~~l ytM nr and exit iGAdMipisrtrAfi':lg~,y[ontroied

• dnppr at civocotrol of ontryand exit through doors is4efmle-d44-wetse exitingthe area. Extended Opeonirg of thle bou1-ndar is q~riaa 1 tpr~FlF 3/4.7.9 (Not Used) 3/4.7.10 (Not Used) 3/4.7.11 (Not used) 3/4.7.12 (Not used) 3/4.7.13 (Not used) 3/4.7.14 ESSENTIAL CHILLED WATER SYSTEM The OPERABILITY of the Essential Chilled Water System ensures that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analyses.

When a risk-important system or component (for example Essential Chilled Water) is taken out of service, it is important to assure that the impact on plant risk of this and other equipment simultaneously taken out of service is assessed. The Configuration Risk Management Program evaluates the impact on plant risk of equipment out of service. A brief description of the Configuration Risk Management Program is in Section 6.8.3 (administrative controls) of the Technical Specifications.

The extended allowed outage time (EAOT) of 7 days for one inoperable Essential Chilled Water System loop is based on establishing compensatory measures that are consistent with the Configuration Risk Management Program and are controlled by plant procedures to offset the risk impacts of entering the EAOT. Refer to the Bases for 3.8.1.1. Action b for further details.

SOUTH TEXAS - UNITS 1 & 2 B 3/4 7-7 Unit 1 - Amendment No.

Unit 2 -Amendment No. 1

NOC-AE-07002127 Attachment 3 ELECTRICAL POWER SYSTEMS BASES A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION (Continued)

The 10-year Frequency is consistent with the recommendations of Regulatory Guide 1.108, paragraph 2.b, and Regulatory Guide 1.137, paragraph C.2.f.

SR 4.8.1.1.2.9 This SR provided assurance that any accumulation of sediment over time or the normal wear on the system has not degraded the diesels.

The Surveillance Requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guides 1.9, "Selection of Diesel Generator Set Capacity for Standby Power Supplies," Revision 2, December 1979; 1.108, "Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants," Revision 1, August 1977; and ASTM D975-81, ASTM D1 552-79, ASTM D262282, ASTM D4294-83, and ASTM D2276-78. The standby diesel generators auxiliary systems are designed to circulate warm oil and water through the diesel while the diesel is not running, to preclude cold ambient starts. For the purposes of surveillance testing, ambient conditions are considered to be the hot prelube condition.

3.8.1.3 The OPERABILITY of the minimum AC sources during MODE 6 with >23' of water in the cavity is based on the following conditions:

a. The unit can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and No Changes on this Page I SOUTH TEXAS - UNITS 1 & 2 B 3/4 8-14 Unit 1 - Amendment No:

Unit 2 - Amendment No. 02-6209

NOC-AE-07002127 Attachment 3 ELECTRICAL POWER SYSTEMS BASES

c. Adequate AC electrical power is provided to mitigate events postulated during shutdowng In general, when the unit is shutdown, the Technical Specifications requirements ensure that the unit has the capability to mitigate the consequences of postulated accidents. However, assuming a single failure and concurrent loss of all offsite or all onsite power is not required.

The rationale for this is based on the fact that many Design Basis Accidents (DBAs) that are analyzed in MODES 1,2,3, and 4 have no specific analyses in MODES 5 and 6. Worst case bounding events are deemed not credible in MODES 5 and 6 because the reduced energy contained within the reactor pressure boundary, reactor coolant temperature and pressure, and the corresponding stresses result in the probabilities of occurrence being significantly reduced or eliminated, and in minimal consequences. These deviations from DBA analysis assumptions and design requirements during shutdown conditions are allowed by the LCO for required systems.

During MODES 1, 2, 3, and 4, various deviations from the analysis assumptions and design requirements are allowed within the Required Actions. This allowance is in recognition that certain testing and maintenance activities must be conducted provided an acceptable level of risk is not exceeded. During MODES 5 and 6, performance of a significant number of required testing and maintenance activities is also required. In MODES 5 and 6, the activities are generally planned and administratively controlled. Relaxations from MODES 1, 2, 3, and 4 LCO requirements are acceptable during shutdown modes based on:

a. U time in an outage - ". This is a riskP.prudent goal as well as a utility economic consideration.

b. Requiring appropriate compensatory measures for certain conditions. These may include administrative controls, reliance on systems that do not necessarily meet typical design requirements applied to systems credited in operating MODE analyses, or both.

c. Prudent utility consideration of the risk associated with multiple activities that could affect multiple systems.

d. Maintaining, to the extent practical, the ability to perform required functions (even if not meeting MODE 1,2, 3, and 4 OPERABILITY requirements) with systems assumed to function during an event.

In the event of an accident during shutdown, this LCO ensures the capability to support systems necessary to avoid immediate difficulty, assuming either a loss of all offsite power or a loss of all onsite diesel generator (DG) power.

3.8.2.1 In order to ensure the ability of the batteries to perform their intended function, the batteries are normally maintained in a fully charged state and the environment in which the batteries are located is maintained within the parameters used to determine battery sizing and SOUTH TEXAS - UNITS 1 & 2 B 3/4 8-15 Unit 1 - Amendment No. 1' Unit 2 - Amendment No.

NOC-AE-07002127 Attachment 3 ELECTRICAL POWER SYSTEMS BASES 3.8.3.2 The OPERABILITY of the required DC sources and electrical distribution system during shutdown is based on the following conditions:

a. The unit can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and

c. Adequate AC electrical power is provided to mitigate events postulated during shutdown.

In general, when the unit is shutdown, the Technical Specifications requirements ensure that the unit has the capability to mitigate the consequences of postulated accidents. However, assuming a single failure and concurrent loss of all offsite or all onsite power is not required.

The rationale for this is based on the fact that many Design Basis Accidents (DBAs) that are analyzed in MODES 1, 2,3, and 4 have no specific analyses in MODES 5 and 6. Worst case bounding events are deemed not credible in MODES 5 and 6 because the energy contained within the reactor pressure boundary, reactor coolant temperature and pressure, and the corresponding stresses result in the probabilities of occurrence being significantly reduced or eliminated, and in minimal consequences. These deviations from DBA analysis assumptions and design requirements during shutdown conditions are allowed by the LCO for required systems.

Specifications 3.8.2.2 and 3.8.3.2 require DC power sources and specified electric power distribution for equipment required to be operable during shutdown. If the DC sources or distribution system is inoperable, then the Specifications require the affected components to be declared inoperable or that core alterations and positive reactivity changes be stopped. For a required system or component to be operable, the definition of OPERABLE/OPERABILITY requires the availability of necessary support systems, instrumentation, and electrical power for the required system to meet the design basis requirements. In MODES 5 and 6, the design basis does not include single failure coincident with loss of off-site power. Consequently, where two trains or channels of equipment are required by the Technical Specifications during MODES 5 and 6, only one of the trains or channels is required to be backed by an emergency power source or battery. Inoperability of the battery for one channel or train does not affect components that have an operable battery on the other required channel or train. Required electric power distribution systems must be operable under accident conditions that are SOUTH TEXAS - UNITS 1 & 2 B 3/4 8-19 Unit 1 - Amendment No.

Unit 2 - Amendment No. k393-Q8

NOC-AE-07002127 Attachment 3 ELECTRICAL POWER SYSTEMS BASES applicable during shutdown, including seismic. For components that have only a detection function and no mitigation function during or after the accident, emergency power and safety related normal power are not required (e.g., Source Range instrumentation in Refueling Mode).

When the function of those components is lost, the required actions to suspend core alterations or positive reactivity changes preclude the accident the components would be required to detect.

The ACTIONS specified during shutdown with less than the minimum required power sources or distribution systems, include suspending operations involving positive reactivity additions that could result in loss of required SHUTDOWN MARGIN or refueling boron concentration necessary to assure continued safe operation. Introduction of coolant inventory must be from sources that have a boron concentration greater than what would be required in the RCS for minimum SHUTDOWN MARGIN or refueling boron concentration. This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation. Introduction of temperature changes, including temperature increases when operating with a positive moderator temperature coefficient, must also be evaluated to not result in operation below the required SHUTDOWN MARGIN or refueling boron concentration limits.

Control rod withdrawal is not allowed except that it is permissible to unlock the control rods for rapid refueling. To unlock the control rods, they must be withdrawn at least one step. However, since the control rods are above the active fuel when the unlocking process occurs, there is no reactivity addition.

3/4.8.4 (Not Used)

I No Changes on this Page I SOUTH TEXAS - UNITS 1 & 2 B 3/4 8-20 Unit 1 -Amendment No. 05-1034-10 Unit 2 - Amendment No. 05-1034-10

NOC-AE-07002127 Attachment 3 3/4.9 REFUELING OPERATIONS BASES 3/4.9.2 INSTRUMENTATION The OPERABILITY of the Source Range and/or Extended Range Neutron Flux Monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

ACTION a. requires suspending the introduction into the RCS of coolant with boron concentration less than required to meet the refueling boron concentration limit necessary to assure continued safe operation. Introduction of coolant inventory must be from sources that have a boron concentration greater than what would be required in the RCS for minimum refueling boron concentration. This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation. Introduction of temperature changes, including temperature increases when operating with a positive moderator temperature coefficient, must also be evaluated to not result in operation below the required refueling boron concentration limit. Control rod withdrawal is not allowed except that it is permissible to unlock the control rods for rapid refueling. To unlock the control rods, they must be withdrawn at least one step. However, since the control rods are above the active fuel when the unlocking process occurs, there is no reactivity addition.

3/4.9.3 (Not Used) 314.9A (Not Used) CONTAINMENT BUILDING PEN~~ATIONS irmoson Gnnf4ýMent bi ilding penEtratiof cIslosur an OEABLTY nc9 c& ha in (3o~taipment will h thi4~a~~ rmlaaet h pie r~eipA-~erne-Ao rfosorilacs dinMODES 1,22,3-4 eperatiOR T-ho EoquiPME)t ha;tch ig geni~

klold toh 1cs nd-sealed dur Rg MODES 1,2,+3, and 4. Durng perid os-f shutdei-vwhe "GaiRIO~ GFtASurE) otiPG~

roth the my br9'ed to allow passage o aoi Art ý;t t P p~~~~~~~~~

5nn+mnann~~nn 'E' 2122 tf+4ý-_94o t4an4aysl abvte Vis i4 i.e.. tls dnriuilco,.... J.tt' o Gables or hoese rU Iogh the P~rGORR91 iclGk- 4ap ýt I faewithin 3_0 Minutpe4 f nififl o ftlhn~n~acdn nnid~taiýneo-44-4~a~oFiash-V tical fo Iocthn15

~ G hAourr' Fi el mROVGMent~ npoifd w th the eaoqter h nboosub(itica1 for 295 hours0.00341 days <br />0.0819 hours <br />4.877645e-4 weeks <br />1.122475e-4 months <br />. ifthe reacioE) hag boon isubcritical fE) 165 heur or8G

~Fe-~ ~ ~ pij~~sonnel afic c0sQ6ll cRo Soo a[; racjicabloebWt is AR"Mbd t SOUTH TEXAS - UNITS 1 & 2 B 3/4 9-l b Unit 1 - Amendment No. 0Z 1 Unit 2 -Amendment No. 0- a41

NOC-AE-07002127 Attachment 3 3/4.9 REFUELING OPERATIONS BASES

-flat L.a st .~ flfi iii . St~St< ne.IrrnA, fl& Sfl~ y~ ~§~r hil III Lil"I" BASES + wM may atse )pe dy~i.A V GQE.SSR4 t*e tG 1.SNASt Wh.ý hours and (2) the eationjhat wo4ldPccrin thobl ovntfbenfuo e floiR handlng ccien ul gR~r-idpR6'R r.q'-pm nt The, _________g AG ilapywhqý@ ____itaiviýi6 2.d d du6 are~i de~lntee aiihisRd Fead ccaviired Wak GloEco ther heqocurene f

,With eg~i~pMont atch~open, if,the Feactor has not boon subcritical for 1165 houre. Eguipme;4 shuIi occr ~b~nas practi~abI, anRd is normal"' assumzed to occurF ;n P hAuIc

~iachc~s~ro Unlike th 4ireek,4hJ eJILLe~ hatJl'm bc6e V byl~Le t~ilo (6L~.~.9i.JiJJ 3/4.9.5 (Not Used) 3/4.9.6 (Not Used) 3/4.9.7 (Not Used)

SOUTH TEXAS - UNITS 1 & 2 B 3/4 9-2 Unit 1 - Amendment No. r----4 6-5-4:14 Unit 2- Amendment No. §&4Q 4-1

NOC-AE-07002127 Attachment 3 REFUELING OPERATIONS BASES 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION As many as three residual heat removal (RHR) loops may be in operation, but at least one loop must be in operation at all times. One loop in operation ensures that: (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor vessel below 140°F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the core to minimize the effect of a boron dilution incident and prevent boron stratification.

The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability. With the reactor vessel head removed and at least 23 feet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

ACTIONS applicable when no RHR loop is in operation require suspending the introduction into the RCS of coolant with boron concentration less than required to meet the refueling boron concentration limit necessary to assure continued safe operation. Introduction of coolant inventory must be from sources that have a boron concentration greater than what would be required in the RCS for minimum refueling boron concentration. This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation. Introduction of temperature changes, including temperature increases when operating with a positive moderator temperature coefficient, must be evaluated to not result in operation below the required refueling boron concentration limit.

The 3000 gpm flow rate in 4.9.8.2 refers to total RHR flow through the core, i.e., cold leg injection flow. (CR 97-908-5)

,3I4.9.9#(Nt -edOT;.iT.N.1.ENT l:'E *NTI I*STION SYSTEM LATlOFSoC T 3/4.9.10 and 3/4.9.11 WATER LEVEL - REFUELING CAVITY AND STORAGE POOLS The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the safety analysis.

SOUTH TEXAS - UNITS 1 & 2 B 3/4 9-3 Unit 1 - Amendment No. 7 Unit 2 - Amendment No. -.!9Q)

NOC-AE-07002127 Attachment 3 REFUELING OPERATIONS BASES 3/4.9.1t2 (Not used) FUELI HANDLTNG BUILDING2PA

£XRAU AIRý SYf SOUTH TEXAS - UNITS 1 & 2 B 3/4 9-3a Unit 1 - Amendment No.

