ULNRC-06636, Enclosure 6 - FSAR Markups (for Information Only)

Enclosure 6 - FSAR Markups (for Information Only)

ML21272A174
Person / Time
Site:	Callaway
Issue date:	09/28/2021
From:	Ameren Missouri, Union Electric Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML21272A167	List: ML21272A168 ML21272A170 ML21272A171 ML21272A172 ML21272A173 ULNRC-06636, Enclosure 6 - FSAR Markups (for Information Only)
References
ULNRC-06636
Download: ML21272A174 (374)

Text

ULNRC-06636 Page 1 of 374 Attachment A - List of Affected FSAR Sections

ULNRC-06636 Page 2 of 374 FSAR Section Title Change Description Page Chapter 1: Introduction and General Description of the Plant 1.1.7 Design Bases replaced 10 CFR 100 with 10 CFR 50.67 1.1-3 Table 1.1-1 (Sheet 3) Acronyms Added CEDE, EDE, and TEDE acronyms Table Table 1.1-1 (Sheet 5) Acronyms Added CEDE, EDE, and TEDE acronyms Table Table 1.3-4 (Sheet 14) Compliance with NRC Regulations Added 50.67 Table Table 1.3-4 (Sheet 26) Compliance with NRC Regulations Deleted text under 10 CFR 100.11 Table Chapter 2: Site Characteristics 2.3.4 Short-Term (Accident) Diffusion Estimates Reference to Regulatory Guide 1.194, removed old methodology 2.3-1 2.3.4 Short-Term (Accident) Diffusion Estimates Added insert 2.3.4.2.2.2, removed reference 1 2.3-2 Table 2.3-1 Limiting Atmospheric Dispersion Factor Revised with NAI-1990-006 Table Chapter 2 Addendum: Site Characteristics Addendum 2 TOC Added new Section 2.3.4.4 2.0-ii Addendum 2 List of Tables Added Tables 2.3-87 through 2.3-98 2.0-xi Addendum 2, 2.3.4 Short-Term Diffusion Estimates Added note that Sections 2.3.4.1-2.3.4.3 are historical 2.3-59 Addendum 2 Alternative Source Term Short-Term Diffusion Estimates Added new Section 2.3.4.4 2.3-64 Addendum 2 References References Section 2.3 Added 2 References 2.3-75 Addendum 2 References References Section 2.3 Add Reference 2.3-76 Chapter 3: Design Of Structures, Components, Equipment and Systems 3.0 Design of Structures, Components, Equipment and Systems Replaced 10 CFR 100 with 10 CFR 50.67 for offsite exposures 3.1-1 3.1.4 Criterion 19 - Control Room replaced "whole body or its equivalent" with TEDE 3.1-16 3.1.4 Criterion 19 - Control Room Replaced 10 CFR 100 with 10 CFR 50.67 for offsite exposures 3.1-16 3.1.4 Criterion 19 - Control Room deleted TID-14844 in discussion 3.1-16 3.1.4 Criterion 19 - Control Room deleted "thyroid" 3.1-16 3.1.6 Criterion 41, Containment Atmosphere Cleanup Replaced RG 1.4 with RG 1.183 3.1-32 3.2c Classification of Structures, Components and Systems replaced 10 CFR 100 with 10 CFR 50.67 3.2-1 3.6.2.3.2.1 replaced 10 CFR 100 with 10 CFR 50.67 3.6-23

ULNRC-06636 Page 3 of 374 FSAR Section Title Change Description Page 3.7(N) Seismic Design replaced 10 CFR 100 with 10 CFR 50.67 3.7(N)-1 3.8.1.2.1 Regulations Added c. 10 CFR 50.67, Accident Source Term 3.8-4 3.11(B).1.1.2c Safety-Related System Listing replaced 10 CFR 100 with 10 CFR 50.67 3.11(B)-3 3.11(B).1.2.2 Accident Environments - Inside Containment Added note pointing to Chapter 15.6.5 3.11(B)-4 3.11(B).1.2.2 Safety-Related System Listing replaced spray removal coefficients, DFs with AST values 3.11(B)-5 3.11(B).5.2 Equipment Operability Replaced RG 1.4 with RG 1.183 3.11(B)-21 Appendix 3A Conformance to NRC Regulatory guides Added 1.183 and removed 1.4 and 1.25 3A-1 Appendix 3A Conformance to NRC Regulatory Guides Noted that RG 1.4 has been replaced with RG 1.183, deleted 3A-3 reference to Table 15.6.7 Appendix 3A Conformance to NRC Regulatory Guides Noted the RG 1.25 has been replaced with RG 1.183, deleted 3A-11 reference to Table 15.7.2 Appendix 3A Conformance to NRC Regulatory Guides Added Regulatory Guide 1.183 to list 3A-55 Appendix 3A Conformance to NRC Regulatory Guides Added Regulatory Guide 1.194 to list 3A-58 Chapter 3 Addendum: Design Of Structures, Components, Equipment and Systems Addendum 3 Appendix 3.A -Confirmance to NRC Regulatory Guides Added Reg Guide 1.183 and 1.194 to the list, deleted REG 1.4 and 3.A-1 1.25 from list Addendum 3 Appendix 3.A -Confirmance to NRC Regulatory Guides Added Rev 1 of Reg Guide 1.145 and referred to Site Addendum 3.A-9 Section 2.3.4.4 Addendum 3 Appendix 3.A -Confirmance to NRC Regulatory Guides Added Reg Guide 1.194 and referred to Site Addendum Section 3.A-9 2.3.4.4 Chapter 4: Reactor 4.2.1 Fuel System Design replaced 10 CFR 100 with 10 CFR 50.67 4.2-1 4.3.1 Nuclear Design, Design Bases replaced 10 CFR 100 with 10 CFR 50.67 4.3-1