Unit 2 - Amendment No.

NOC-AE-07002127 Attachment 4 Attachment 4 Planned Changes to the Technical Requirements Manual (information only)

TRM 3/4.9.4 - one page TRM 3/4.9.12 - two pages TRM 3/4.9.14 - two pages TRM 3/4.9.15 - one page TRM Bases - three pages

NOC-AIE-07002127 Attachment 4 REFUELING OPERATIONS 3/4.9.12 CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITIONS FOR OPERATIONS 3.9.4 The containment building penetrations shall be in the following status:

a. The equipment hatch closed and held in place by a minimum of four bolts OR

1) The Reactor has been subcritical for Ž42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br />, AND If open, the equipment hatch is capable of being closed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

b. A minimum of one door in the containment Auxiliary Airlock (AAL) and a minimum of one door in the containment Personnel Airlock (PAL) are closed.

OR The water level is _>23 feet above the reactor vessel flange.

AND The Reactor has been subcritical for > 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> AND Individuals are available to close a PAL door and AAL door when directed (after the initiation of a fuel handling accident inside containment) as soon as possible but within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

c. All other penetrations providing direct access from the containment atmosphere to the outside atmosphere shall be either:

1) Closed by an isolation valve, blind flange, or manual valve, or

2) Be capable of being closed (after the initiation of a fuel handling accident inside containment) as soon as possible but within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

APPLICABILITY: During movement of irradiated fuel within the containment.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving movement of irradiated fuel in the containment building.

SURVEILLANCE REQUIREMENTS 4.9.4 Each of the above required containment building penetrations shall be determined to be either in its required condition or capable of being closed as required in specification 3.9.4 within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of and at least once per 7 days during movement of irradiated fuel in the containment building by (as applicable):

a. Verifying the penetrations are in their required condition or capable of being placed in their required condition.

b. Proper tools are staged and trained personnel are designated to close the equipment hatch if open.

SOUTH TEXAS - UNITS 1 & 2 3/4 9-xx TRM

NOC-AE-07002127 Attachment 4 REFUELING OPERATIONS 3/4.9.12 FUEL HANDLING BUILDING EXHAUST AIR SYSTEM LIMITING CONDITION FOR OPERATION 3.9.12 One FHB Exhaust Air Train 1 shall be OPERABLE OR If not OPERABLE, capable of being restored to an OPERABLE status within two hours APPLICABILITY: During the movement of fuel within the spent fuel pool or when conducting crane operation with loads over the spent fuel pool.

ACTION: With no FHB exhaust air train OPERABLE or capable of being restored to an OPERABLE status within two hours, suspend all operations involving movement of fuel within the spent fuel pool or crane operation with loads over the spent fuel pool.

SURVEILLANCE REQUIREMENTS 4.9.12 The FHB Exhaust Air System shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and~verifying that the system operates for at least 10 continuous hours with the heaters operating with the operable exhaust booster fans and the operable main exhaust fans operating to maintain adequate air flow rate;

b. At least once per 18 months and (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system by:

1) Verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% for HEPA filter banks and 0.10% for charcoal adsorber banks and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 29,000 cfm + 10%;

2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of ASTM D3803-1989, "Standard Test Method for Nuclear-Grade Activated Carbon," for a methyl iodide penetration of less than 1.0% when tested at a temperature of 30 0 C and a relative humidity of 70%; and At least one FHB exhaust air filter train, one FHB exhaust booster fan, and one FHB main exhaust fan are capable of being powered from an OPERABLE onsite emergency power source.

SOUTH TEXAS - UNITS 1 & 2 3/4 9-xx TRM

NOC-AE-07002127 Attachment 4 REFUELING OPERATIONS 3/4.9.12 FUEL HANDLING BUILDING EXHAUST AIR SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

3) Verifying a system flow rate of 29,000 cfm + 10% during system operation with two of the three exhaust booster fans and two of the three main exhaust fans operating when tested in accordance with ANSI N510-1980.

All combinations of two exhaust booster fans and two main exhaust fans shall be tested.

c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of ASTM D3803-1989 for a methyl iodide penetration of less than 1.0% when tested at a temperature of 3000 and a relative humidity of 70%.

d. At least once per 18 months by:

1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the system at a flow rate of 29,000 cfm + 10%,

2) Verifying that the system maintains the spent fuel storage pool area at a negative pressure of greater than or equal to 1/8 inch Water Gauge relative to the outside atmosphere during system operation, and

3) Verifying that the heaters dissipate 38 + 2.3 kW when tested in accordance with ANSI N510-1980.*

e. After each complete or partial replacement of a HEPA filter bank, by verifying that the HEPA filter bank satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N51 0-1980 for a DOP test aerosol while operating the system at a flow rate of 29,000 cfm + 10%.

f. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the charcoal adsorber bank satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.10% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 29,000 cfm + 10%.

SOUTH TEXAS - UNITS 1 & 2 3/4 9-xx. TRM

NOC-AE-07002127 Attachment 4 REFUELING OPERATIONS 3/4.9.14 CONTROL ROOM MAKEUP AND CLEANUP FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.14 One Control Room Makeup and Cleanup Filtration System shall be OPERABLE OR Ifnot OPERABLE, capable of being restored to an OPERABLE status within two hours.

APPLICABILITY: During the movement of irradiated fuel or when conducting crane operation with loads over the spent fuel pool.

ACTION: With no Control Room Makeup and Cleanup Filtration Systems OPERABLE or capable of being restored to an OPERABLE status within two hours, suspend all operations involving movement of irradiated fuel and crane operation with loads over the spent fuel pool.

SURVEILLANCE REQUIREMENTS 4.9.14 Each Control Room Makeup and Cleanup Filtration System shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the control room air temperature is less than or equal to 780 F;

b. At least once per 92 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers of the makeup and cleanup air filter units and verifying that the system operates for at least 10 continuous hours with the makeup filter unit heaters operating;

c. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system by:

1) Verifying that the makeup and cleanup systems satisfy the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% for HEPA filter banks and 0.10% for charcoal adsorber banks and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 6000 cfm + 10% for the cleanup units and 1000 cfm + 10% for the makeup units;

2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of ASTM D3803-1989, "Standard Test Method for Nuclear-Grade Activated Carbon," for a methyl iodide penetration of less than 1.0% when tested at a temperature of 300 C and a relative humidity of 70%; and

3) Verifying a system flow rate of 6000 cfm 10% for the cleanup units and 1000 cfm +

10% for the makeup units during system operation when tested in accordance with ANSI N510-1980.

SOUTH TEXAS - UNITS 1 & 2 3/4 9-xx TRM

NOC-AE-07002127 Attachment 4 REFUELING OPERATIONS 3/4.9.14 CONTROL ROOM MAKEUP AND CLEANUP FILTRATION SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

d. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of ASTM D3803-1989 for a methyl iodide penetration of less than 1.0% when tested at a temperature of 300C and a relative humidity of 70%.

1

e. At least once per 18 months by:

1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6.1 inches Water Gauge for the makeup units and 6.0 inches Water Gauge for the cleanup units while operating the system at a flow rate of 6000 cfm + 10% for the cleanup units and 1000 cfm

& 10% for the makeup units.

2) Verifying that the system maintains the control room envelope at a positive pressure of greater than or equal to 1/8 inch Water Gauge at less than or equal to a pressurization flow of 2000 cfm relative to adjacent areas during system operation; and

3) Verifying that the makeup filter unit heaters dissipate 4.5 + 0.45 kW when tested in accordance with ANSI N510-1980.

f. After each complete or partial replacement of a HEPA filter bank, by verifying that the HEPA filter bank satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0,05% in accordance with ANSI N510-1980 for a DOP test aerosol while operating the system at a flow rate of 6000 cfm + 10% for the cleanup units and 1000 cfm + 10% for the makeup unIts; and

g. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the charcoal adsorber bank satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.10% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 6000 cfm + 10% for the cleanup units and 1000 cfm + 10% for the makeup units.

SOUTH TEXAS - UNITS 1 & 2 3/4 9-xx TRM

NOC-AE-07002127 Attachment 4 REFUELING OPERATIONS 3/4.9.15 RADIOACTIVE MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.15 The Unit Vent Gas Activity Monitor RT-8010B shall be functional.

APPLICABILITY: During the movement of irradiated fuel or when conducting crane operation with loads over the spent fuel pool.

ACTION: With the requirements of the above specification not met, ensure that a method is available for monitoring a radioactive release in the event of a fuel handling accident OR suspend operations involving the movement of irradiated fuel or loads over the spent fuel pool.

SURVEILLANCE REQUIREMENTS 4.9.15 Surveillance testing for the Unit Vent Gas Activity Monitor RT-8010B will be in accordance with the methodology in the ODCM.

SOUTH TEXAS - UNITS 1 & 2 3/4 9-xx TRM

NOC-AE-07002127 Attachment 4 REFUELING OPERATIONS BASES 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS Containment building penetration closure is not credited or required as a mitigation function to meet the accident analyses. The OPERABILITY requirements provide the restoration of a monitored release path as a defense-in-depth measure to mitigate the consequences of a postulated FHA. This is consistent with NUMARC 93-01, Revision 3, Section 11.2.6, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants", subheading "Containment - Primary (PWR)/Secondary (BWR)." The closure requirement is a regulatory commitment for licensing the Alternative Source Term. (REF: Licensing Amendments No. xx and No. xx for Units 1 and 2 respectively)

The containment personnel airlock and auxiliary airlock, which are part of the containment pressure boundary, provide a means for personnel access during MODES 1, 2, 3, and 4 operation.

The equipment hatch is required to be closed and sealed during MODES 1, 2, 3, and 4. During periods of shutdown, when containment closure is not required, the equipment hatch may be opened to allow passage of material needed to support activities in the containment building. The personnel and auxiliary airlock door interlock mechanisms may be disabled during shutdown, allowing both airlock doors to remain open for extended periods when frequent containment entry is necessary.

Both containment personnel airlock doors and/or auxiliary airlock doors may be open during the movement of irradiated fuel when specific limitations are satisfied. The specification requires: (1) there is 23 feet of water above the reactor vessel flange, (2) the reactor has been subcritical for 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br />, (3) one airlock door in each containment entry point is OPERABLE and, (4) an individual is available to close one door in each entry point (if open) following a fuel handling accident inside containment.

The requirement to have 23 feet of water above the reactor vessel flange is consistent with the fuel handling accident analysis assumptions, Regulatory Guide 1.183, and Technical Specification 3.9.10, Water Level - Refueling Cavity.

Operability of each airlock entry door requires that the doors are capable of being closed, i.e., that the door is unblocked, no cables or hoses run through the airlock, and at least one door seal is capable of being inflated. Containment airlock door closure should occur as soon as practicable, but within at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

The equipment hatch may also be open during the movement of irradiated fuel when specific limitations are satisfied. The specification requires: (1) the reactor has been subcritical for 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> and, (2) the equipment hatch (if open) is capable of being closed following a fuel handling accident inside containment. The following administrative requirements will apply whenever the equipment hatch is open during the movement of irradiated fuel in containment:

1. Appropriate personnel are aware of the open status of the containment during movement of irradiated fuel.

2. Specified individuals are designated and readily available to close the equipment hatch following an evacuation that would occur in the event of a fuel handling accident

3. Obstructions (e.g., cables, hoses, and runway) that would prevent closure of the equipment hatch can be quickly removed.

SOUTH TEXAS - UNITS 1 & 2 B 3/4 9-xx TRM Amendment No.

NOC-AE-07002127 Attachment 4 REFUELING OPERATIONS BASES 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS (continued)

The containment equipment hatch closure is required to take place upon the occurrence of a fuel handling accident inside containment if the hatch is open. Fuel movement is not permitted with the equipment hatch open ifthe reactor has not been subcritical for 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br />. Equipment hatch closure should occur as soon as practicable, and is normally assumed to occur in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Unlike the airlock, the equipment hatch may be blocked by an obstruction (e.g. the removable equipment hatch runway). Fuel movement is not allowed with the runway installed unless the capability to remove all obstructions and close the hatch within the required time is maintained.

A surveillance requirement verifies that the proper tools are staged at the equipment hatch location and qualified personnel assigned to close the equipment hatch on a seven-day frequency.

3/4.9.12 FHB EXHAUST AIR SYSTEM The FHB exhaust air system is comprised of two independent exhaust air filter trains and three exhaust ventilation trains. Each of the three exhaust ventilation trains has a main exhaust fan, an exhaust booster fan, and associated dampers. The main exhaust fans share a common plenum and the exhaust booster fans share a common plenum. An OPERABLE FHB Exhaust Air Train consists of any OPERABLE exhaust filter train, any OPERABLE main exhaust fan, any OPERABLE exhaust booster fan and appropriate OPERABLE dampers.

The Fuel Handling Building Exhaust Air System is not credited or required as a mitigation function to meet the accident analyses. The OPERABILITY requirements provide the restoration of a filtered release path as a defense-in-depth measure to mitigate the consequences of a postulated FHA. This is consistent with NUMARC 93-01, Revision 3, Section 11.2.6, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants", subheading "Containment - Primary (PWR)/Secondary (BWR)." The OPERABILITY requirement is a regulatory commitment for licensing the Alternative Source Term. (REF: Licensing Amendments No. xx and No. xx for Units 1 and 2 respectively)

Operation of the system with the heaters operating for at least 10 continuous hours in a 31-day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. The iodine removal capacity of the system is consistent with NRC RG 1.52. ANSI N510-1980 will be used as a procedural guide for surveillance testing. This Specification has been modified by a note that states, at least one FHB exhaust air filter train, one FHB exhaust booster fan, and one FHB main exhaust fan are capable of being powered from an Onsite emergency power source. This note ensures that required FHB exhaust train components will have an emergency power source available.

3/4.9.14 CONTROL ROOM MAKEUP AND CLEANUP FILTRATION SYSTEM The Control Room Makeup and Cleanup Filtration System is comprised of three 50-percent redundant systems (trains) that share a common intake plenum and exhaust plenum. Each system/train is comprised of a makeup fan, a makeup filtration unit, a cleanup filtration unit, a SOUTH TEXAS - UNITS 1 & 2 B 3/4 9-xx TRM Amendment No.

NOC-AE-07002127 Attachment 4 REFUELING OPERATIONS BASES 3/4.9.14 CONTROL ROOM MAKEUP AND CLEANUP FILTRATION SYSTEM (continued) cleanup fan, a control room air handling unit, a supply fan, a return fan, and associated ductwork and dampers.

The Control Room Makeup and Cleanup Filtration System is not credited or required as a mitigation function to meet the accident analyses for a fuel handling accident. The OPERABILITY requirement during the movement of irradiated fuel or when irradiated fuel is in the spent fuel pool provides for the restoration of a filtered path as a defense-in-depth measure to further lower the consequences to the control room operator from a postulated FHA. This is consistent with NUMARC 93-01, Revision 3, Section 11.2.6, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants", subheading "Containment - Primary (PWR)/Secondary (BWR)" The OPERABILITY requirement is a regulatory commitment for licensing the Alternative Source Term. (REF: Licensing Amendments No. xx and No. xx for Units 1 and 2 respectively) 3/4.9.15 RADIOACTIVE MONITORING INSTRUMENTATION The Radioactive Monitoring Instrumentation provides the capability of monitoring a release from a fuel handling accident in either the Reactor Containment Building or the Fuel Handling Building.

This is consistent with NUMARC 93-01, Revision 3, Section 11.2.6, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants", subheading "Containment -

Primary (PWR)/Secondary (BWR)". Maintaining this instrumentation functional when moving irradiated fuel or performing operations involving crane operation with loads over the spent fuel pool is a regulatory commitment for licensing the Alternative Source Term. (REF: Licensing Amendments No. xx and No. xx for Units 1 and 2 respectively)

SOUTH TEXAS - UNITS 1 & 2 B 3/4 9-xx TRM Amendment No.

NOC-AE-07002127 Attachment 5 List of Commitments

NOC-AE-07002127 Attachment 5 Page 1 List of Commitments The following table identifies those actions committed to by STPNOC in this document. Any statements in this submittal with the exception of those in the table below are provided for information purposes and are not considered commitments. Please direct questions regarding these commitments to Ken Taplett at (361) 972-8416.

Using insights from TSTF-51, and consistent with the guidance in NUMARC 93-01, Revision 3, Section 11.3.6.5, "Safety Assessment for Removal of Equipment from Service During Shutdown Conditions," subheading "Containment - Primary (PWR)/Secondary (BWR),"

STPNOC makes the following commitments to mitigate the consequences of a potential fuel handling accident.

NOTE: The purpose of these commitments are to maintain the Fuel Handling Building (FHB)

Ventilation System and associated radiation monitoring availability to reduce doses even further below that provided by the natural decay and to avoid unmonitored releases; and to enable the FHB Ventilation System to draw the release from a postulated fuel handling accident in the FHB in the proper direction such that it can be treated and monitored.

In addition, the purpose of these commitments is to isolate the Reactor Containment Building (RCB) for postulated fuel handling accident in the RCB and further reduce dose by natural decay and to enable the Ventilation System to draw the release from a postulated fuel handling accident in the containment in the proper direction such that it can be monitored (not treated).

Commitment Continuing Scheduled Compliance Completion Date

1. Whenever fuel is being moved in the spent fuel X Upon pool or when conducting crane operation with Implementation loads over the spent fuel pool, at least one train of FHB Exhaust Air shall be OPERABLE or capable of being restored to an OPERABLE status within two hours .

2. Whenever irradiated fuel is being moved or X Upon when conducting crane operation with loads Implementation over the spent fuel pool, at least one Control Room Makeup and Cleanup Filtration System shall be OPERABLE or capable of being restored to an OPERABLE status within two hours'.

NOC-AE-07002127 Attachment 5 Page 2 Commitment Continuing Scheduled Compliance Completion Date

3. Whenever irradiated fuel is being moved within X Upon the Reactor Containment Building, the Implementation following will be closed or capable of being closed within two hours'.

a. The equipment hatch

b. At least one door in the Auxiliary Airlock and one door in the Personnel Airlock.

c. All other penetrations providing direct access from the containment atmosphere to the outside atmosphere.

4. Within two hours' of a fuel handling accident X Upon in the FHB, at least one train of FHB Exhaust Implementation Air will be placed in operation.

5. Within two hours' of a fuel handling accident, X Upon at least one Control Room Makeup and Implementation Cleanup Filtration System will be placed in operation.

6. Within two hours' of a fuel handling accident X Upon in the Reactor Containment Building, the Implementation following actions will be taken:

a. Close the equipment hatch,

b. Close at least one of the Auxiliary Airlock doors and one of the Personnel Airlock doors, and

c. Close all other penetrations providing direct access from the containment atmosphere to the outside atmosphere.

7. Whenever irradiated fuel is being moved or X Upon when conducting crane operation with loads Implementation over the spent fuel pool, radiation monitoring instrumentation will remain functional to ensure that a release following a fuel handling accident is monitored.

The two hours to restore the FHB Exhaust Air System and the Control Room Makeup and Cleanup Filtration System to OPERABLE status and to close containment penetrations or openings in the event of a fuel handling accident is reasonable because these systems are not required to mitigate the accident. These systems are not credited in the accident analyses. Dose limits are within requirements assuming an instantaneous release from the FHA. These additional administrative actions are taken to further filter and monitor the release as a defense-in-depth measure.

NOC-AE-07002127 Attachment 5 Page 3 For License Amendments 139/128, Units 1 and 2 respectively, two hours to close the equipment hatch was acceptable to meet the dose guidelines of 10 CFR 100 and General Design Criterion

19. For this proposed licensing amendment request, the dose guidelines of 10 CFR 100 and General Design Criterion 19 are met without restoring the systems described above. Therefore, two hours is a reasonable time to put these defense-in-depth measures in place.

Reference:

South Texas Project, Units 1 and 2 - Issuance of Amendments 139/128, Units 1 and 2 respectively, dated July 18, 2003 on Equipment Hatch Open During Refuel Operations (TAC NOS. MB3587 and MB3591)

Other Commitment Commitment Continuing Scheduled Compliance Completion Date

8. Until a plant modification is completed for X Upon supporting the limiting single failure Implementation assumptions in the steam generator tube rupture (SGTR) analysis, STP will maintain an administrative limit for reactor coolant system dose equivalent iodine so that the radiological dose limits for the SGTR analysis remain bounding.

NOC-AE-07002127 Attachment 6 Attachment 6 Regulatory Guide 1.183 Conformance Tables 1 of 37

NOC-AE-07002127 Attachment 6 REGULATORY GUIDE 1.183 CONFORMANCE TABLE Notes: a Any reference to Tables or Sections in this column refers to Attachment 1, Section 4.0, "Technical Analysis" of the Licensee Evaluation Table A: Conformance with Reeulatorv Guide 1.183 Main Sections RG Section Regulatory Position Analysis Comments a 3.1 - AST The inventory of fission products in the reactor core and Conforms The core power level assumed for the DBA Fission available for release to the containment should be based analyses is 4100 MWt. This is greater than the Product on the maximum full power operation of the core with, TS rated thermal power of the core of 3853 Inventory as a minimum, current licensed values for fuel MWt with an uncertainty factor of 0.6%. The enrichment, fuel burnup, and an assumed core power peak burnup assumed is 60030 MWD/MTU.

equal to the current licensed rated thermal power times The inventory of fission products is based on the ECCS evaluation uncertainty. The uncertainty factor the current licensed values for fuel used in determining the core inventory should be that enrichment. (Section 4.2.4.2) value provided in Appendix K to 10 CFR Part 50, typically 1.02.

The period of irradiation should be of sufficient duration The assumed period of irradiation was to allow the activity of dose-significant sufficient (three-region equilibrium cycle core radionuclides to reach equilibrium or to reach maximum at end of life with the three regions having values. operated at 39.31 MW/MTU for 509, 1018, and 1527 EFPD, respectively) to allow the activity of dose-significant radionuclides to reach equilibrium or to reach maximum values. (Section 4.2.4.2)

The core inventory should be determined using an The ORIGEN 2.1 code was used to calculate appropriate isotope generation and depletion computer plant-specific fission product inventories.

code such as ORIGEN 2 or ORIGEN-ARP. (Section 4.2.4.2) 2 of 37

NOC-AE-07002127 Attachment 6 Table A: Conformance with Regulatory Guide 1.183 Main Sections RG Section Regulatory Position Analysis Comments a For the DBA LOCA, all fuel assemblies in the core are For the DBA LOCA, all fuel assemblies in the assumed to be affected and the core average inventory core are assumed to be affected and the core should be used. For DBA events that do not involve the average inventory is used. (Section 4.3.5) entire core, the fission product inventory of each of the The fission product inventory for the FHA is damaged fuel rods is determined by dividing the total provided in Table 4.4-1. The core inventory core inventory by the number of fuel rods in the core. To for each fission product in Table 4.4-1 is account for differences in power level across the core, multiplied by the peaking factor of 1.7, which radial peaking factors from the facility's core operating bounds values in the Core Operating Limits limits report (COLR) or technical specifications should Report, and by the fraction of the fuel in the be applied in determining the inventory of the damaged core that is damaged (314 pins out of 50952).

rods. (Section 4.4.3) For the remaining accidents, the fuel damage is specified in the accident discussion in Section 4.

No adjustment to the fission product inventory should be All accident analyses were performed made for events postulated to occur during power assumed 4100 Mwt. Rated full power for the operations at less than full rated power or those STP units is 3853 Mwt. Each accident postulated to occur at the beginning of core life. analyses in Section 4.0.

For events postulated to occur while the facility is A radioactive decay time of 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> while shutdown, e.g., a fuel handling accident, radioactive the facility is shutdown in modeled in the decay from the time of shutdown may be modeled. FHA. (Section 4.4.3) 3.2 - Release The core inventory release fractions, by radionuclide Conforms The DBA LOCA release fractions given as fractions groups, for the gap release and early in-vessel damage release rates over the given duration are found phases for DBA LOCAs are listed in Table 2 for PWRs. in Table 4.3-10.

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

For non-LOCA events, the fractions of the core For non-LOCA events, the fractions of the 3 of 37

NOC-AE-07002 127 Attachment 6 Table A: Conformance with Regulatorv Guide 1.183 Main Sections Attachment 6 NOC-AE-07002127 RG Section Regulatory Position Analysis Comments a inventory assumed to be in the gap for the core inventory assumed to be in the gap for various radionuclides are given in Table 3. The release the various radio-nuclides are based on Table fractions from Table 3 are used in conjunction with the 3. See sections 4.4.3 (FHA), 4.7.3.1.1 fission product inventory calculated with the maximum (CREA), and 4.8.3 (LRA). The MSLB and core radial peaking factor. SGTR accidents do not assume clad damage.

3.3 - Timing The activity released from the core during each release Conforms The DBA LOCA release fractions given as of Release phase should be modeled as increasing in a linear release rates over the given duration are found Phases fashion over the duration of the phase. in Table 4.3-10. The LOCA activity released from the core is modeled in a linear fashion over the duration of the release phases.

For non-LOCA DBAs in which fuel damage is This is true, except that the elemental and projected, the release from the fuel gap and the fuel particulate iodines released from the fuel for pellet should be assumed to occur instantaneously with an accident-induced spike during a SGTR are the onset of the projected damage. released into the RCS over time. (4.6.3.1.1)

For facilities licensed with leak-before-break Leak before break is not credited in the AST methodology, the onset of the gap release phase may be analyses.

assumed to be 10 minutes. A licensee may propose an alternative time for the onset of the gap release phase, based on facility-specific calculations using suitable analysis codes or on an accepted topical report shown to be applicable to the specific facility. In the absence of approved alternatives, the gap release phase onsets in Table 4 should be used.

3.4 - Table 5 lists the elements in each radionuclide group that Conforms The fission product inventory for the LOCA Radionuclide should be considered in design basis analyses. is listed in Table 4.3-1. This table does not list Composition Br, Se, Pd, Co, Eu, Pm & Sm - elements listed in Table 5 of RG 1.183 In addition to the radionuclides appearing in 4 of 37

NOC-AE-07002127 Attachment 6 Table A: Conformance with Regulatory Guide 1.183 Main Sections RG Section Regulatory Position Analysis I Comments a the RADTRAD list, Kr83m, Xel3 lm, Xel33m, and Xe135m were added for dose analysis purposes based on their inclusion in TID-14844. Xe138 was also added. Co58 and Co60 were deleted from the list because only 63 radionuclides can be used. A study performed for another licensee indicated that omitting Co58 and Co 60 decreased the control room dose by about 0.01 percent while adding the noble gas isotopes increased the control room dose by about 0.1 percent.

(Section 4.3.2) 3.5- Of the radioiodine released from the reactor coolant Conforms 95 percent of the iodine released from the Chemical system (RCS) to the containment in a postulated reactor coolant system to the containment is Form accident, 95 percent of the iodine released should be assumed to be cesium iodide (CsI), 4.85 assumed to be cesium iodide (CsI), 4.85 percent percent elemental iodine, and 0.15 percent elemental iodine, and 0.15 percent organic iodide. This organic iodide. This includes releases from includes releases from the gap and the fuel pellets. With the gap and the fuel pellets. Fission products the exception of elemental and organic iodine and noble are assumed to be in particulate form, gases, fission products should be assumed to be in with the exception of elemental and organic particulate form. iodine and noble gases.

(Section 4.3.5)

The same chemical form is assumed in releases from The same chemical form is assumed in fuel pins in FHAs and from releases from the fuel pins releases from fuel pins in FHAs and from through the RCS in DBAs other than FHAs or LOCAs. releases from the fuel pins through the RCS in However, the transport of these iodine species following DBAs other than FHAs or LOCAs. The release from the fuel may affect these assumed fractions. accident-specific appendices to this regulatory The accident-specific appendices to this regulatory guide guide provide additional details.

provide additional details.

3.6 - Fuel The amount of fuel damage caused by non-LOCA Conforms The non-LOCA design bases analyses used 5 of 37

NOC-AE-07002127 Attachment 6 Table A: Conformance with Regulatory Guide 1.183 Main Sections RG Section Regulatory Position Analysis Comments a Damage in design basis events should be analyzed to determine, for DNBR as a fuel damage criterion. (Section Non-LOCA the case resulting in the highest radioactivity release, the 4.2.4.2)

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

4.1 - Offsite Dose Consequences The following assumptions should be used in determining the TEDE for persons located at or beyond the boundary of the exclusion area (EAB) 4.1.1 The dose calculations should determinethe TEDE. Conforms TEDE is calculated, with significant progeny TEDE is the sum of the committed effective dose included.

equivalent (CEDE) from inhalation and the deep dose equivalent (DDE) from external exposure. The The DBA radiological consequences are listed calculation of these two components of the TEDE should in a "Summary and Conclusions" table at the consider all radionuclides, including progeny from the end of the discussion of each DBA in Section decay of parent radionuclides, that are significant with 4.

regard to dose consequences and the released radioactivity.