ULNRC-06636 Page 4 of 374 FSAR Section Title Change Description Page Chapter 6: Engineered Safety Features 6 TOC Revised title, added section 6.0-vi 6.0 Engineered Safety Features replaced 10 CFR 100 with 10 CFR 50.67 6.1-1 6.2.1.1.1 Design Bases replaced 10 CFR 100 with 10 CFR 50.67 6.2.1-1 6.2.1.1.1f Design Bases replaced 10 CFR 100 with 10 CFR 50.67 6.2.1-4 6.3.3 Safety Evaluation replaced 10 CFR 100 with 10 CFR 50.67 6.3-26 6.4 SAFETY DESIGN BASIS NINE added (event) to SAFETY DESIGN BASIS NINE description 6.4-2 6.4.4 SAFETY EVALUATION SEVEN revised to address inhalation, immersion, and transit 6.4-6 6.5.1.3 Safety Evaluation replaced 10 CFR 100 with 10 CFR 50.67 6.5-3 6.5.2.1.1 Safety Design Bases replaced 10 CFR 100 with 10 CFR 50.67 6.5-4 6.5.2.3 Safety Evaluation change to reference Section 6.5A-4, change spray removal 6.5-6 coefficient from 10 to 20 6.5.2.3 Safety Evaluation deleted "iodine" 6.5-7 6.5.2.3 Safety Evaluation replaced DF's and spray removal coefficients, replaced 10 CFR 100 6.5-8 with 10 CFR 50.67, deleted "iodine" 6.5.3.1 Primary Containment replaced 10 CFR 100 with 10 CFR 50.67 6.5-9 6.5.5 References Added reference 4 6.5-10 Table 6.5-1 ESF Filtration Systems Input Parameters A.3 and 6.5A.44 Table Table 6.5-2 Input Parameters and Results of Spray Removal Analysis Updated input parameters based on NAI-1990-004. Additional work required for some values Table 6.5A Appendix 6.5A Removal Models for the Containment Spray System Deleted "iodine" from title 6.5A-1 6.5A.1 Particulate Iodine Model (for EQ Dose Consequences) Renamed title 6.5A-2 6.5A.2 Elemental iodine Model for EQ Dose Calculations removed pointer to use in Chapter 15.6.5 6.5A-10 6.5A.3 Elemental Iodine Model for Offsite and Control Room Dose Calculations Replaced entire section 6.5A-11 PARTICULATE IODINE MODEL FOR OFFSITE AND CONTROL ROOM DOSE New section, moved references to 6.5.A.5 6.5A.4 6.5A-11 CALCULATIONS 6.5.A5 References Added references 14 and 15 6.5A-12 Chapter 7: Instrumentation and Controls 7.1.1 Identification of Safety-Related Systems added 10 CFR 50.67 and RG 1.183, as appropriate 7.1-1 Chapter 9 9.1.4.1.1 Safety Design Bases (of the Fuel Handling System), Replaced 10 CFR 100 with 10 CFR 50.67 and RG 1.183 9.1-26 SAFETY DESIGN BASIS SIX 9.1.4.3 Safety Evaluation Replaced 10 CFR 100 with 10 CFR 50.67 and RG 1.183 9.1-50 Table 9.1-3 DESIGN COMPARISION TO REGULATORY POSITIONS OF REGULATORY GUIDE Regulatory Guide 1.13, position 4 changed to point to RG 1.183 Table 1.13 REVISION 1, DATED DECEMBER 1975, TITLED SPENT FUEL STORAGE instead of RG 1.25 for assumptions for the inventory of radioactive FACILITY DESIGN BASIS materials available for leakage from the building 9.4.1.2.1 General Description Added reference to the model for control room dose analysis in 9.4-3 Appendix 15A.

9.4.2 Fuel Building HVAC added credit for emergency exhaust system 9.4-16 9.4.2.3 Safety Evaluation (of the Fuel Building HVAC) replaced 10 CFR 100 with 10 CFR 50.67 9.4-24

ULNRC-06636 Page 5 of 374 FSAR Section Title Change Description Page Chapter 12 12.3.2.2.6 Control Room Shielding Design Replaced whole body with TEDE 12.3-12 12.3.3.2c Design Criteria Replaced 10 CFR 100 with 10 CFR 50.67 12.3-14 12.3.4.1.1.1 Safety Design Bases replaced 10 CFR 100 with 10 CFR 50.67 12.3-18 Chapter 15 TOC Table of Contents Updated Section 15A 15.0-viii TOC Table of Contents Added Sections 15A.5, 15A.6 and Appendix 15B 15.0-ix List of Tables Table of Contents deleted RG 1.4, 1.25 (Table 15.6-7, 15.7-2) 15.0-xii List of Tables Table of Contents Renamed Table 15A-3, Added Tables 15A-6, -7 and -8 15.0-xiii Added Conformance Tables 15B-1 through 15B-7 for RG 1.183 List of Figures List of Figures Replaced figures 15.0-xxv, -xxvi,

-xxxiv, -xxxv List of Figures List of Figures Figures 15A-1 and 15A-2 to be replaced 15.0-xxxvii 15.0.1.4 Condition IV - Limiting Faults Replaced 10 CFR 100 with 10 CFR 50.67 15.0-5 15.0.9 Fission Product Inventories Replaced with data from NAI-1990-002 15.0-12, -13 15.0.11.8 RETRAN added discussion of RETRAN-3D 15.0-17 15.0.14 References Added reference 22 to SCALE; added Reference 23 to RETRAN-3D 15.0-22 Table 15.0-2(Sheet 6) Table 15.0-2 corrected typo Table Table 15.0-7 Single Failures Assumed in Accident Analyses Added SGTR information Table MSLB 15.1.5.1 Steam system piping failure replaced 10 CFR 100 with 10 CFR 50.67 15.1-14 15.1.5.3.1.2 Assumptions and Conditions Added pointer to Tables 15B-1 and 15B-4 15.1-21 Added reference to Tables 15B-1 and 15B-4 for RG 1.183, revised 15.1-22 15.1.5.3.1.2 Assumptions and Conditions timing 15.1.5.3.1.2 and 15.1.5.3.1.3 Assumptions and Conditions added clarification text 15.1-23 15.1.5.3.2 Identification of Uncertainties and Conservatisms in the Analysis revised paragraph a, added paragraph d. 15.1-24 15.1.5.3.3.2 Steam system piping failure replaced 10 CFR 100 with 10 CFR 50.67, duration of accident 15.1-25 Table 15.1-3 Parameters used in evaluating radiological consequences of MSLB revised input parameters Table Table 15.1-4 Radiological Consequences of MSLB revised values Table Loss of Offiste Power 15.2.6.3.1.1 Physical Model for LOOP Radiological Consequences replace 8 with 7.275 hours0.00318 days <br />0.0764 hours <br />4.546958e-4 weeks <br />1.046375e-4 months <br /> 15.2-13 15.2.6.3.1.1 Physical Model for LOOP Radiological Consequences revised assumptions and conditons 15.2-14 15.2.6.3.2 Identification of Uncertainties in, and Conservative of, the Analysis revised values 15.2-15 15.2.6.3.3.2 Loss of NonEmergency AC Power Radiological Consequences replaced 10 CFR 100 with 10 CFR 50.67 15.2-16 Table 15.2-2 Parameters Used In Evaluating Radiological Consequences of Loss of revised parameters Table Nonemergency AC Power Table 15.2-3 Radiological Consequences revised doses Table

ULNRC-06636 Page 6 of 374 FSAR Section Title Change Description Page Locked Rotor 15.3.3.3 Radiological Consequences revised assumptions, models used, conservatisms and results 15.3-9, -10, -11, -

12 Table 15.3-3 Parameters Used in Evaluating the Rsdiological Consequences of a Locked revised parameters Table Rotor Event Table 15.3-4 Radiological Consequences of a Locked Rotor Event revised doses Table Rod Ejection 15.4.8.3 Radiological Consequences revised assumptions, models used, conservatisms and results 15.4-43 through 15.4-47 Table 15.4-3 Parameters Used in Evaluating the RCCA Ejection Accident revised parameters sheet 1 and 2 Table 15.4-4 Radiological Consequences of a Rod-Ejection Accident revised doses Table Letdown Line 15.6.2.1.1.2 Assumptions and Conditions (Radiological Consequences of Letdown Line revised assumptions and conditons 15.6-4 Break) 15.6.2.1.1.3 Mathematical Models Uned in the Analysis added in control room 15.6-5 15.6.2.1.3.2 Dose to Receptor for Letdown Line renamed title, replaced 10 CFR 100 with 10 CFR 50.67 15.6-6 Steam Generator Tube Failure 15.6.3.1.2 Analysis of Effects and Consequences revised description 15.6-10, -11, -12, -

13 15.6.3.1.3 Radiological Consequences revised description 15.6-14 through -

19 15.6.3.2.1 Identification of Causes and Accident Description Revised discussion of ASD conservatism 15.6-20 15.6.3.2.1, Identification of Causes and Accident Description Revised length of time to RHR cut-in, revised feedwater 15.6-21 15.6.3.2.2 temperature 15.6.3.2.2 Analysis of Effects and Consequences Revised assumptions 15.6-22 15.6.3.2.2 Analysis of Effects and Consequences Revised AFW flow to SGs, revised assumptions 15.6-23,