4.1.2 The exposure-to-CEDE factors for inhalation of Conforms Table 2.1 of Federal Guidance Report 11 radioactive material should be derived from the data were used. (Section 4.2.4.1) provided in ICRP Publication 30, "Limits for Intakes of Radionuclides by Workers" (Ref. 19). Table 2.1 of Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion" (Ref. 20), provides tables of conversion 6 of 37

NOC-AE-07002127 Attachment 6 Table A: Conformance with Regulatory Guide 1.183 Main Sections RG Section Regulatory Position Analysis Comments a factors acceptable to the NRC staff.

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 Conforms The standard breathing rates specified in RG should be assumed to be 3.5 x 10-4cubic meters per 1.183. (Table 4.2-1) 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 10-4 cubic meters per second. After that and until the end of the accident, the rate should be assumed to be 2.3 x 10.4 cubic meters per second.

4.1.4 The DDE should be calculated assuming submergence in Conforms The analyses were performed using the NRC semi-infinite cloud assumptions with appropriate credit RADTRAD computer code. (Section 4.2.1) for attenuation by body tissue. EDE may be used in lieu of DDE in determining the contribution of external dose to the TEDE.

4.1.5 The TEDE should be determined for the most limiting Conforms The TEDE was determined for the most person at the EAB. The maximum EAB TEDE for any limiting person at the EAB and the maximum two-hour period following the start of the radioactivity two-hour dose has been reported. (Section release should be determined and used in determining 4.2.1) 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.

4.1.6 TEDE should be determined for the most limiting Conforms The TEDE was determined for the most receptor at the outer boundary of the low population limiting person at the LPZ. (Section 4.2.1) 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 No plume depletion has been credited.

7 of 37

NOC-AE-07002127 Attachment 6 Table A: Conformance with Regulatory Guide 1.183 Main Sections RG Section Regulatory Position Analysis Comments a effluent plume by deposition on the ground. I (Section 4.2.1) 4.2 - Control Room Dose Consequences 4.2.1 The TEDE analysis should consider all sources of Conforms The DBA LOCA radiation dose to personnel radiation that will cause exposure to control room in the CR and TSC includes the gamma shine personnel. from the primary containment airborne activity (CR and TSC), from airborne activity in the electrical penetration area (CR only),

from activity in the radioactive cloud surrounding the plant structures (CR and TSC), and from trapped activity on filters (CR and TSC). (Section 4.3.4.2) 4.2.2 The radioactive material releases and radiation levels Conforms Gamma shine dose contribution to the control used in the control room dose analysis should be room is discussed in Section 4.3.4.2. Source determined using the same source term, transport, and to receptor models are discussed in Section release assumptions used for determining the EAB and 4.1.3.

the LPZ TEDE values, unless these assumptions would result in non-conservative results for the control room.

4.2.3 The models used to transport radioactive material into Conforms Gamma shine dose contribution to the control and through the control room, and the shielding models room is discussed in Section 4.3.4.2.

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 Conforms For the DBA LOCA, no credit is taken for airborne radioactive material within the control room any filtration other than for the recirculation may be assumed. Such features may include control filters for the CR. The recirculation filter room isolation or pressurization, or intake or features are qualified and acceptable per the recirculation filtration. Refer to Section 6.5.1, "ESF referenced guidance. (Section 4.2.2)

Atmospheric Cleanup System," of the SRP (Ref. 3) and Regulatory Guide 1.52, "Design, Testing, and For the FHA, no credit is taken for any Maintenance Criteria for Postaccident Engineered- I filtration (make-up or recirculation clean-up) 8 of 37

NOC-AE-07002127 Attachment 6 Table A: Conformance with Regulatory Guide 1.183 Main Sections RG Section Regulatory Position Analysis Comments a Safety-Feature Atmosphere Cleanup System Air for either the CR or the TSC. (Section 4.2.2 Filtration and Adsorption Units of Light-Water-Cooled and 4.2.3)

Nuclear Power Plants" (Ref. 25), for guidance.

4.2.5 Credit should generally not be taken for the use of Conforms No credit is taken for the use of personal personal protective equipment or prophylactic drugs. protective equipment or prophylactic drugs.

(Section 4.2.2 and 4.2.3) 4.2.6 The dose receptor for these analyses is the hypothetical Conforms The standard breathing rates specified in RG maximum exposed individual who is present in the 1.183 and the standard CR and TSC control room for 100% of the time during the first 24 occupancy factors specified in RG 1.183 have hours after the event, 60% of the time between 1 and 4 been used. (Table 4.2-3) 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 10- cubic meters per second.

4.2.7 Control room doses should be calculated using dose Conforms See response to Position 4.1.2.

conversion factors identified in Regulatory Position 4.1 above for use in offsite dose analyses.

4.3 - Other The guidance provided in Regulatory Positions 4.1 and Technical Support Center (TSC) doses were Dose 4.2 should be used, as applicable, in re-assessing the calculated for the analyzed accidents and the Consequences radiological analyses identified in Regulatory Position results are found in the "Summary and 1.3.1, such as those in NUREG-0737. Design envelope Conclusions" section of each accident source terms provided in NUREG-0737 should be discussion in Section 4.

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 The radiation doses used for the current be determined using the guidance of Appendix I of this licensing basis environmental qualification guide. analyses were calculated using source terms determined by TID- 14844 methodology.

9 of 37

NOC-AE-07002127 Attachment 6 Table A: Conformance with Regulatory Guide 1.183 Main Sections RG Section Regulatory Position Analysis Comments a AST impact on these doses has been considered. (Section 1.0) 4.4- The radiological criteria for the EAB, the outer boundary All accident analyses consequences meet 10 Acceptance of the LPZ, and for the control room are in 10 CFR CFR 50.67 and the SRP. See the "Summary Criteria 50.67. These criteria are stated for evaluating reactor and Conclusions" section of each accident accidents of exceedingly low probability of occurrence discussion in Section 4.

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 The accident analyses consequences are well occurrence, postulated EAB and LPZ doses should not within the criteria tabulated in Table 6. See exceed the criteria tabulated in Table 6. the "Summary and Conclusions" section of each accident discussion in Section 4.

The acceptance criteria for the various NUREG-0737 For post-accident vital area access, the results (Ref. 2) items generally reference General Design of the assessment of dose impact of Criteria 19 (GDC 19) from Appendix A to 10 CFR Part containment shine demonstrates that the 50 or specify criteria derived from GDC-19. These current calculated doses (based on TID-14844 criteria are generally specified in terms of whole body source terms) bound the corresponding doses dose, or its equivalent to any body organ. For facilities that would be calculated based on the AST.

applying for, or having received, approval for the use of (Section 4.9) an AST, the applicable criteria should be updated for consistency with the TEDE criterion in 10 CFR The post-accident containment high range 50.67(b)(2)(iii). radiation monitors are determined not to be impacted by the AST. (Section 4.9)

The CR radiological dose impact of AST is specifically calculated for the six Design Basis Accidents. See the "Summary and Conclusions" section of each accident 10 of 37

NOC-AE-07002127 Attachment 6 Table A: Conformance with Regulatory Guide 1.183 Main Sections RG Section Regulatory Position Analysis Comments a

_discussion I_ _ in Section 4.

5.1 - General Considerations for Analysis ssumptions and Methodology '

5.1.1 The evaluations required by 10 CFR 50.67 are re- Conforms Appendix B, "Quality Assurance Criteria for analyses of the design basis safety analyses and Nuclear Power Plants and Fuel Reprocessing evaluations required by 10 CFR 50.34; they are Plants," to 10 CFR Part 50 was applied.

considered to be a significant input to the evaluations (Section 4.0) 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.

These design basis analyses were structured to provide a Care has been taken to ensure that dose conservative set of assumptions to test the performance analyses have not been "tuned" to a specific of one or more aspects of the facility design. Licensees set of accident progression assumptions.

should exercise caution in proposing deviations based (Section 4.0) upon data from a specific accident sequence since the DBAs were never intended to represent any specific accident sequence -- the proposed deviation may not be conservative for other accident sequences.

5.1.2 Credit may be taken for accident mitigation features that For the DBA LOCA, credit is taken for the are classified as safety-related, are required to be CR recirculation filters. These are safety-operable by technical specifications, are powered by related and are required to be operable by TS emergency power sources, and are either automatically 3.7.7. They are powered by emergency actuated or, in limited cases, have actuation power sources and are automatically actuated requirements explicitly addressed in emergency by SI signal. TS 3.3.2 applies. (Section 4.2.2) operating procedures.

Without credit being taken for the FHB filters The single active component failure that results in or for the CR make-up filters (and the the most limiting radiological consequences should be I associated heaters to control intake humidity),

11 of 37

NOC-AE-07002127 Attachment 6 Table A: Conformance with Regulatory Guide 1.183 Main Sections RG Section Regulatory Position Analysis Comments a assumed. the single-failure assessment becomes much simpler for application of the AST than that of the CLB. For the AST DBA LOCA, an electrical division electrical failure is assumed as a single failure to minimize containment mixing via the containment fan coolers. This assumption maximizes dose. Only two out of three trains of containment ventilation are assumed to operate, and one fan-cooler on one of the operating trains is assumed to be out of service, as well. The spray removal lambdas used are also consistent with the loss of one spray train, as are the assumptions regarding CR ventilation and filtration.

(Section 4.3.4.1)

Assumptions regarding the occurrence and timing of a The LOCA, MSLB, SGTR, CREA and LRA loss of offsite power should be selected with the analyses assume a loss of offsite power objective of maximizing the postulated radiological concurrent with the accident.

consequences.

5.1.3 The numeric values that are chosen as inputs to the Conforms Conservative parameters were used when analyses required by 10 CFR 50.67 should be selected calculating components in the dose analyses.

with the objective of determining a conservative (Section 4.0) postulated dose. In some instances, a particular parameter may be conservative in one portion of an analysis but be non-conservative in another portion of the same analysis.

5.1.4 In order to issue a license amendment authorizing the Conforms The analysis assumptions and methods are use of an AST and the TEDE dose criteria, the NRC compatible with the ASTs and the TEDE staff must make a current finding of compliance with criteria.

regulations applicable to the amendment. The 12 of 37

NOC-AE-07002127 Attachment 6 Table A: Conformance with Regulatory Guide 1.183 Main Sections RG Section Regulatory Position Analysis Comments a characteristics of the ASTs and the revised dose calculational methodology may be incompatible with many of the analysis assumptions and methods currently reflected in the facility's design basis analyses.

Licensees should ensure that analysis assumptions and methods are compatible with the ASTs and the TEDE criteria.

5.2 - Licensees should analyze the DBAs that are affected by Conforms The postulated accident radiological Accident- the specific proposed applications of an AST. consequence analyses specified in this RG are Specific updated for AST implementation impact.

Assumptions The NRC staff has determined that the analysis Each assumption is addressed with for Analysis assumptions in the appendices to this guide provide an conformance with the RG accident analyses Assumptions integrated approach to performing the individual assumptions provided by this table.

and analyses and generally expects licensees to address each Methodology assumption or propose acceptable alternatives.

The NRC will consider licensee proposals for changes in No changes were made to analysis analysis assumptions based upon risk insights. The staff assumptions based upon risk insights.

will not approve proposals that would reduce the defense (Section 4.0) in depth deemed necessary to provide adequate protection for public health and safety. Defense-in-depth has not been compromised by the changes proposed in this application.

(Section 5.2) 5.3 - Atmospheric dispersion values (y/Q) for the EAB, the Conforms All XIQ values have been recalculated for the Meteorology LPZ, and the control room that were approved by the AST application. (Section 4.1)

Assumptions staff during initial facility licensing or in subsequent licensing proceedings may be used in performing the radiological analyses identified by this guide.

References 22 and 28 of this RG should be used if the Reference 28 of RG 1.183 has been used to I FSAR X/Q values are to be revised or if values are to be calculate offsite XIQ values (for the EAB, the 13 of 37

NOC-AE-07002127 Attachment 6 Table A: Conformance with Regulatory Guide 1.183 Main Sections RG Section Regulatory Position Analysis Comments a determined for new release points or receptor distances. LPZ). RG 1.194 guidance has been used, as well. Reference 22 of RG 1.183 has not been used. (Sections 4.1.2 and 4.1.3)

Fumigation should be considered where applicable for Fumigation has not been included since no the EAB and LPZ. For the EAB, the assumed fumigation credit is taken for an elevated release.

period should be timed to be included in the worst 2-hour exposure period.

The NRC computer code PAVAN implements The PAVAN code was used for determining Regulatory Guide 1.145 and its use is acceptable to the of offsite X/ Q values. (Section 4.1.2)

NRC staff.

The methodology of the NRC computer code ARCON96 ARCON96 was used for determining X/ Q is generally acceptable to the NRC staff for use in values for onsite receptors near building determining control room XIQ values. structures. (Section 4.1.3)

Meteorological data collected in accordance with the Recently acquired meteorological data site-specific meteorological measurements program (five years from 2000 to 2004) is used to described in the facility FSAR should be used in calculate onsite and offsite atmospheric generating accident X/ Q values. Additional guidance is dispersion. (Section 4.1) provided in Regulatory Guide 1.23.

All changes in X/ Q analysis methodology should be ,/ Q values for radiological dose calculations reviewed by the NRC staff. are found in Tables 4.1-24 and 4.1-37.

5.6 - The assumptions in Appendix I to this guide are Conforms The radiation doses used for the CLB Assumptions acceptable to the NRC staff for performing environmental qualification analyses were for Evaluating radiological assessments associated with equipment calculated using source terms determined by the Radiation qualification. The assumptions in Appendix I TID-14844 methodology. (Section 1.0)

Doses for will supersede Regulatory Positions 2.c(1) and 2.c(2)

Equipment and Appendix D of Revision 1 of Regulatory 14 of 37

NOC-AE-07002127 Attachment 6 Table A: Conformance with Regulatory Guide 1.183 Main Sections RG Section Regulatory Position Analysis Comments a Qualification Guide 1.89, for operating reactors that have amended their licensing basis to use an alternative source term.