-24, -25 15.6.3.2.2 results Analysis of Effects and Consequences, Results revised operator timing 15.6-27 15.6.3.2.3 Radiological Consequences revised results 15.6-28, -29,

-30 15.6.3.3.2 Conclusions revised description 15.6-31 LOCA 15.6.5.4 Radiological Consequences revised description 15.6-44, -45,

-46 15.6.5.4.1.2 Radioactive Releases Due to Leakage from ECCS and Containment Spray added "air" to RWST description, revised RWST leakage discussion 15.6-47 Recirculation Lines and release percentages 15.6.5.4.2 Identification of Uncertainties and Conservatisms in the Analysis revised conservatisms 15.6-48, -49 15.6.5.4.3.2 and Doses at EAB and LPZ replaced total body and thyroid with TEDE, revised 10 CFR 100 (and 15.6-50 15.6.5.4.3.3 GDC-19) with 10 CFR 50.67 Chapter 15.6 references 15.6.7 References Added FGR 11 and 12 15.6-53

ULNRC-06636 Page 7 of 374 FSAR Section Title Change Description Page Chapter 15.6 Tables Table 15.6-1 (Sheet 2) Time Sequence of Events for Incidents which Result in a Decrease in Reactor Revised timing SGTR with overfill Coolant Inventory Table 15.6-2 Parameters used In Evaluating the Radiological Consequence of the CVCS revised parameters Table Letdown line Rupture Outside of Containment Table 15.6-3 Radiological Consequences of a CVCS Letodown Line Break Outside revised doses Table Containment Table 15.6-4 Parameters used in Evaluating the Radiological Consequences of a SGTR revised input papameters Table Table 15.6-5 Radiological Consequences of a Steam Generator Tube Rupture with Stuck- revised dose consequence values Table Open Atmospheric Steam Dump Valve Table 15.6-5A Radiological Consequences of a Steam Generator Tube Rupture with revised dose consequence values Table Overfill Table 15.6-6 Parameters used in Evaluating the Radiological Consequences of a Loss-of- revised inputs Table Coolant Accident Table 15.6.7 Design Comparison to RG 1.4 Deleted table Table Table 15.6-8 Radiological Consequences of a LOCA Replaced values Table 15.7 Radioactive Release from A Subsystem or Component 15.7.1.5.1.3 Mathematical Models Used in the Analysis Revised referenced section numbers 15.7-3 15.7.2.5.1.3 Mathematical Models Used in the Analysis Revised referenced section numbers 15.7-7 Fuel Handling Accident 15.7.4.4 Barrier Performance added pointer to Section 15.7.4.5.1.2, removed factor of 100 15.7-9 15.7.4.5.1.2 Assumptions and Conditions revised description 15.7-12, -13 15.7.4.5.2 Identification fo Uncertainties and Conservatisms in Analysis revised to reference RG 1.183 guidance instead of description 15.7-14 15.7.4.5.2.2 Doses to Receptor at the EAB, LPZ and Control Room revised to reference 50.67 and RG 1.183, added insert 15.7.5 15.7-15 Table 15.7-2 Design Comparison to RG 1.25 Deleted table Table Table 15.7-7 Parameters Used in Evalauting the Radiological Consequences of a Fuel- Revised parameters Table Handling Accident Table 15.7-8 Radiological Consequences of a Fuel Handling Accident Revised values Table Appendix 15A 15A General Acccident Parameters entire section revision 15A-1 through 15A-14 Table 15A-1 Parameters used in Accident Analysis revised Table Table 15A-2 Limiting short term Atmospheric Dispersion Factors replaced Table Table 15A-3 Core Inventory (Ci) Replaced and Renamed Table Table 15A-4 Dose Conversion Factors Used in Accident Analysis Added DCFs for events reanalyzed with AST Table Table 15A-5 Initial Radioactivity for Accidents that use the Primary-to-Secondary added nuclides Sheets 1-5 Leakage Relase Pathway

ULNRC-06636 Page 8 of 374 1.1.6 SCHEDULE FOR FUEL LOADING AND OPERATION On June 11, 1984 Union Electric received a low-power (5%) license to operate the Callaway Plant with initial criticality being achieved on October 2, 1984. The full-power license was issued to Union Electric on October 18, 1984 and commercial operation began on April 9, 1985.

1.1.7 DESIGN BASES As used within this FSAR, the design bases are a list of requirements that the system must meet in order to:

a. Perform directly a specified safety or power generation function including support of another function (e.g., provide cooling water flow for other components, maintain a given compartment temperature).

b. Comply with a regulatory or statutory requirement or guideline (e.g., a jurisdictional building code).

c. Meet a specific operator interface, startup, or specific testing requirement.

d. Meet a design classification or code requirement (e.g., be designed to withstand the safe shutdown earthquake). Items implicit in contemporary design practices (e.g., use of the English system of weights and measures or the exercise of good engineering practice) are not specified as design bases.

Safety design bases are engineering objectives which must be met by safety-related structures, systems, or components. Safety-related items are defined as those plant features necessary to ensure the following:

a. The integrity of the reactor coolant pressure boundary

b. The capability to shut down the reactor and maintain it in a safe shutdown condition and control room operator

c. The capability to prevent or mitigate the consequences of accidents that could potentially result in offsite exposures approaching the guideline exposures of 10 CFR 100.

Items which are associated with safety-related equipment, but which in themselves are not absolutely essential to the safety function of the equipment, are not considered safety-related.

50.67 and Regulatory Guide 1.183, or 10 CFR 100, as applicable.

Power generation design bases support, either directly or indirectly, the major electrical power generation function of the station. Examples of power generation design bases 1.1-3 Rev. OL-17

ULNRC-06636 Page 9 of 374 TABLE 1.1-1 ACRONYMS USED IN THE FSAR AC Alternating Current ACI American Concrete Institute ACRS Advisory Committee on Reactor Safeguards A/E Architect/Engineer AFAS Auxiliary Feedwater Actuation System AFS Auxiliary Feedwater System AISC American Institute of Steel Construction ALARA As Low as Reasonably Achievable ANSI American National Standards Institute APRM Average Power Range Monitor ARM Area Radiation Monitor ARW Chemical Waste ASCE American Society of Civil Engineers ASME American Society of Mechanical Engineers ASTM American Society for Testing and Materials ATWS Anticipated Transients Without Scram AVT All Volatile Treatment AWS American Welding Society BOP Balance of Plant B&PVC Boiler and Pressure Vessel Codes BRS Boron Recycle System BTP Branch Technical Position CAS Compressed Air System CCS Condensate Cleanup System CCWS Component Cooling Water System CDS Condensate Demineralizer System CeCWS Central Chilled Water System CFR Code of Federal Regulations CFS Condensate and Feedwater System CGCS Combustible Gas Control System CHC Cask Handling Crane CHF Critical Heat Flux CIS Containment Isolation Signal ClCWS Closed Cooling Water System CLP Cask Loading Pit CM Center of Mass CMAA Crane Manufacturing Association of America CEDE Committed Effective Dose Equivalent

ULNRC-06636 Page 10 of 374 TABLE 1.1-1 (Sheet 2)