Except as stated in Appendix I, all other assumptions, methods, and provisions of Revision 1 of Regulatory Guide 1.89 remain effective.

The NRC staff is assessing the effect of increased cesium releases on EQ doses to determine whether licensee action is warranted. Until such time as this generic issue is resolved, licensees may use either the AST or the TID 14844 assumptions for performing the required EQ analyses. However, no plant modifications are required to address the impact of the difference in source term characteristics (i.e., AST vs TID14844) on EQ doses pending the outcome of the evaluation of the generic issue.

Table B - Comparison with Regulatory Guide 1.183 Appendix A (PWR Loss-of-Coolant Accident)

RG Section Regulatory Position Analysis Comments a 1 - Source Acceptable assumptions regarding core inventory and Conforms The core inventory and release of radio-Term the release of radionuclides from the fuel are provided in nuclides in the AST analysis were derived Assumptions Regulatory Position 3 of this guide. using the guidance outlined in this RG.

ORIGIN 2.1 code was used to calculate plant-specific fission product inventories. (Section 4.3.2)

The release fractions are provided in Table 4.3-10.

2 - Source If the sump or suppression pool pH is controlled at Sump pH is A calculation was performed to evaluate Term values of 7 or greater, the chemical form of radioiodine less than 7. containment sump pH in the event of a DBA Assumptions released to the containment should be assumed to be The plant- LOCA. The objective of the analysis was to 15 of 37

NOC-AE-07002127 Attachment 6 Table B - Comparison with Regulatory Guide 1.183 Appendix A (PWR Loss-of-Coolant Accident)

RG Section Regulatory Position Analysis Comments a 95% cesium iodide (CsI), 4.85 percent elemental iodine, specific determine the transient containment sump pH and 0.15 percent organic iodide. Iodine species, calculation so that the removal of elemental and including those from iodine re-evolution, for sump or will be particulate iodine (cesium iodide - CsI) from suppression pool pH values less than 7 will be evaluated provided, the containment atmosphere in the course of on a case-by-case basis. Evaluations of pH should the DBA LOCA would not be overstated.

consider the effect of acids and bases created during the The analysis credits the pH buffering effect of LOCA event, e.g., radiolysis products. With the trisodium phosphate (TSP) stored in the exception of elemental and organic iodine and noble containment sump. The pH decreases slightly gases, fission products should be assumed to be in below 7.0 over the 30-day duration of the particulate form. radiological consequence analysis for the DBA-LOCA, and the impact of that decrease has been reflected in the CR, TSC, and offsite doses. Because the pH of the containment sump falls below 7.0 after one day, a fractional iodine release for ESF leakage greater than 10% was considered per RG 1.183. (Sections 4.3.3.1.1) 3 - Assumptions on Transport in Primary Containment 3.1 The radioactivity released from the fuel should be Conforms The radioactivity release from the fuel is assumed to mix instantaneously and homogeneously assumed to mix instantaneously and throughout the free air volume of the primary homogeneously throughout the containment containment in PWRs as it is released. This distribution air space as it is released. (Section 4.3.5) should be adjusted if there are internal compartments that have limited ventilation exchange.

3.2 Reduction in airborne radioactivity in the containment Conforms The natural removal rate for elemental iodine by natural deposition within the containment may be is 4.5 per hour (Section 4.3.5).

credited. The prior practice of deterministically assuming that a 50% plate-out of iodine is released from the fuel is no longer acceptable to the NRC staff as it is inconsistent with the characteristics of the revised 16 of 37

NOC-AE-07002127 Attachment 6 Table B - Comnarison with Reeulatorv Guide 1.183 A pendix A (PWR Loss-of-Coolant Accident)

RG Section Regulatory Position Analysis Comments a I source terms. I I 3.3 Reduction in airborne radioactivity in the containment Conforms The containment spray systems are designed by containment spray systems that have been designed and maintained in accordance with Chapter and are maintained in accordance with Chapter 6.5.2 of 6.5.2 of the SRP.

the SRP may be credited.

The evaluation of the containment sprays should address Values of reduction in airborne radioactivity areas within the primary containment that are not in the containment by containment spray covered by the spray drops. The mixing rate attributed to systems is given in Table 4.3-13 and natural convection between sprayed and unsprayed discussed in Section 4.3.5. Forced mixing regions of the containment building, provided that crediting containment fan-cooler units is used.

adequate flow exists between these regions, is assumed to be two turnovers of the un-sprayed regions per hour, unless other rates are justified. The containment building atmosphere may be considered a single, well-mixed volume if the spray covers at least 90% of the volume and if adequate mixing of unsprayed compartments can be shown.

The SRP sets forth a maximum decontamination factor The AST spray removal parameters are given (DF) for elemental iodine based on the maximum iodine in Table 4.3-13. The table demonstrates that activity in the primary containment atmosphere when the the particulate iodine removal rate is reduced sprays actuate, divided by the activity of iodine from 6.9 to 0.7 when a DF of 50 is reached.

remaining at some time after decontamination.

The SRP also states that the particulate iodine removal rate should be reduced by a factor of 10 when a DF of 50 is reached. The reduction in the removal rate is not required if the removal rate is based on the calculated time-dependent airborne aerosol mass. There is no specified maximum DF for aerosol removal by sprays.

3.7 1 The primary containment should be assumed to leak at Conforms I The volumetric leak rate from containment is 17 of 37

NOC-AE-07002127 Attachment 6 Table B - Comnarison with Reeulatorv Guide 1.183 A pendix A (PWR Loss-of-Coolant Accident)

RG Section Regulatory Position Analysis Comments a the peak pressure technical specification leak rate for the assumed to be 0.3%/day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 and 0.15%/day for the remainder of the 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 accident - this is consistent with the CLB.

specification leak rate. (Section 4.3.3.1) Maximum post-LOCA containment temperature and pressure were assumed. (Section 4.3-11) 3.8 If the primary containment is routinely purged during Bounding Primary containment is routinely purged power operations, releases via the purge system prior to during power operations. Releases via the containment isolation should be analyzed and the purge system prior to containment isolation resulting doses summed with the postulated doses from are analyzed and the resulting doses are other release paths. The purge release evaluation summed with the postulated doses from other should assume that 100% of the radionuclide inventory release paths.

in the reactor coolant system liquid is released to the containment at the initiation of the LOCA. This The reactor coolant concentrations are based inventory should be based on the technical specification on 1% failed fuel that is greater than the TS reactor coolant system equilibrium activity. Iodine limit of 1.0 ltCi/gm.

spikes need not be considered. If the purge system is not isolated before the onset of the gap release phase, the Containment leakage is assumed via open release fractions associated with the gap release and purge lines for the first 23 seconds of the early in-vessel phases should be considered as accident. This leakage is released to the applicable. environment via the plant vent.

(Section 4.3.3.1.1) 4- Not Assumptions Applicable of Dual Containment 5- ESF systems that re-circulate sump water outside of the Conforms The radiological consequences from the Assumptions primary containment are assumed to leak during their postulated ESF systems leakage is analyzed on ESF intended operation. The radiological consequences from and combined with consequences postulated System the postulated leakage should be analyzed and combined for other fission product release paths.

18 of 37

NOC-AE-07002127 Attachment 6 Table B - Comparison with Reiulatorv Guide 1.183 A pendix A (PWR Loss-of-Coolant Accident)

RG Section Regulatory Position Analysis Comments a Leakage with consequences postulated for other fission product (Section 4.3.3.2) release paths to determine the total calculated radiological consequences from the LOCA.

5.1 With the exception of noble gases, all the fission Conforms For determination of the dose contribution products released from the fuel to the containment from ESF leakage, all radionuclides assumed should be assumed to instantaneously and to be released from the core (except noble homogeneously mix in the primary containment sump gases) are assumed to be instantaneously and water at the time of release from the core. In lieu of this homogeneously mixed in the containment deterministic approach, suitably conservative sump. (Section 4.3.3.2) mechanistic models for the transport of airborne activity in containment to the sump water may be used.

5.2 The leakage should be taken as two times the sum of the Conforms ESF leak rate is 4140 cc/hr and analyzed as simultaneous leakage from all components in the ESF 8280 cc/hr per the CLB.

recirculation systems above which the technical specifications, or licensee commitments to item III.D. 1.1 ESF leakage is assumed to begin at time = 0.

of NUREG-0737 would require declaring such systems inoperable. The leakage should be assumed Leakage past valves and into tanks vented to to start at the earliest time the recirculation flow occurs the atmosphere has been evaluated to have no in these systems and end at the latest time the releases impact on the dose consequences because of from these systems are terminated. Consideration should the distances involved and the time of travel.

also be given to design leakage through valves isolating ESF recirculation systems from tanks vented to (Section 4.3.3.2) atmosphere, e.g., emergency core cooling system (ECCS) pump mini-flow return to the refueling water storage tank.

5.3 With the exception of iodine, all radioactive materials in Conforms Complies. (Section 4.3.3.2) the re-circulating liquid should be assumed to be retained in the liquid phase.

5.5 If the temperature of the leakage is less than 212'F or Higher The fractional release of iodine from ESF the calculated flash fraction is less than 10%, the amount amount sump water leakage uses a value of 10% only of iodine that becomes airborne should be assumed to be assumed for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Beyond that time, the 19 of 37

NOC-AE-07002127 Attachment 6 Table B - Comparison with Regulatory Guide 1.183 Appendix A (PWR Loss-of-Coolant Accident)

RG Section Regulatory Position Analysis Comments a 10% of the total iodine activity in the leaked fluid, because release fraction is increased to 16% (24-480 unless a smaller amount can be justified based on the sump pH hours) and then to 25% (480 hours0.00556 days <br />0.133 hours <br />7.936508e-4 weeks <br />1.8264e-4 months <br /> to 720 actual sump pH history and area ventilation rates. goes below hours) based on the calculated volatility of the 7.0 iodine for the pH values over those intervals relative to the volatility for a pH of 7.0.

(Section 4.3.3.2) 5.6 The radioiodine that is postulated to be available for Conforms The radioiodine that is postulated to be release to the environment is assumed to be 97% available for release to the environment is elemental and 3% organic. Reduction in release activity assumed to be 97% elemental and 3%

by dilution or holdup within buildings, or by ESF organic. The ESF leakage is assumed to ventilation filtration systems, may be credited where release directly into the FHB from the leaked applicable, reactor coolant. It is then assumed to be released instantaneously into the environment without benefit of filtration via the plant vent.

(Table 4.3-11 & Section 4.3.3.2) 6- Not Assumption Applicable on Main Steam Isolation Valve Leakage in BWRs 7- The radiological consequences from post-LOCA Not Hydrogen control by purge is not part of the Assumption primary containment purging as a combustible gas or Applicable licensing basis.

on pressure control measure should be analyzed. If the Containment installed containment purging capabilities are (Ref: Amendment 165 for Unit 1 and 155 for Purging maintained for purposes of severe accident management Unit 2 - TAC Nos. MC4229 and MC4290) and are not credited in any design basis analysis, radiological consequences need not be evaluated.

20 of 37

NOC-AE-07002127 Attachment 6 Table C - Comparison with Regulatory Guide 1.183 Appendix B (Fuel Handling Accident)

RG Section Regulatory Position Analysis Comments a 1 - Source Acceptable assumptions regarding core inventory and Conforms The assumptions regarding core inventory and Term the release of radionuclides from the fuel are provided in the release of radionuclides from the fuel are Regulatory Position 3 of this guide. consistent with Regulatory Position 3 of this guide. (Tables 4.4-1 and 4.4-2) 1.1 The number of fuel rods damaged during the accident Conforms The mechanical part of the analysis remains should be based on a conservative analysis that considers unchanged from the STPEGS CLB; the total the most limiting case., number of failed fuel rods is 314 (out of 50952 for an entire core). (Section 4.4.5) 1.2 The fission product release from the breached fuel is Conforms The fission product release from the breached based on Regulatory Position 3.2 of this guide and the fuel is based on Regulatory Position 3.2 of estimate of the number of fuel rods breached. All the gap this guide and the estimate of the number of activity in the damaged rods is assumed to be fuel rods breached. All the gap activity in the instantaneously released. Radionuclides that should damaged rods is assumed to be be considered include xenons, kryptons, halogens, instantaneously released. Since alkali metal cesiums, and rubidiums. releases (as particulates) are assumed to experience an infinite DF due to the water submergence (per RG 1.183), no alkali metals (e.g., Cs and Rb) are included. (Sections 4.4.4 and 4.4.5) 1.3 The chemical form of radioiodine released from the fuel Conforms This position is not explicitly used. Given the to the spent fuel pool should be assumed to be 95% 23 feet of water depth and the effective DF of cesium iodide (CsI), 4.85 percent elemental iodine, and 200, the iodine released to the atmosphere is 0.15 percent organic iodide. The CsI released from the 57% elemental and 43% organic, as discussed fuel is assumed to completely dissociate in the pool in Position 2 of Appendix B of RG 1.183.

water. Because of the low pH of the pool water, the (Section 4.4.4) iodine re-evolves as elemental iodine. This is assumed to occur instantaneously.

2 - Water If the depth of water above the damaged fuel is 23 feet Conforms The depth of water over the damaged fuel is Depth or greater, the decontamination factors for the elemental not less than 23 feet.

21 of 37

NOC-AE-07002127 Attachment 6 Table C - Comparison with Regulatory Guide 1.183 Appendix B (Fuel Handling Accident)

RG Section Regulatory Position Analysis Comments a and organic species are 500 and 1, respectively, giving an overall effective decontamination factor of 200 (i.e., Due to the submergence of the damaged fuel, 99.5% of the total iodine released from the damaged the iodine release is assumed to experience a rods is retained by the water). This difference in DF of 200. The assumed iodine chemical decontamination factors for elemental(99.85%) and form after decontamination by the water pool organic iodine (0.15%) species results in the iodine is 43% organic and 57% elemental.

above the water being composed of 57% elemental and 43% organic species. (Section 4.4.4) 3 - Noble The retention of noble gases in the water in the fuel pool Conforms No DF is applied to the noble gas. The DF Gases or reactor cavity is negligible for particulate is assumed to be infinite.