CP Construction Permit CPR Critical Power Ratio CPIS Containment Purge Isolation System/Signal CR Center of Rigidity CRD Control Rod Drive CRDA Control Rod Drop Accident CRDM Control Rod Drive Mechanism CREA Control Rod Ejection Accident CRVIS Control Room Ventilation Isolation System/Signal CRW Tritiated Waste CSD Cold Shutdown CST Condensate EDE Storage Effective Tank Dose Equivalent for cloudshine CSTS Condensate Storage and Transfer System CtCS Containment Cooling System CVCS Chemical and Volume Control System CWP Cask Washdown Pit CWS Circulating Water System DBA Design Basis Accident DBE Design Basis Event DC Direct Current DCSS Dry Cask Storage System DEHC Digital Electrohydraulic Control DEPSG Double Ended Pump Suction Guillotine DG Diesel Generator DGB Diesel Generator Building DoWS Domestic Water System DNB Departure From Nucleate Boiling DNBR Departure From Nucleate Boiling Ratio DRW Potentially Radioactive Nontritiated Waste DWMS Demineralized Water Make-up System DWST Demineralized Water Storage Tank DWSTS Demineralized Water Storage and Transfer System DWT Dead Weight Test ECCS Emergency Core Cooling System EHC Electrohydraulic Control EOL End of Life EDECAIES Emergency Diesel Engine Combustion Air Intake and Exhaust System EDECWS Emergency Diesel Engine Cooling Water System EDEFSTS Emergency Diesel Engine Fuel Oil Storage and Transfer System

ULNRC-06636 Page 11 of 374 TABLE 1.1-1 (Sheet 5)

PSAR Preliminary Safety Analysis Report PSS Process Sampling System PWR Pressurized Water Reactor RCP Reactor Coolant Pumps RCPB Reactor Coolant Pressure Boundary RCS Reactor Coolant System RHR Residual Heat Removal RMWCS Reactor Makeup Water Control System RMWS Reactor Makeup Water System RMWST Reactor Makeup Water Storage Tank RO Reactor Operator RPV Reactor Pressure Vessel RRS Required Response Spectrum RSG Replacement Steam Generator RWB Radwaste Building RWST Refueling Water Storage Tank SACF Single Active Component Failure SAR Safety Analysis Report SFSF Spent Fuel Storage Facility SGB Steam Generator Blowdown SGBIS Steam Generator Blowdown Isolation System/Signal SGBS Steam Generator Blowdown System SIS Safety Injection Signal SIT Structural Integrity Test SLWS Secondary Liquid Waste System SMA Strong Motion Accelerometer SNUPPS Standard Nuclear Unit Power Plant System SRO Senior Reactor Operator SRP Standard Review Plan SRS Solid Radwaste System SRSS Square Root of the Sum of the Squares SRW Detergent Waste SSE Safe Shutdown Earthquake SWS Service Water System TBS Turbine Bypass System TG Turbine Generator TGSS Turbine Gland Sealing System TRS Test Response Spectrum UHS Ultimate Heat Sink TEDE Total Effective Dose Equivalent

ULNRC-06636 Enclosure 6 Page 12 of 374 TABLE 1.3-4 (Sheet 14)

Regulation (10 CFR) Compliance 50.56 This regulation provides that the Commission will, in the absence of good cause shown to the contrary, issue an operating license upon completion of the construction of a facility in compliance with the terms and conditions of the construction permit. This imposes no independent obligations on the applicant.

50.57(a) This regulation requires the Commission to make certain findings prior to the issuance of an operating license.

50.57(b) The license, as issued, will contain appropriate conditions to ensure that items of contruction or modification are completed on a schedule acceptable to the Commission.

50.57(c) This regulation provides for a low-power testing license.

50.58 This regulation provides for the review and report of the Advisory Committee on Reactor Safeguards.

50.59 This regulation provides for the licensing of certain changes, tests, and experiments at a licensed facility. Technical Specifications and procedures provide implementation of this regulation.

50.70 The Commission has assigned resident inspectors to the SNUPPS plants and space will be provided in conformance with 50.70(b)(1) through (3).

50.71 Records are and will be maintained in accordance with the requirements of sections (a) through (e) of this regulation and the license.

50.80 This regulation provides that licenses may not be transferred without NRC consent. No application for transfer has been made by the SNUPPS utilities.

50.81 This regulation permits the creation of mortgages, pledges, and liens on licensed facilities, subject to certain provisions. The regulation prohibits secured creditors from violating the Atomic Energy Act and the Commission's regulations.

50.82 This regulation provides for the termination of licenses. It does not apply to SNUPPS' because no termination of licenses has been requested.

50.90 This regulation governs applications for amendments to licenses.

Future request for license amendments will be made in accordance with these requirements.

50.91 This regulation provides guidance to the NRC in issuing license amendments.

50.67 The FSAR accident analyses, in particular those in Chapter 6.0 and 15.0, demonstrate that offsite and control room doses resulting from postulated accidents would not exceed the criteria in this section of the regulation.

ULNRC-06636 Page 13 of 374 TABLE 1.3-4 (Sheet 26)

Regulation (10 CFR) Compliance 100.3 This regulation is explanatory and does not impose independent obligations on licensees.

100.10 The factors listed related to both the unit design and the site have been provided in the application. Site specifics, including seismology, meteorology, geology, and hydrology, are presented in Chapter 2.0 of the FSAR. The exclusion area, low population zone, and population center distance are provided and described. The FSAR also describes the characteristics of reactor design and operation.

100.11 Exclusion areas have been established, as described in each FSAR Site Addendum Section 2.1. The low population zone for each unit has been established in accordance with this requirement.

The FSAR accident analyses, particularly those in Chapters 6.0 and 15.0, demonstrate that offsite doses resulting from postulated accidents would not exceed the criteria in this section of the regulation.

Appendix A Appendix A to 10 CFR Part 100 provides seismic and geologic siting criteria for nuclear power plants. Site suitability was determined at the construction permit stage.

ULNRC-06636 Page 14 of 374 2.3 METEOROLOGY 2.3.4 SHORT-TERM (ACCIDENT) DIFFUSION ESTIMATES 2.3.4.1 Objective The objective of this section is to provide short-term atmospheric dispersion factors

(/Qs) for the postulated accident analyses presented in Chapter 15.0.

2.3.4.2 Calculations Section 2.3.4.4 2.3.4.2.1 Site Boundary and LPZ The short-term atmospheric dispersion factors (/Qs) are based on onsite meteorological data for the Callaway Plant site. The diffusion equations and assumptions used in the calculations were those outlined in NRC Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Assessment at Nuclear Power Plants." Table 2.3-1 lists the limiting /Qs for the Callaway site. The detailed procedures used in the calculations are given in Section 2.3.4.2 of the Site Addendum.

2.3.4.2.2 Control Room Intake 2.3.4.2.2.1 Waste Gas Decay Tank Rupture and Liquid Waste The basic model employed for the distribution of relative concentrations (/Qs) within a building wake at the Callaway control room intakes following an accident is given by Reference 1 to be:

K (1)

Q = -------

C AV-Where A = reference cross-sectional building area, m2 V = reference wind speed, m/sec KC = nondimensional concentration coefficient KC is a function of nondimensional space coordinates x/L, y/L, and z/L, building configuration, wind direction, and source configuration. The KC field for a given building configuration, source configuration, and wind configuration is considered to be invariant.

Accordingly, KC values determined by wind tunnel tests with a model structure are expected to be the same as those that would be obtained with a geometrically similar building in the full-scale atmosphere in the same wind direction, with a similar leak. The Callaway Plant contiguous building arrangement is shown in Figure 2.3-1. The KC data

ULNRC-06636 Page 15 of 374 used in the analysis for low level release are presented in Figure 2.3-2 and were derived from two sets of tests. One used rectangular prisms (Ref. 2), the other used a model of the EBR-II complex (Ref. 1). Both tests were described and portions of the data presented in Reference 3. The KC data for the unit vent release from the top of the containment were extracted from Figure 10 of Reference 1 and are presented in Table 2.3-2. The value of A used in conjunction with KC in Figure 2.3-2 and Table 2.3-2 is the Callaway Plant equivalent of the EBR-II area, A = 1.12 D2 = 2280 m2 with the diameter of the reactor D = 45.1 m.

The value of V used in conjunction with Figure 2.3-2 is the mean velocity of the approach flow at an elevation corresponding to the anemometer elevation of the EBR-II model tests. Reference 3 reports this elevation to be 62 feet or 0.77D above the top of the dome. The Callaway Plant equivalent height becomes 63.4 + 0.77 x 45.1 = 98.1m above ground. The V values were obtained by extrapolating wind speeds at anemometer elevations equivalent to 98.1 meters by the power law.

n (2)

V= u1(98.1 z1)

Where u1 = mean speed at elevation z1, m/sec z1 = anemometer elevation at a given site, m n = atmospheric stability exponent Values of n were arbitrarily assumed for the various stability classes as follows:

Pasquill Stability Class A B C D E F G n 0.20 0.25 0.29 0.33 0.40 0.50 0.60 Insert 2.3.4.2.2.2 A cumulative frequency distribution was constructed for the /Q values calculated by equations 1 and 2 above, using 3 years combined onsite meteorological data. The corresponding highest 5 percent, 10 percent, 20 percent, and 40 percent /Q values are given in Table 2.3-3.

2.

3.5 REFERENCES

1. Halitsky, J., Golden, J., Halpern, P., (1963): "Wind Tunnel Tests of Gas Diffusion From a Leak in the Shell of a Nuclear Power Reactor and from a Nearby Stack,"

N. Y. University Department of Met. & Ocean, GSL Rep. 63-2 under USWB Contract Cwb-10321

ULNRC-06636 Page 16 of 374 TABLE 2.3-1 LIMITING ATMOSPHERIC DISPERSION FACTOR, /Q(sec/m3) 2.05E-04 Site Boundary /Q 0-2 hr. 2.OE-4 Low Population Zone 0-8 hr. 2.6E-5 8-24 hr. 1.7E-5 24-96 hr. 7.2E-6 96-720 hr. 2.0E-6 3.42E-05 2.42E-05 1.13E-05 3.83E-06

ULNRC-06636 Page 17 of 374 3.0 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS This chapter identifies, describes, and discusses the principal architectural and engineering design features of those structures, components, equipment, and systems which are necessary to assure:

or control room

a. The integrity of the reactor coolant pressure boundary

b. The capability to shut down the reactor and maintain it in a safe shutdown condition

c. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline values of 10 CFR 100.

50.67, Regulatory Guide 1.183, or 10 CFR 100, as 3.1 CONFORMANCE WITH NRC GENERAL appropriate.

DESIGN CRITERIA This section briefly discusses the extent to which the design criteria for SNUPPS plant structures, systems, and components important to safety comply with Title 10, Code of Federal Regulations, Part 50 (10 CFR 50), Appendix A, "General Design Criteria for Nuclear Power Plants" (GDC). As presented in this section, each criterion is first quoted and then discussed in enough detail to demonstrate SNUPPS compliance with each criterion. For some criteria, additional information may be required for a complete discussion. In such cases, detailed evaluations of compliance with the various general design criteria are incorporated in more appropriate FSAR sections, but are located by reference.

3.1.1 DEFINITION OF SINGLE FAILURE The single failure criterion is a constraint used in the design of safety systems to improve the reliability of the system to perform its safety function following a design-basis event or design occurrence.

A single failure means an occurrence which results in the loss of the capability of a component to perform its intended safety functions. Multiple failures resulting from a single occurrence are considered to be a single failure. Fluid and electrical systems are considered to be designed against an assumed single failure if neither (1) a single failure of any active component (assuming that passive components function properly) nor (2) a single failure of a passive component (assuming that active components function properly) results in a loss of the capability of the system to perform its safety functions.

Single failures are random occurrences imposed upon safety systems that are required to respond to a design basis event. They are postulated despite the fact that the systems were designed to remain functional under the adverse condition imposed by the accident. No mechanism for the cause of the single failure need be postulated. Single

ULNRC-06636 Page 18 of 374 total effective dose CRITERION 19 - CONTROL ROOM equivalent (TEDE)

"A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent, to any part of the body, for the duration of the accident.

"Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures."

total effective dose DISCUSSION equivalent (TEDE)

A separate control room is provided for the control of each unit from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain in a safe manner under accident conditions, including LOCAs. Operator action outside of the control room to mitigate the consequences of an accident is permitted. The control room and its post-accident ventilation systems are designed to satisfy seismic Category I requirements, as discussed in Chapter 3.0. Adequate concrete shielding and radiation protection are provided against direct gamma radiation and inhalation doses postulated to result from a TID-14844 release of fission products inside the containment structure.

The shielding and the control room standby air-conditioning system allow access to and occupancy of the control rooms under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body or its equivalent to any part of the body for the duration of the accident. Refer to Chapter 15.0. Fission product removal is provided in the control room recirculation equipment to remove iodine and particulate matter, thereby minimizing the thyroid dose which could result from the accident. The control room habitability features are described in Chapter 6.0.

In the event that the operators are forced to abandon the control room, panel-mounted local instrumentation and controls are provided to achieve and maintain the plant in the hot shutdown condition (see Chapter 7.0). The capability for bringing the plant to a cold shutdown is also provided outside the control room through the use of local controls.

3.1.5 PROTECTION AND REACTIVITY CONTROL SYSTEMS CRITERION 20 - PROTECTION SYSTEM FUNCTIONS "The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational

ULNRC-06636 Page 19 of 374 offsite electric power system operation (assuming onsite power is not available) its safety function can be accomplished, assuming a single failure."

DISCUSSION The containment spray system serves to remove radioiodine and other airborne particulate fission products from the containment atmosphere following a LOCA. The system consists of two independent systems, each supplied from separate electrical power busses, as described in Chapter 8.0. Either subsystem alone can provide the fission product removal capacity for which credit is taken in Chapter 15.0, in compliance with Regulatory Guide 1.4. 1.183 The generation of hydrogen in the containment under post-accident conditions has been evaluated, using the assumptions of Regulatory Guide 1.7 (see Chapter 6.0). A post-accident hydrogen recombiner system is provided with redundancy of vital components so that a single failure does not prevent timely operation of the system.

This system is described in Section 6.2.5. A hydrogen purge system is provided as a backup. No single failure causes both subsystems to fail to operate.

CRITERION 42 - INSPECTION OF CONTAINMENT ATMOSPHERE CLEANUP SYSTEMS "The containment atmosphere cleanup systems shall be designed to permit appropriate periodic inspection of important components, such as filter frames, ducts, and piping to assure the integrity and capability of the systems."

DISCUSSION The containment atmosphere cleanup systems are designed and located so that they can be inspected periodically, as required. The essential equipment of the containment spray system is outside the containment, except for risers, distribution header piping, and spray nozzles in the containment. The hydrogen purge and monitoring components of the hydrogen control system are located outside the containment. The equipment outside the containment may be inspected during normal power operation. Components of the containment spray system and the hydrogen control system located inside the containment can be inspected during refueling shutdowns. See Chapter 6.0 for details on the containment spray system and details of the hydrogen control system.