(i.e., decontamination factor of 1). Particulate (Section 4.4.4) radionuclides are assumed to be retained by the water in the fuel pool or reactor cavity (i.e., infinite decontamination factor).

4 - FHA Within the Fuel Building 4.1 The radioactive material that escapes from the fuel pool Assumption Following accident initiation at t = 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> to the fuel building is assumed to be released to the more after shutdown, the radionuclide inventory environment over a 2-hour time period. conservative from the damaged fuel pins is assumed to leak out to the environment instantaneously.

(Sections 4.4.1 and 4.4.2) 4.2 A reduction in the amount of radioactive material Not No credit is taken for filtration by the FHB released from the fuel pool by engineered safety feature Applicable filters or for hold-up in the FHB. (Section (ESF) filter systems may be taken into account provided 4.4.2) these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02.

4.3 The radioactivity release from the fuel pool should be Not Following accident initiation at t = 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> assumed to be drawn into the ESF filtration system Applicable after shutdown, the radionuclide inventory without mixing or dilution in the fuel building. from the damaged fuel pins is assumed to leak I I_out to the environment instantaneously.

22 of 37

NOC-AE-07002127 Attachment 6 Table C - Comparison with Regulatory Guide 1.183 Appendix B (Fuel Handling Accident)

RG Section Regulatory Position Analysis I Comments a (Sections 4.4.1 and 4.4.2) 5 - FHA Within Containment 5.3 If the containment is open during fuel handling Assumption Following accident initiation at t = 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> operations (e.g., personnel air lock or equipment hatch is more after shutdown, the radionuclide inventory open), the radioactive material that escapes from the conservative from the damaged fuel pins is assumed to leak reactor cavity pool to the containment is released to the out to the environment instantaneously.

environment over a 2-hour time period. (Sections 4.4.1 and 4.4.2) 5.4 A reduction in the amount of radioactive material Not Following accident initiation at t = 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> released from the containment by ESF filter systems Applicable after shutdown, the radionuclide inventory may be taken into account provided that these systems from the damaged fuel pins is assumed to leak meet the guidance of Regulatory Guide 1.52 and Generic out to the environment instantaneously.

Letter 99-02. There is no credit taken for activity removal other than by scrubbing by the water in the refueling cavity. (Sections 4.4.1 and 4.4.2) 5.5 Credit for dilution or mixing of the activity released Not Following accident initiation at t = 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> from the reactor cavity by natural or forced convection Applicable after shutdown, the radionuclide inventory inside the containment may be considered on a case-by- from the damaged fuel pins is assumed to leak case basis. out to the environment instantaneously.

(Section 4.4.2)

Table D: Conformance with Regulatory Guide 1.183 Appendix E (PWR Main Steam Line Break)

RG Section Regulatory Position Analysis Comments a I Assumptions acceptable to the NRC staff regarding core Conforms Assumptions regarding core inventory and the inventory and the release of radionuclides from the fuel release of radionuclides from the fuel are are provided in Regulatory Position 3 of this regulatory consistent with Regulatory Position 3 of this guide. The release from the breached fuel is based on regulatory guide. No fuel damage is Regulatory Position 3.2 of this guide and the estimate postulated to occur for the Main Steam Line 23 of 37

NOC-AE-07002127 Attachment 6 Table D: Conformance with Regulatory Guide 1.183 Appendix E (PWR Main Steam Line Break)

RG Section Regulatory Position Analysis Comments a of the number of fuel rods breached. The fuel damage Break. (Section 4.5.3) estimate should assume that the highest worth control rod is stuck at its fully withdrawn position.

2 If no or minimal 2 fuel damage is postulated for the Conforms The activity assumed in the analysis is based limiting event, the activity released should be the on the activity associated with the maximum maximum coolant activity allowed by the technical technical specification values. In determining specifications. Two cases of iodine spiking should be dose equivalent 1-131, only the radioiodine assumed. associated with normal operations or iodine spikes is included. (Section 4.5.5)

Footnote 2: The activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximum technical specification values, whichever maximizes the radiological consequences. In determining dose equivalent 1-131 (DE 1-131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.

2.1 Case 1 A reactor transient has occurred prior to the Conforms This analyzed case involves a 60 ptCi/gm pre-postulated main steam line break (MSLB) and has raised accident iodine spike, consistent with the the primary coolant iodine concentration to the Technical Specification operational Reactor maximum value (typically 60 ptCi/gm DE 1-13 1) Coolant System activity concentration limit permitted by the technical specifications (i.e., a for assumed spike. All of the spike activity is preaccident iodine spike case). homogeneously mixed in the primary coolant, prior to accident initiation. (Section 4.5.1) 2.2 Case 2 The primary system transient associated with the Conforms This case involves an accident initiated iodine MSLB causes an iodine spike in the primary system. The spike that occurs concurrently with the release increase in primary coolant iodine concentration is of the fluid from the primary and secondary estimated using a spiking model that assumes that the coolant systems. This transient is associated iodine release rate from the fuel rods to the primary with an iodine spike which assumes that the 24 of 37

NOC-AE-07002127 Attachment 6 Table D: Conformance with Regulatory Guide 1.183 Appendix E (PWR Main Steam Line Break)

RG Section Regulatory Position Analysis Comments a coolant (expressed in curies per unit time) increases to a iodine release rate from the fuel rods to the value 500 times greater than the release rate primary coolant increases to a value 500 times corresponding to the iodine concentration at the greater than the release rate corresponding to equilibrium value (typically 1.0 jtCi/gm DE 1-131) the 1.0 ptCi/gm DE 1-131 RCS equilibrium specified in technical specifications (i.e., concurrent iodine. The elemental and particulate iodines iodine spike case). A concurrent iodine spike need not be release rate spike is assumed to occur for considered if fuel damage is postulated. The assumed eight hours. (Section 4.5.1) iodine spike duration should be 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Shorter spike durations may be considered on a case-by-case basis if it can be shown that the activity released by the 8-hour spike exceeds that available for release from the fuel gap of all fuel pins.

3 The activity released from the fuel should be assumed to Conforms The activity released from the fuel is assumed be released instantaneously and homogeneously through to be released instantaneously and the primary coolant. homogeneously through the primary coolant.

(Section 4.5.5) 4 The chemical form of radioiodine released from the fuel Bounds STP has taken a more conservative approach should be assumed to be 95% cesium iodide (CsI), 4.85 and assumes 4.2% elemental iodine, 13.1%

percent elemental iodine, and 0.15 percent organic organic iodine and 82.7% particulate released iodide. Iodine releases from the steam generators to the from the steam generators to the environment.

environment should be assumed to be 97% elemental (Section 4.5.4) and 3% organic. These fractions apply to iodine released as a result of fuel damage and to iodine released during normal operations, including iodine spiking.

5.1 For facilities that have not implemented alternative Conforms The primary-to-secondary leak rate in the repair criteria (see Ref. E-1, DG- 1074), the primary-to- steam generators is based on one secondary leak rate in the steam generators should be gallon/minute (gpm)/1440 gallon/day (gpd) assumed to be the leak rate limiting condition for total leakage. The leak rate is apportioned operation specified in the technical specifications. between the faulted steam generator as 0.35 25 of 37

NOC-AE-07002127 Attachment 6 Table D: Conformance with Regulatory Guide 1.183 Appendix E (PWR Main Steam Line Break)

RG Section Regulatory Position Analysis Comments a For facilities with traditional generator specifications gpm/504 gpd and 0.65 gpm/936 gpd from the (both per generator and total of all generators), the intact steam generators. This is the leakage leakage should be apportioned between affected and assumption in the current accident analysis.

unaffected steam generators in such a manner that the This assumption is conservative when calculated dose is maximized. compared with the Technical Specification limit of 0.1 gpm/150 gpd limit. (Sections 4.5.4 and 4.5.5) 5.2 The density used in converting volumetric leak rates Conforms The density is assumed to be 8.33 lbs/gal.

(e.g., gpm) to mass leak rates (e.g., lbm/hr) should be (Section 4.5.5) consistent with the basis of the parameter being converted. The ARC leak rate correlations are generally based on the collection of cooled liquid. Surveillance tests and facility instrumentation used to show compliance with leak rate technical specifications are typically based on cooled liquid. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbM/ft3).

5.3 The primary-to-secondary leakage should be assumed to Conforms The steaming release and primary-to-continue until the primary system pressure is less than secondary coolant leakage is postulated to end the secondary system pressure, or until the temperature at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, when the primary and secondary of the leakage is less than 100°C (212'F). The release of loops have reached equilibrium. This radioactivity from unaffected steam generators should be consistent with the current licensing basis.

assumed to continue until shutdown cooling is in (Section 4.5.4) operation and releases from the steam generators have been terminated.

5.4 All noble gas radionuclides released from the primary Conforms Noble gases are released without reduction or system are assumed to be released to the environment mitigation. (Section 4.5.4) without reduction or mitigation.

5.5 The transport model described in Appendix E of RG Conforms The transport model described is utilized for 26 of 37

NOC-AE-07002127 Attachment 6 Table D: Conformance with Regulatory Guide 1.183 Appendix E (PWR Main Steam Line Break)

RG Section Regulatory Position Analysis Comments a 1.183 should be utilized for iodine and particulate iodine and particulate releases from the steam releases from the steam generators. This model is shown generators. (Section 4.5.1.1) in Figure E-1 of RG 1.183.

5.5.1 A portion of the primary-to-secondary leakage will flash Conforms Primary-to-secondary coolant leakage through to vapor, based on the thermodynamic conditions in the the faulted steam generator conservatively reactor and secondary coolant. goes directly to the environment without

" During periods of steam generator dryout, all of mixing with the secondary coolant.

the primary-to-secondary leakage is assumed to Therefore, under the assumed dry-out flash to vapor and be released to the environment conditions, no partitioning of any nuclides is with no mitigation. postulated to occur in this release pathway.

" With regard to the unaffected steam generators (Section 4.5.4) used for plant cooldown, the primary-to-secondary leakage can be assumed to mix with For all post-accident releases via the intact the secondary water without flashing during steam generator loops, the mechanism for periods of total tube submergence. release to the environment is steaming of the secondary coolant. Because of this release dynamic, a reduction is taken in the amount of activity released to the environment based on partitioning of nuclides between the liquid and gas states of water. For iodine, the partitioning factor of 100 is used, per RG 1.183. (Sections 4.5.4 and 4.5.5) 5.5.2 The leakage that immediately flashes to vapor will rise Conforms A partition factor of 100 is used for elemental through the bulk water of the steam generator and enter and particulate iodines from primary-to-the steam space. Credit may be taken for scrubbing secondary leakage in the intact steam in the generator, using the models in NUREG-0409, generators. Noble gases and organic iodines "Iodine Behavior in a PWR Cooling System Following a are released with no partitioning. (Sections Postulated Steam Generator Tube Rupture Accident" 4.5.4 and 4.5.5)

, during periods of total submergence of the tubes.

27 of 37

NOC-AE-07002127 Attachment 6 Table D: Conformance with Regulatory Guide 1.183 Appendix E (PWR Main Steam Line Break)

RG Section Regulatory Position Analysis Comments a 5.5.3 The leakage that does not immediately flash is assumed Conforms See comments for Section 5.5.1 above.

to mix with the bulk water.

5.5.4 The radioactivity in the bulk water is assumed to become Conforms A partition coefficient of 100 for iodine is vapor at a rate that is the function of the steaming rate assumed in the analysis. (Section 4.5.5) and the partition coefficient. A partition coefficient for iodine of 100 may be assumed. The retention of particulate radionuclides in the steam generators is limited by the moisture carryover from the steam generators.

5.6 Operating experience and analyses have shown that for Conforms Tube uncovery does not occur in the intact some steam generator designs, tube uncovery may occur steam generators. (Section 4.5.5) for a short period following any reactor trip. The potential impact of tube uncovery on the transport model parameters (e.g., flash fraction, scrubbing credit) needs to be considered. The impact of emergency operating procedure restoration strategies on steam generator water levels should be evaluated.

Table E: Conformance with Regulatory Guide 1.183 Appendix F (PWR Steam Generator Tube Rupture)

RG Section Regulatory Position Analysis Comments a 1 Assumptions acceptable to the NRC staff regarding core Conforms Assumptions regarding core inventory and the inventory and the release of radionuclides from the fuel -release of radio-nuclides from the fuel are are in Regulatory Position 3 of this guide. The release consistent with Regulatory Position 3 of this from the breached fuel is based on Regulatory Position guide. No fuel damage is assumed due to the 3.2 of this guide and the estimate of the number of SGTR accident. (Section 4.6.3) fuel rods breached.

2 If no or minimal 2 fuel damage is postulated for the Conforms For this analysis, only the iodine and noble 28 of 37

NOC-AE-07002127 Attachment 6 Table E: Conformance with Regulatory Guide 1.183 Appendix F (PWR Steam Generator Tube Rupture)

RG Section Regulatory Position Analysis Comments a limiting event, the activity released should be the gas activities, which are conservatively maximum coolant activity allowed by technical characterized by operation with 1% core fuel specification. Two cases of iodine spiking should be defects and the equilibrium and spiked release assumed. rates from that fuel, define the source terms.

The AST release fractions associated with Footnote 2. The activity assumed in the analysis should iodines and noble gases are assumed to be be based on the activity associated with the projected 100%, and are released to the reactor coolant.

fuel damage or the maximum technical specification values, whichever maximizes the radiological No additional fuel damage is assumed due to consequences. In determining dose equivalent 1-131 (DE this accident. Two different cases of iodine 1-13 1), only the radioiodine associated with normal spiking are analyzed, in accordance with operations or iodine spikes should be included. Activity regulatory guidance as previously described.

from projected fuel damage should not be included. (Section 4.6.3) 2.1 Case 1 A reactor transient has occurred prior to the Conforms This analyzed case involves a 60 ýiCi/gm pre-postulated steam generator tube rupture (SGTR) and has accident iodine spike, consistent with the raised the primary coolant iodine concentration to the Technical Specification operational Reactor maximum value (typically 60 ptCi/gm DE 1-13 1) Coolant System activity concentration limit permitted by the technical specifications (i.e., a for assumed spike. All of the spike activity is preaccident iodine spike case). homogeneously mixed in the primary coolant, prior to accident initiation. (Section 4.6.1) 2.2 Case 2 The primary system transient associated with Conforms This case involves an accident initiated iodine the SGTR causes an iodine spike in the primary system. spike that occurs concurrently with the release The increase in primary coolant iodine concentration is of the fluid from the primary and secondary estimated using a spiking model that assumes that the coolant systems. This spike results in a iodine release rate from the fuel rods to the primary release rate that is 335 times greater than the coolant (expressed in curies per unit time) increases to a release rate corresponding to the 1.0 ýtCi/gm value 335 times greater than the release rate DE 1-131 RCS equilibrium iodine corresponding to the iodine concentration at the concentration, and lasts for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. (Section equilibrium value (typically 1.0 ptCi/gm DE 1-131) 4.6.1) 29 of 37

NOC-AE-07002127 Attachment 6 Table E: Conformance with Regulatory Guide 1.183 Appendix F (PWR Steam Generator Tube Rupture)

RG Section Regulatory Position Analysis Comments a specified in technical specifications (i.e., concurrent iodine spike case). A concurrent iodine spike need not be considered if fuel damage is postulated. The assumed iodine spike duration should be 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Shorter spike durations may be considered on a case-by-case basis if it can be shown that the activity released by the 8-hour spike exceeds that available for release from the fuel gap of all fuel pins.