CRITERION 43 - TESTING OF CONTAINMENT ATMOSPHERE CLEANUP SYSTEMS "The containment atmosphere cleanup systems shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the active components of the systems such as fans, filters, dampers, pumps, and valves and (3) the operability of the systems as a whole and, under conditions as close to design as practical, the

ULNRC-06636 Page 20 of 374 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS Certain structures, components, and systems of the nuclear plant are important to safety because they:

a. Assure the integrity of the reactor coolant pressure boundary.

or Control Room

b. Assure the capability to shut down the reactor and maintain it in a safe condition.

c. Assure the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline exposures of 10 CFR 100.

d. Contain or may contain radioactive material.

The purpose of this section is to classify 50.67, Regulatory structures, Guide 1.183,systems, and or 10 CFR components, 100, as appropriate.

according to the importance of the item, in order to provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public.

Table 3.2-1 delineates each of the items in the plant which fall under the above-mentioned categories and the respective associated classification that the NRC, ANS, and industrial codes committees have developed. Each of the classification categories in Table 3.2-1 is addressed in the following sections.

For identification of system and subsystem boundaries, Table 3.2-1 is supplemented (i.e., referenced to applicable figures) by piping and instrument diagrams which have been marked to clearly show the limits of the seismic Category I and various quality group classifications on a system. The legend for the piping and instrument diagrams is provided in Figure 1.1-1.

Classification of power supplies, instrumentation and controls, valve operators, supports, hangers, and restraints is not delineated in Table 3.2-1 because of the extensive listing required. Generic listings for piping/valves and ductwork/ dampers are included for completeness, since for some systems these are the only items serving a safety function. Containment penetrations are not included in these generic listings as there is a separate subheading for containment penetrations. The classification for all of these unlisted and generically listed items is consistent with the boundaries shown on the piping and instrumentation drawings. A listing of the piping and instrumentation drawings and their associated FSAR figures is found in Table 1.7-2 and in Section 1.7 of each Site Addendum.

ULNRC-06636 Page 21 of 374 3.6.2.3.2 Dynamic Analysis Methods to Verify Integrity and Operability for the Reactor Coolant Loop 3.6.2.3.2.1 General A LOCA is assumed to occur for a branch line break down to the restraint of the second normally open automatic isolation valve (Case II in Figure 3.6-2) on outgoing* and down to and including the second check valve (Case III in Figure 3.6-2) on incoming lines normally with flow. A pipe break beyond the restraint or second check valve will not result in an uncontrolled loss of reactor coolant if either of the two valves in the line closes.

Accordingly, both of the automatic isolation valves are suitably protected and restrained as close to the valves as possible so that a pipe break beyond the restraint will not jeopardize the integrity and operability of the valves. Further, periodic testing capability of the valves to perform their intended function is essential. This criterion takes credit for only one of the two valves performing its intended function. For normally closed isolation or incoming check valves (Cases I and IV in Figure 3.6-2), a LOCA is assumed to occur for pipe breaks on the reactor side of the valve.

Branch lines connected to the reactor coolant loop (RCL) are defined as "large" for the purpose of this criteria and as having an inside diameter greater than 4 inches up to the largest connecting line, generally the pressurizer surge line. Rupture of these lines results in a rapid blowdown from the RCL, and protection is basically provided by the accumulators and the low head safety injection pumps (residual heat removal pumps).

Branch lines connected to the RCL are defined as "small" if they have an inside diameter equal to or less than 4 inches. This size is such that emergency core cooling system analyses, using realistic assumptions, show that no clad damage is expected for a break area of up to 12.5 square inches, corresponding to 4-inch inside diameter piping.

Engineered safety features are provided for core cooling and boration, pressure reduction, and activity confinement in the event of a LOCA or steam or feedwater line break accident to ensure that the public is protected in accordance with 10 CFR 100 guidelines. These safety systems have been designed to provide protection for a reactor coolant system pipe rupture of a size up to and including a double-ended severance of a reactor coolant loop.

In order to assure the continued integrity of the vital components and the engineered safety systems, consideration is given to the consequential effects of the pipe break itself to the extent that:

50.67, Regulatory Guide 1.183, or 10 CFR 100 guidelines, as appropriate.

• It is assumed that motion of the unsupported line containing the isolation valves could cause failure of the operators of both valves to function.

ULNRC-06636 Page 22 of 374 3.7(N) SEISMIC DESIGN For the OBE loading condition, the nuclear steam supply system is designed to be capable of continued safe operation. The design for the SSE is intended to ensure:

a. That the integrity of the reactor coolant pressure boundary is not compromised;

b. That the capability to shut down the reactor and maintain it in a safe condition is not compromised; and

c. That the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline exposures of 10 CFR 100 is not compromised.

It is necessary to ensure that required critical structures and components do not lose their capability to perform their safety function. Not all critical components have the same functional safety requirements. For example, a safety injection pump must retain its capability to function normally during the SSE. Therefore, the deformation in the pump must be restricted to appropriate limits in order to ensure its ability to function. On the other hand, many components can experience significant permanent deformation without loss of function. Piping and vessels are examples of the latter where the principal requirement is that they retain their contents and allow fluid flow.

50.67, Regulatory Guide 1.183, or 10 CFR 100, as appropriate.

The seismic requirements for safety-related instrumentation and electrical equipment are covered in Sections 3.10(N) and (B). The safety class definitions, classification lists, operating condition categories, and the methods used for seismic qualification of mechanical equipment are given in Section 3.2.

3.7(N).1 SEISMIC INPUT 3.7(N).1.1 Design Response Spectra Refer to Section 3.7(B).1.1.

3.7(N).1.2 Design Time History Refer to Section 3.7(B).1.2.

3.7(N).1.3 Critical Damping Values The damping values given in Table 3.7(N)-1 are used for the systems analysis of Westinghouse equipment and for the component analysis of the Integrated Head Assembly (IHA) and replaced steam generators (SGs). These are consistent with the damping values recommended in Regulatory Guide 1.61, Rev. 0, except in the case of the primary coolant loop system components and large piping (excluding reactor 3.7(N)-1 Rev. OL-21

ULNRC-06636 Page 23 of 374 anchorage surfaces of the buttress are normal to the tangent line of the anchored hoop tendons. Details are shown in Figure 3.8-30.

The concrete shell around the equipment hatch opening is thickened by the method shown in Figures 3.8-31 and 3.8-32.

3.8.1.1.5 Special Reinforcing Requirements Special reinforcing is required in such areas as the major penetrations. Refer to Figures 3.8-31 through 3.8-35 for typical details in these areas.

3.8.1.2 Applicable Codes, Standards, and Specifications The following codes, regulations, standards, and specifications are utilized in the reactor building design.

3.8.1.2.1 Regulations

a. 10 CFR 50, "Licensing of Production and Utilization Facilities"

b. 10 CFR 100, "Reactor Site Criteria" 3.8.1.2.2 Codes c. 10 CFR 50.67, "Accident Source Term"

a. American Concrete Institute, Building Code Requirements for Reinforced Concrete (ACI-318-71)

b. American Institute of Steel Construction (AISC), Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings, 7th Edition, adopted February 12, 1969, and Supplement Numbers 1, 2, and 3 (See FSAR Table 3.2-1, Note 19)

c. ASME Boiler and Pressure Vessel Code - 1974 Edition or later Section II - Material SpecificationsSection III, Division 1 - Nuclear Power Plant ComponentsSection V - Nondestructive Examination Section VIII - Pressure VesselsSection IX - Welding and Brazing Qualifications

d. American Welding Society, Structural Welding Code (AWS D1.1-75) (See FSAR Table 3.2-1, Note 19) 3.8-4 Rev. OL-22

ULNRC-06636 Page 24 of 374 3.11(B).1.1.2 Safety-Related System Listing Safety-related systems are those plant systems necessary to ensure:

a. The integrity of the reactor coolant pressure boundary.

b. The capability to shut down the reactor and maintain it in a safely shutdown condition.

c. The capability to prevent or mitigate the consequences of accidents which could result in offsite exposures comparable to the guidelines of 10 CFR 100.

Systems that perform these type functions are those systems required to achieve or support emergency reactor shutdown, containment isolation, reactor core cooling, containment heat removal, core residual heat removal, and prevention of significant 50.67, release of radioactive Regulatory material Guide 1.183, The to the environment. or 10listing CFR of100, as appropriate.

systems that perform or support these functions is contained in the Callaway Equipment List (CEL). The specific safety function of each system is described in FSAR system description sections and in the CEL database.

Class 1E powered I&C devices are included in the system that they serve (e.g.,

EG-FT-0108 is a flow transmitter in the component cooling water system [EG]). The I&C devices can be divided into two categories, NSSS and BOP supplied. Each type can be identified in the fourth column of Table 3.11(B)-3. The BOP supplied devices that are purchased by the Bechtel I&C Group have a specification number that begins with the letter J (e.g., J-301 for EG-FT-0108). The NSSS-supplied devices are identified in the fourth column by the respective Westinghouse EQDP number (e.g., ESE-4).

3.11(B).1.2 Plant Environments 3.11(B).1.2.1 Normal Environments Pressure, Temperature, Humidity, and Radiation Normal operating environmental conditions are defined as conditions existing during routine plant operations. These environmental conditions, as listed in Table 3.11(B)-1, represent the normal maximum and minimum conditions expected during routine plant operations.

Dust In the NUREG-0588 review, dust was considered and was determined to be an insignificant factor in equipment qualification because outside air sources and ventilation units are typically equipped with filters which remove airborne dust. Also concrete

ULNRC-06636 Page 25 of 374 coating, plant housekeeping, dust seals, and equipment maintenance requirements provide assurance that dust will not degrade equipment performance.

3.11(B).1.2.2 Accident Environments - Inside Containment Accident environmental conditions are defined as those deviating from the normal operating environmental conditions. These conditions are specified in Table 3.11(B)-2.

In the NUREG-0588 review, Callaway LOCA/HELB/MSLB pressure, temperature, humidity, radiation, chemical spray, and submergence environmental conditions were evaluated. Where required, plant-unique environmental conditions were developed using the Category I criteria of NUREG-0588. The development of these conditions is described below. The post-accident parameters used in the equipment review are provided in summary form in Table 3.11(B)-2 and as used in the review, in Figures 3.11(B)-1 through 84. HELB P/T curves are also located in Reference 24.

Radiation Using the guidance of NUREG-0588, post-LOCA radiation environments were determined in all areas of the containment. The original fission product release data used in this analysis were obtained from Westinghouse. The isotopic inventory provided by Westinghouse was for an equilibrium cycle Callaway core. The data were calculated at the end of cycle life and, therefore, represent maximums suitable for post-accident evaluations. This source term is referred to as the licensing basis EQ source term, applicable It is notedto the thatinitial core load.

the offsite Subsequent and control room have cyclesseen changes in fuel type (from STD/LOPAR doses discussedtoinOFA to VANTAGE Section 15.6.5 were5 to VANTAGE+), power level (from 3425 MWt to 3579 calculated using the elemental iodine spray MWt), and burnup (up to 60,000 MWd/MTU a s discussed in Section 4.2.1). The doses removal model reported discussed in Table in Section 3.11(B)-4 6.5A.3.

have been increas ed by 5% to account for these effects. In addition, the airborne gamma doses were increased by another 3% to account for the replacement of the active spray additive system with a passive system of baskets adjacent to the containment recirculation sumps containing trisodium phosphate. The following discussion refers to the initial calculations performed with the licensing basis EQ source term and a 50% cesium release fraction.

The accident scenario assumed that a LOCA event occurred causing core damage. The entire source of 100 percent noble gas inventory, 50 percent of the core halogen inventory, 50 percent of the cesium, and 1 percent of the other solids was released to the containment. This release was conservatively assumed to occur at time zero. For the liquid source, 50 percent of the halogens, 50 percent of the cesium, and 1 percent of the remaining fission product solids were assumed to go directly to the sump and were diluted by the volume of the refueling water storage tank (RWST) and the liquid volume of the reactor coolant system. For the airborne source, 100 percent of the noble gases and 50 percent of core halogens were assumed to be released to the free volume of the containment. The simultaneous release of 50 percent of the halogens to the atmosphere and to the sump introduced additional conservatism.

ULNRC-06636 Page 26 of 374 Credit was taken for mechanistic removal of the airborne iodine via containment spray and plateout. The spray removal lambdas for elemental and particulate iodine (25.7 hr-1 + 0.73 hr-1) were taken from the calculated values listed in Table 6.5-2. The plate-out removal lambda (15.8 hr-1) was calculated using methodology outlined in NUREG/CR-0009. The surface area available for plateout was assumed to be equivalent to the heat sink area used in the containment pressure analysis given in Table 6.2.1-4. In addition, two of the four hydrogen mixing fans were assumed to be operating, at 42,500 cfm each, to provide mixing between the sprayed (86 percent) and unsprayed (14 percent) regions of the containment. These removal processes were assumed to persist until the elemental and particulate iodine in the sprayed region were reduced by factors of 200 and 10,000, respectively.

These decontamination factors (DFs) were taken from Reference 22. The spray removal rate for elemental iodine was calculated in Section 6.5A.2 to be 25.7 hr-1. This spray removal rate plus the plateout removal rate (25.7 hr-1 + 1.58 hr-1) were assumed to be effective in the sprayed region until an elemental iodine decontamination factor (DF) of 200 was reached in the EQ dose calculations. Only the plateout removal rate was assumed to be effective in the unsprayed region until an elemental iodine DF of 2 was reached in the EQ dose calculations. The spray removal rate for particulate iodine was calculated to be 0.73 hr-1 in Section 6.5A.1 and was assumed to be effective in the sprayed region until a particulate iodine DF of 10,000 was reached in the EQ dose calculations.

It is noted that the offsite and control room doses discussed in Section 15.6.5 were calculated using an elemental iodine spray removal rate of 10 hr-1 and a particulate iodine spray removal rate of 0.45 hr-1 , until a DF of 28.7 was reached for elemental species and a DF of 50 was reached for particulate species. No plateout removal lambda was used in the Section 15.6.5 dose calculations since credit was taken for the instantaneous plateout of half of the iodines released to the containment atmosphere (i.e. 25% of the core iodines).

With the replacement of the spray additive system with trisodium phosphate baskets, the minimum equilibrium sump fluid pH is reduced to 7.1. This reduced pH results in a reduced spray partition coefficient (H, from Equation 6.5A-15 on page 6.5A-7) of 1100 per Reference 22. Using Equation 6.5A-15, the resulting elemental iodine DF was calculated to be 28.7 for the analysis of offsite and control room doses discussed in Section 15.6.5. Per Reference 23, the particulate iodine spray removal rate, calculated using Equation 6.5A-1 on page 6.5A-2, can conservatively be based on an assumed E/D of 10 per meter initially, changing to 1 per meter after a DF of 50. After the particulate iodine spray removal rate is reduced, there is no DF limit. However, for simplicity and conservatism, removal was assumed to stop after a DF of 50 was reached in the analysis of offsite and control room doses. With consideration given to these reduced DF values for elemental and particulate iodines, airborne gamma doses listed in Table

ULNRC-06636 Page 27 of 374 and Control Room During and Following an Accident." The response has been included in Appendix 7A.