3 The activity released from the fuel, if any, should be Conforms Mixing in the primary coolant is assumed to assumed to be released instantaneously and be released instantaneously and homogeneously through the primary coolant. homogeneously. (Section 4.6.5) 4 Iodine releases from the steam generators to the Bounds STP has taken a more conservative approach environment should be assumed to be 97% elemental and assumes 4.2% elemental iodine, 13.1%

and 3% organic. organic iodine and 82.7% particulate. (Section 4.6.4) 5.1 The primary-to-secondary leak rate in the steam Conforms The primary-to-secondary leak rate in the generators should be assumed to be the leak rate limiting steam generators is based on one condition for operation specified in the technical gallon/minute (gpm)/1440 gallon/day (gpd) specifications. The leakage should be apportioned total leakage. The leak rate is apportioned between affected and unaffected steam generators in between the ruptured steam generator as 0.35 such a manner that the calculated dose is maximized. gpm/504 gpd and 0.65 gpm/936 gpd from the intact steam generators. This is the leakage assumption in the current accident analysis.

This assumption is conservative when compared with the Technical Specification limit of 0.1 gpm/150 gpd limit. (Sections 4.6.4 and 4.6.5) 5.2 The density used in converting volumetric leak rates Conforms The density is assumed to be 8.33 lbs/gal.

(e.g., gpm) to mass leak rates (e.g., lbm/hr) should be (Section 4.6.5) 30 of 37

NOC-AE-07002127 Attachment 6 Table E: Conformance with Regulatory Guide 1.183 Appendix F (PWR Steam Generator Tube Rupture)

RG Section Regulatory Position Analysis Comments a consistent with the basis of surveillance tests used to show compliance with leak rate technical specifications.

These tests are typically based on cool liquid. Facility instrumentation used to determine leakage is typically located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft,).

5.3 The primary-to-secondary leakage should be assumed to Conforms Release of activity terminates after 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> continue until the primary system pressure is less than when shutdown cooling has been established.

the secondary system pressure, or until the temperature (Section 4.6.5) of the leakage is less than 100°C (2120 F). The release of radioactivity from the unaffected steam generators should be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated.

5.4 The release of fission products from the secondary Conforms A coincident loss of offsite power is assumed.

system should be evaluated with the assumption of a (Section 4.6.5) coincident loss of offsite power.

5.5 All noble gas radionuclides released from the primary Conforms Noble gases are released without reduction or system are assumed to be released to the environment mitigation. (Section 4.6.4) without reduction or mitigation.

5.6 The transport model described in Regulatory Positions Conforms The transport model described in Regulatory 5.5 and 5.6 of Appendix E of RG 1.183 should be Positions 5.5 and 5.6 of Appendix E of RG utilized for iodine and particulates. 1.183 is utilized for iodine and particulates.

(Section 4.6.1.1) 31 of 37

NOC-AE-07002127 Attachment 6 Table F: Conformance with Regulatory Guide 1.183 Appendix H (PWR Control Rod Ejection Accident)

RG Section Regulatory Position Analysis Comments a Assumptions acceptable to the NRC staff regarding core Conforms The core inventory in Regulatory Position 3 inventory are in Regulatory Position 3 of this guide. For of RG 1.183 is assumed. The CREA results in the rod ejection accident, the release from the breached damage to 10% of the core. 10% of the core fuel is based on the estimate of the number of fuel rods inventory of the noble gases and iodines is in breached and the assumption that 10% of the core the fuel gap and available for release. One inventory of the noble gases and iodines is in the fuel quarter percent of the core experiences fuel gap. The release attributed to fuel melting is based on melting. 100% of the noble gases, 25% of the the fraction of the fuel that reaches or exceeds the iodines, and 50% of the cesium and rubidium initiation temperature for fuel melting and the contained in the fraction of melted fuel are assumption that 100% of the noble gases and 25% of the available for release from containment. For iodines contained in that fraction are available for the secondary system release pathway, 100%

release from containment. For the secondary system of the noble gases and 50% of the iodines in release pathway, 100% of the noble gases and 50% of the fraction of melted fuel are released to the the iodines in that fraction are released to the reactor coolant. (Sections 4.7.2 and 4.7.3) reactor coolant.

2 If no fuel damage is postulated for the limiting event, a Not Since fuel damage is postulated, a radiological analysis is not required as the consequences Applicable radiological consequence analysis is of this event are bounded by the consequences projected performed.

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

3 Two release cases are to be considered. In the first, Conforms For the first case, the activity for leakage from 100% of the activity released from the fuel should be the containment assumes that the activity in assumed to be released instantaneously and the fuel pellet-clad gap and the activity homogeneously through the containment atmosphere. In released due to fuel melting is instantaneously the second, 100% of the activity released from the fuel mixed in the containment.

should be assumed to be completely dissolved in the primary coolant and available for release to the For the second case, 100% of the noble gases secondary system. and iodines released by fuel failed during the 32 of 37

NOC-AE-07002127 Attachment 6 Table F: Conformance with Regulatory Guide 1.183 Appendix H (PWR Control Rod Ejection Accident)

RG Section Regulatory Position Analysis Comments a accident is available for release to the secondary system.

(Section 4.7.2) 4 The chemical form of radioiodine released to the Conforms Iodines released to the containment (from the containment atmosphere should be assumed to be 95% fuel and RCS) are assumed to be 95%

cesium iodide (CsI), 4.85% elemental iodine, and 0.15% particulate, 4.85% elemental, and 0.15%

organic iodide. If containment sprays do not actuate or organic. (Section 4.7.5) are terminated prior to accumulating sump water, or if the containment sump pH is not controlled at values of 7 or greater, the iodine species should be evaluated on an individual case basis. Evaluations of pH should consider the effect of acids created during the rod ejection accident event, e.g., pyrolysis and radiolysis products.

With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form.

5 Iodine releases from the steam generators to the Bounds STP has taken a more conservative approach environment should be assumed to be 97% elemental and assumes 4.2% elemental iodine, 13.1%

and 3% organic. organic iodine and 82.7% particulate. (Section 4.7.5) 6.1 A reduction in the amount of radioactive material Conforms No credit is assumed for removal of iodine in available for leakage from the containment that is due to the containment due to containment sprays.

natural deposition, containment sprays, recirculating (Section 4.7.4) filter systems, dual containments, or other engineered safety features may be taken into account. Refer to Appendix A to this guide for guidance on acceptable methods and assumptions for evaluating these mechanisms.

33 of 37

NOC-AE-07002127 Attachment 6 Table F: Conformance with Regulatory Guide 1.183 Appendix H (PWR Control Rod Ejection Accident)

RG Section Regulatory Position Analysis Comments a 6.2 The containment should be assumed to leak at the leak Conforms The containment leaks for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at rate incorporated in the technical specifications at peak its design leak rate of 0.3 percent per day (i.e.,

accident pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at 50% of the Technical Specification limit). Thereafter, this leak rate for the remaining duration of the accident. the containment leak rate is 0.15 percent per Peak accident pressure is the maximum pressure defined day. (Section 4.7.5) in the technical specifications for containment leak testing. Leakage from subatmospheric containments is assumed to be terminated when the containment is brought to a subatmospheric condition as defined in technical specifications.

7.1 A leak rate equivalent to the primary-to-secondary leak Conforms For the case of the secondary release rate limiting condition for operation specified in the pathway, the assumed primary-to-secondary technical specifications should be assumed to exist until leak rate is 1 gpm (1440 gpd) which is more shutdown cooling is in operation and releases from the conservative than the Technical Specification steam generators have been terminated. limit of 0.1 gpm per steam generator.

Shutdown cooling is assumed to be in operation within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after accident initiation. Leakage via the MSIV above seat drain orifices is assumed to continue for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. (Section 4.7.4.2)

No releases from the secondary side are postulated for the case of the containment release pathway. (Section 4.7.4.1) 7.2 The density used in converting volumetric leak rates Conforms The density is assumed to be 8.33 Ibm/gal.

(e.g., gpm) to mass leak rates (e.g., lbm/hr) should be (Section 4.7.5) consistent with the basis of surveillance tests used to show compliance with leak rate technical specifications.

These tests typically are based on cooled liquid.

The facility's instrumentation used to determine leakage I 34 of 37

NOC-AE-07002127 Attachment 6 Table F: Conformance with Regulatory Guide 1.183 Appendix H (PWR Control Rod Ejection Accident)

RG Section Regulatory Position Analysis Comments a typically is located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft3).

7.3 All noble gas radionuclides released to the secondary Conforms 100% of the noble gases released via the system are assumed to be released to the environment secondary system are assumed to be released without reduction or mitigation. to the environment without reduction or mitigation.. (Section 4.7.4.2) 7.4 The transport model described in assumptions 5.5 and Conforms (Section 4.7.1.1) 5.6 of Appendix E of RG 1.183 should be utilized for iodine and particulates.

Table G: Conformance with Regulatory Guide 1.183 Appendix G (PWR Locked Rotor Accident)

RG Section Regulatory Position Analysis Comments a 1 Assumptions acceptable to the NRC staff regarding core Conforms Assumptions regarding core inventory and the inventory and the release of radionuclides from the fuel release of radionuclides from the fuel are are in Regulatory Position 3 of this regulatory guide. The those in Regulatory Position 3 of RG 1.183.

release from the breached fuel is based on Regulatory The release from the breached fuel is based Position 3.2 of this guide and the estimate of the on Regulatory Position 3.2 of RG 1.183. 10%

number of fuel rods breached. of the fuel rods are assumed to fail. (Section 4.8.3) 2 If no fuel damage is postulated for the limiting event, a Not 10% fuel failure is assumed. A radiological radiological analysis is not required as the consequences Applicable calculation analysis was performed.

of this event are bounded by the consequences projected for the main steam line break outside containment.

3 The activity released from the fuel should be assumed to Conforms (Section 4.8.5) be released instantaneously and homogeneously through the primary coolant.

35 of 37

NOC-AE-07002127 Attachment 6 Table G: Conformance with Regulatory Guide 1.183 Appendix G (PWR Locked Rotor Accident)

RG Section Regulatory Position Analysis Comments a 4 The chemical form of radioiodine released from the fuel Bounds The chemical form of the iodine in the RCS is should be assumed to be 95% cesium iodide (CsI), 4.85 95% CsI, 4.85% elemental, and 0.15%

percent elemental iodine, and 0.15 percent organic organic. The chemical form of the iodine iodide. Iodine releases from the steam generators to the released from the secondary side to the environment should be assumed to be 97% elemental environment is 4.2% elemental, 13.1%

and 3% organic. These fractions apply to iodine released organic, and 82.7% particulate. (Section as a result of fuel damage and to iodine released during 4.8.3.4) normal operations, including iodine spiking.

5.1 The primary-to-secondary leak rate in the steam Conforms The assumed primary-to-secondary leak rate generators should be assumed to be the leak-rate- is 1 gpm (1440 gpd) which is more limiting condition for operation specified in the technical conservative than the Technical Specification specifications. The leakage should be apportioned limit of 0.1 gpm per steam generator. A 0.35 between the steam generators in such a manner that the gpm (504 gpd) leak rate is assumed for one calculated dose is maximized. steam generator that experiences tube uncovery during the accident due to a postulated single failure in the feedwater system. A 0.65 gpm (936 gpd) leak rate is assumed for the remaining steam generators.

(Section 4.8.4 and 4.8.5) 5.2 The density used in converting volumetric leak rates Conforms The density is assumed to be 8.33 Ibm/gal.

(e.g., gpm) to mass leak rates (e.g., lbm/hr) should be (Section 4.8.5) consistent with the basis of surveillance tests used to show compliance with leak rate technical specifications.

These tests are typically based on cool liquid. Facility instrumentation used to determine leakage is typically located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft3).

5.3 The primary-to-secondary leakage should be assumed to Conforms The primary-to-secondary leakage of 1 gpm is 36 of 37

NOC-AE-07002127 Attachment 6 Table G: Conformance with Regulatory Guide 1.183 Appendix G (PWR Locked Rotor Accident)

RG Section Regulatory Position Analysis Comments a continue until the primary system pressure is less than assumed to continue for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the the secondary system pressure, or until the temperature accident. Eight hours after the accident, the of the leakage is less than 100'C (2120 F). The release of Residual Heat Removal System starts radioactivity should be assumed to continue until operation to cool down the plant. No further shutdown cooling is in operation and releases from the steam or activity is released to the steam generators have been terminated, environment. (Section 4.8.5) 5.4 The release of fission products from the secondary Conforms The release of fission products from the system should be evaluated with the assumption of a secondary system is evaluated with the coincident loss of offsite power. assumption of a coincident loss of offsite power. The secondary system condensers are not available for dumping steam. (Section 4.8.5) 5.5 All noble gas radionuclides released from the primary Conforms Noble gas radionuclides released from the system are assumed to be released to the environment primary system are assumed to be released to without reduction or mitigation. the environment without reduction or mitigation. (Section 4.8.4) 5.6 The transport model described in assumptions 5.5 and Conforms (Section 4.8.1.1) 5.6 of Appendix E of RG 1.183 should be utilized for iodine and particulates.

a The Section or Table number indicated in the parentheses, in this column, refers to the Section or Table in Attachment 1, "Licensee's Evaluation" of this licensing amendment request, where the regulatory position is addressed.

37 of 37

NOC-AE-07002127 Attachment 7 NRC Regulatory Issue Summary 2006-04 Table

NOC-AE-07002127 Page 1 of 7 NRC Regulatory Issue Summary 2006-04 Table The purpose of RIS 2006-04 is to discuss the more frequent and significant issues encountered by the NRC staff during its review of AST submittals and to provide information for licensees to consider when developing submittals for implementation of an AST. This table provides comments describing how STPNOC addressed each RIS issue in the application request for implementing AST.

RIS Issue I Licensee Comments a

1. Level of Detail Contained in LARs (1) The AST amendment request should provide justification for Provided in Section 2.0 of the Licensee's Evaluation for each each individual proposed change to the technical specifications individual proposed change to the TS.

(TS)

(2) The AST amendment request should identify and justify each Section 2.0 of the Licensee's Evaluation provides an overview change to the licensing basis accident analyses justification. Section 4.0 and 5.0 provide the detailed justification.

(3) The AST amendment request should contain enough details Sufficient detail in tabular format is provided in Section 4.0 of the (e.g., assumptions, computer analyses input and output) to allow Licensee's Evaluation to allow the NRC staff to confirm the dose the NRC staff to confirm the dose analyses results in independent analyses results in independent calculations. In addition, the dose calculations. consequent calculations will be submitted under separate cover letter.

Licensees should identify the most current analyses, assumptions, The most current analyses, assumptions, and TS changes are and TS changes in their submittal and supplements to the identified throughout Attachment 1 of the Licensing Amendment submittal. Request.

2. Main Steam Isolation Valve (MSIV) Leakage and Not applicable to PWRs Fission Product Deposition in Piping

NOC-AE-07002127 Page 2 of 7 RIS Issue I Licensee Comments a

3. Control Room Habitability Use of non-ESF ventilation systems during a DBA should not be No credit is taken for use of non-ESF ventilation systems during a assumed unless the systems have emergency power and are part DBA. (Section 4.2.2) of the ventilation filter testing program in Section 5 of the TS.

Generic Letter (GL) 2003-01, "Control Room Habitability" The CR make-up flow is increased from 2000 cfm to 2200 cfm requested licensees to confirm the ability of their facility's control for conservatism. The CR make-up filtration is conservatively room to meet applicable habitability regulatory requirements. The ignored. 100 cfm of unfiltered inleakage is assumed in addition GL placed emphasis on licensees confirming that the most to the 2200 cfm of make-up flow that is assumed to experience no limiting unfiltered inleakage into the control room envelope filtration. (Section 4.2.2)

(CRE) was not greater than the value assumed in the DBA analyses. Unfiltered inleakage testing performed in Unit 1 using the tracer gas method in response to GL 2003-01 measured 9.4 scfm for the limiting case. Unfiltered inleakage testing performed in Unit 2 using the tracer gas method in response to GL 2003-01 measured 62 scfm for the limiting case. (Section 4.2.2)

Some AST amendment requests proposed operating schemes for Control room and other ventilation systems which affect areas the control room and other ventilation systems which affect areas adjacent to the CRE and are the same as the operation and adjacent to the CRE and are different from the manner of performance described in the response to the GL. No credit is operation and performance described in the response to the GL taken for the control room ventilation system makeup filters in the without providing sufficient justification for the proposed changes AST application. (Section 4.2.2) in the operating scheme.

4. Atmospheric Dispersion Licensees have the option to adopt the generally less conservative Updated CR y/Q values for releases from the containment, from (more realistic) updated NRC staff guidance on determining X/Q the plant vent, and from the PORV nearest the CR intake were values in support of design basis control room radiological calculated using the computer code ARCON96 using the methods habitability assessments provided in RG 1.194, "Atmospheric of Regulatory Guide 1.194. The revision to the atmospheric Relative Concentrations for Control Room Radiological dispersion analyses is provided in Section 4.1 of the Licensee's Habitability Assessments at Nuclear Power Plants". Evaluation. MET data and inputs to ARCON 96 plus marked-up updates to Chapter 2 of the UFSAR will be provided as part of the submittal.

NOC-AE-07002127 Page 3 of 7 RIS Issue Licensee Comments a Regulatory positions on X/Q values for offsite (i.e., exclusion area The XJQ values for offsite locations were evaluated using the boundary and low population zone) accident radiological methods of Regulatory Guide 1.145. (Section 4.1) MET data and consequence assessments are provided in RG 1.145, inputs to PAVAN plus marked-up updates to Chapter 2 of the "Atmospheric Dispersion Models for Potential Accident UFSAR will be provided as part of the submittal.

Consequence Assessments at Nuclear Power Plants".

The submittal should include a site plan showing true North and Figure 4.1-13 provides a simplified plot plan with release points indicating locations of all potential accident release pathways and and receptors.

control room intake and unfiltered inleakage pathways (whether assumed or identified during inleakage testing).

The submittal should include a justification for using control Section 4.2.2.1 of Attachment 1 provides justification.

room intake X/Q values for modeling the unfiltered inleakage, if applicable.

The submittal should include a copy of the meteorological data The revised X/Q values used for the AST application have been inputs and program outputs along with a discussion of developed using more recent meteorological data than that used assumptions and potential deviations from staff guidelines, for the CLB. These more recent data were obtained for the years Meteorological data input files should be checked to ensure 2000 to 2004 (five years worth of data) and are documented in quality (e.g., compared against historical or other data and against ABS Consulting Report R-1459208-01, September 2005.

the raw data to ensure that the electronic file has been properly Submittal will include a copy of the meteorological data inputs formatted, any unit conversions are correct, and invalid data are and program outputs. The MET data was collected per station properly identified). procedures. (Section 4.1)

When running the control room atmospheric dispersion model No credit is taken for an elevated release. (Section 4.1.3)

ARCON96, two or more files of meteorological data representative of each potential release height should be used if X7Q values are being calculated for both ground-level and elevated releases.

In addition, licensees should be aware that All releases are assumed to be at ground level, and therefore, only 10 m (lower) elevation wind speed data is relevant. (Section (1) two levels of wind speed and direction data should always be 4.1.2) provided as input to each data file, (2) fields of "nines" (e.g., 9999) should be used to indicate invalid Valid wind direction data provided from 1 to 3600.

or missing data, and (Tables 4.1-1 and Figures 4.1-2 through 4.1-7)

(3) valid wind direction data should range from 1' to 3600. 1

NOC-AE-07002127 Page 4 of 7 RIS Issue Licensee Comments a Licensees should also provide detailed engineering information Buoyancy or mechanical jets of high energy releases are not when applying the default plume rise adjustment cited in RG credited. (Section 4.1.3) 1.194 to control room X/Q values to account for buoyancy or mechanical jets of high energy releases.

This information should demonstrate that the minimum effluent No credit is taken for an elevated release. (Sections 4.1.2 and velocity during any time of the release over which the adjustment 4.1.3) is being applied is greater than the 9 5 th percentile wind speed at the height of release.

When running the offsite atmospheric dispersion model PAVAN, No credit is taken for an elevated release. (Section 4.1.2) two or more files of meteorological data representative of each potential release height should be used if X/Q values are being calculated for pathways with significantly different release heights (e.g., ground level versus elevated stack).

The joint frequency distributions of wind speed, wind direction, An adequate number of categories at lower wind speed were used.

and atmospheric stability data used as input to PAVAN should Seven wind speed groups for STP are used which are as follows:

have a large number of wind speed categories at the lower wind 0.5 mph, 3.5 mph, 7.5 mph, 12.5 mph, 18.5 mph, 24.5 mph, and speeds in order to produce the best results 36.0 mph (i.e., 0.22 m/s, 1.56 m/s, 3.35 m/s, 5.59 m/s, 8.27 m/s, 10.95 m/s, and 16.09 m/s). These are judged to be adequate for determining the offsite X/Q values using PAVAN.

(Section 4.1.2)

5. Modeling of ESF Leakage The radiological consequences from the postulated [ESF] leakage The postulated [ESF] leakage is analyzed and combined with should be analyzed and combined with consequences postulated consequences postulated for other fission product release paths to for other fission product release paths to determine the total determine the total calculated radiological consequences from the calculated radiological consequences from the [loss-of-coolant [loss-of-coolant accident] LOCA. (Section 4.3.3.2) accident] LOCA.

Licensees should account for ESF leakage at accident conditions ESF leakage was accounted for at accident conditions. (Section in their dose analyses so as not to underestimate the release rate. 4.3.3.2)

NOC-AE-07002127 Page 5 of 7 RIS Issue Licensee Comments a In Appendix A to RG 1.183, Regulatory Position 5.5, the NRC The RG 1.183 recommended value of 10% is used only for the staff provided a conservative value of 10 percent as the assumed first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Beyond that time, the release fraction is increased amount of iodine that may become airborne from ESF leakage to 16% (24-480 hours) and then to 25% (480 hours0.00556 days <br />0.133 hours <br />7.936508e-4 weeks <br />1.8264e-4 months <br /> to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />) that is less than 212 'F. based on the calculated volatility of the iodine for the pH values over those intervals relative to the volatility for a pH of 7.0.

(Section 4.3.3.2)

Figure 3.1 in NUREG/CR-5950 can be used to quantify the The calculation methodology for containment sump pH control amount of elemental iodine as a function of the sump water pH was based on the approach outlined in NUTREG-1465 and and the concentration of iodine in the solution. In some cases, NUREG/CR-5950.

however, licensees have misapplied this figure. Rather than using the total concentration of iodine (i.e., stable and radioactive), The assessment included 1-127 and 1-129.

licensees based their assessment on only the radioactive iodine in the sump water. By using only the radioactive iodine, licensees (Section 4.3.3.1.2.1) have underestimated how much iodine evolves during post-accident conditions.

6. Release Pathways Changes to the plant configuration associated with an LAR (e.g., The AST application re-analyzes the design basis dose an "open" containment during refueling) may require a re- calculations for an open containment during refueling.

analysis of the design basis dose calculations. A request for TS (Section 4.4.2) modifications allowing containment penetrations (i.e., personnel air lock, equipment hatch) to be open during refueling cannot rely The current dose analysis considered the open personnel air lock on the current dose analysis if this analysis has not already and equipment hatch as release paths.

considered these release pathways. Releases from personnel air (Section 4.4.2) locks and equipment hatches exposed to the environment and containment purge releases prior to containment isolation need For the LOCA analysis, the duration of flow through open purge to be addressed. valves is assumed to be 23 seconds - same as the current licensing basis. (Section 4.3.5)

NOC-AE-07002127 Page 6 of 7 RIS Issue Licensee Comments a Licensees are responsible for identifying all release pathways and Revised control room, exclusion area boundary, and low for considering these pathways in their AST analyses, consistent population zone atmospheric dispersion factors (X/Q) for the with any proposed modification. containment leakage, plant vent, and steam generator (SG) secondary side power-operated relief valve (PORV) release paths were calculated. (Section 4.1)

A new release path consisting of the containment electrical penetration area volume, the leak rate per penetration, the number of penetrations, and the ventilation exhaust rate for the electrical penetration area are analyzed to determine the post-accident transient activity airborne in the electrical penetration area to support the calculation of the gamma shine contribution to the CR. (Section 4.2.2.1 and Table 4.3-11)

7. Primary to Secondary Leakage Some analysis parameters can be affected by density changes that All analyses assume a water density of 1 gram/cubic centimeter occur in the process steam. The NRC staff continues to find errors (i.e. 8.33 Ibm/gallon).

in LAR submittals concerning the modeling of primary to (Sections 4.5.5, 4.6.5, 4.7.5 and 4.8.5) secondary leakage during a postulated accident. This issue is discussed in Information Notice (IN) 88-31, "Steam Generator Tube Rupture Analysis Deficiency," (Ref. 11) and Item 3.f in RIS 2001-19. An acceptable methodology for modeling this leakage is provided in Appendix F to RG 1.183, Regulatory Position 5.2.

8. Elemental Iodine Decontamination Factor (DF)

Appendix B to RG 1.183, provides assumptions for evaluating the The depth of water over the damaged fuel is not less than 23 feet.

radiological consequences of a fuel handling accident. If the water Due to the submergence of the damaged fuel, the iodine release is depth above the damaged fuel is 23 feet or greater, Regulatory assumed to experience a DF of 200 per RG 1.183.

Position 2 states that "the decontamination factors for the (Section 4.4.4) elemental and organic [iodine] species are 500 and 1, fespectively, giving an overall effective decontamination factor of 200." However, an overall DF of 200 is achieved when the DF for elemental iodine is 285, not 500.

NOC-AE-07002127 Page 7 of 7 RIS Issue Licensee Comments a

9. Isotopes Used in Dose Assessments For some accidents (e.g., main steamline break and rod drop), Noble gas and cesium isotopes were included in the dose licensees have excluded noble gas and cesium isotopes from the assessment.

dose assessment. The inclusion of these isotopes should be addressed in the dose assessments for AST implementation.

10. Definition of Dose Equivalent 131i In the conversion to an AST, licensees have proposed a The definition of DE 1-131 is modified to reflect that the dose modification to the TS definition of dose equivalent 131I. conversion factors are those listed in Federal Guidance Report 11.

Although different references are available for dose conversion (Section 2.0 and 4.2.4.1) factors, the TS definition should be based on the same dose conversion factors that are used in the determination of the reactor coolant dose equivalent iodine curie content for the main steamline break and steam generator tube rupture accident analyses.

11. Acceptance Criteria for Off-Gas or Waste Gas System Release As part of full AST implementation, some licensees have Accident not included with this submittal.

included an accident involving a release from their off-gas or waste gas system.

12. Containment Spray Mixing Some plants with mechanical means for mixing containment air The volumetric flow rate between sprayed and unsprayed regions have assumed that the containment fans intake air solely from a of containment is found in Table 4.3-11 in the Licensee's sprayed area and discharge it solely to an unsprayed region or Evaluation.

vice versa. Without additional analysis, test measurements or further justification, it should be assumed that the intake of air by Less than 10% of this total will re-circulate in the unsprayed containment ventilation systems is supplied proportionally to the region. The dose is insensitive to mixing flow bypass of this sprayed and unsprayed volumes in containment, magnitude.

a The Section or Table number indicated in the parentheses, in this column, refers to the Section or Table in Attachment 1, "Licensee's Evaluation" of this licensing amendment request.