All Category I instruments are included in the NUREG-0588 program.

3.11(B).5.2 Equipment Operability 1.183 For the NUREG-0588 review, a post-DBA maximum operability requirement of 6 months (180 days) was utilized. Equipment was evaluated against this period for operability unless a shorter operability duration was justified. This value was selected as a conservative bounding time for termination of accident effects within the containment.

The containment pressure-temperature analysis, as reflected in Figures 3.11(B)-3 and 6, indicates that containment conditions return to normal or below normal operating conditions within 30 days. It should also be noted that Regulatory Guide 1.4 provides criteria for evaluating the offsite radiological consequences of a LOCA event for a maximum of 30 days following the accident.

Margins of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or more for equipment with required operability times of less than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> have generally been used for the Callaway equipment qualification review.

However, margins of less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> have been used when adequate technical justification could be provided. Union Electric concurs with the AIF position on the 1-hour time margin, as stated in a letter to Mr. Harold Denton dated January 4, 1982, in that an arbitrary time margin of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> appears inappropriate and should not be required when adequate technical justification for a shorter period exists.

3.11(B).5.3 Margins The discussions in Section 3.11(B).1 show that post-accident environmental parameters were conservatively and uniquely determined using plant-specific data. Hence, the guideline generic techniques discussed in NUREG-0588 are not applicable.

The values for margin identified in Section 6.3.1.5 of IEEE-323-1974 were used as acceptance criteria during the NUREG-0588 review. The only regular exception to the IEEE-323-1974 margins was for radiation. As identified in Item 1.4 of NUREG-0588, additional margin need not be added to the radiation parameters if the methods identified in Appendix D of NUREG-0588 are utilized. The methods used to determine the Callaway radiation parameters are consistent with the Appendix D methodology. Hence, the radiation margins required by Section 6.3.1.5 of IEEE-323-1974 were not necessary.

3.11(B).5.4 Aging During the NUREG-0588 review, two general observations were made concerning equipment aging:

1. Some IEEE-323-1974 equipment underwent accelerated thermal aging based on the Arrhenius method. This approach was considered acceptable.

3.11(B)-21 Rev. OL-25

ULNRC-06636 Page 28 of 374 APPENDIX 3A - CONFORMANCE TO NRC REGULATORY GUIDES This appendix briefly discusses the extent to which the standard plant conforms to NRC published regulatory guides, Division 1. The Standard Plant FSAR Appendix 3A may refer to the Addendum Appendix 3A or the Union Electric Company Operational Quality Assurance Manual (OQAM) for the specific regulatory commitment for certain regulatory guides. However, in cases where a reference is not made to the Addendum Appendix 3A or the OQAM, the commitment is as stated in the Standard Plant Appendix 3A and the same regulatory position is not repeated in the Addendum Appendix 3A or the OQAM. The statement of specific regulatory commitment for the following regulatory guides is located as indicated:

Callaway FSAR, Standard Plant - Regulatory Guides 1.1, 1.2, 1.3, 1.4, 1.5, 1.6, 1.7, 1.9, 1.10, 1.11, 1.12, 1.13, 1.14, 1.15, 1.18, 1.20, 1.22, 1.24, 1.25, 1.26, 1.29, 1.31, 1.32, 1.34, 1.35, 1.36, 1.40, 1.41, 1.42, 1.43, 1.44, 1.45, 1.46, 1.47, 1.48, 1.49, 1.50, 1.51, 1.52, 1.53, 1.54, 1.55, 1.56, 1.57, 1.59, 1.60, 1.61, 1.62, 1.63, 1.65, 1.66, 1.67, 1.68, 1.68.1, 1.68.2, 1.69, 1.70, 1.71, 1.72, 1.73, 1.75, 1.76, 1.77, 1.78, 1.79, 1.80, 1.81, 1.82, 1.83, 1.84, 1.85, 1.87, 1.89, 1.90, 1.92, 1.93, 1.95, 1.96, 1.97, 1.98, 1.99, 1.100, 1.101, 1.102*, 1.103, 1.104, 1.105, 1.106, 1.107, 1.108, 1.110, 1.112, 1.115, 1.117, 1.118, 1.119, 1.120, 1.121, 1.122, 1.124, 1.126, 1.128, 1.129, 1.130, 1.131, 1.133, 1.136, 1.137, 1.139, 1.140, 1.141, 1.142, 1.143, 1.147, 1.150, 1.152, 1.155, 1.158, 1.160, 1.163, 1.181, 1.182, 1.187, 1.195, and 1.205.

1.183 Callaway FSAR, Site Addendum - Regulatory Guides 1.17, 1.21, 1.23, 1.27, 1.59, 1.86, 1.91, 1.102*, 1.109, 1.111, 1.113, 1.114, 1.125, 1.127, 1.132, 1.134, 1.138, and 1.145.

Union Electric Operational Quality Assurance Manual - Regulatory Guides 1.8, 1.28, 1.30, 1.33, 1.37, 1.38, 1.39, 1.58, 1.64, 1.74, 1.88, 1.94, 1.116, 1.123, 1.144, and 1.146.

Exceptions to the guides are identified, and justification is presented or referenced. In the discussion of each guide, the sections or tables of the FSAR, where more detailed information is presented, are referenced. The referenced tables provide a position-by-position comparison to each regulatory position of section C of the regulatory guides. All statements within the Regulatory Position Section (C) of the Regulatory Guides are considered requirements unless a specific exception or clarification has been committed to by Union Electric. This is true regardless of the qualifier (i.e., "shall" or "should") which prefaces the statement. As regards to standards endorsed by the Regulatory Guide, unless further qualified within the Regulatory Guide, "shall" statements denote requirements while "should" statements denote recommendations.

• Refer to both the Callaway FSAR Standard Plant and the Callaway FSAR Site Addendum for the complete statement of regulatory commitment.

ULNRC-06636 Page 29 of 374 post-irradiation fracture toughness data have been obtained. Evaluation of the data obtained to date on material irradiated to fluences between 2.2 and 4.5 x 1019 n/cm2 indicates that the reference toughness curve, as contained in the American Society of Mechanical Engineers (ASME) Code,Section III, remains a conservative lower bound for toughness values for pressure vessel steels.

Details of progress and results obtained in the HSST program are available in the HSST program progress reports issued by Oak Ridge National Laboratory.

Regulatory Position C.2 is followed, inasmuch as no significant changes have been made in approved core or reactor designs.

Regulatory Position C.3 is followed, since the vessel design does not preclude the use of an engineering solution to assure adequate recovery of the fracture toughness properties of the vessel material. If additional margin is needed, the reactor vessel can be annealed at any point in its service life. This solution is already feasible, in principle, and could be performed with the vessel in place.

NOTE: Regulatory Guide 1.2 (Safety Guide 2) has been withdrawn by the NRC Staff letter to Regulatory Guide Distribution List, June 17, 1991. The guide has been superseded by 10 CFR 50, Section 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events. 10 CFR 50, Section 50.61 establishes screening criteria to effectively limit the extent of irradiation embrittlement permitted for reactor pressure vessel materials. The pressurized thermal shock requirements are sufficient to address thermal shock concerns. The withdrawal of Regulatory Guide 1.2 (Safety Guide 2) does not alter prior or existing licensing commitments based on its use.

REGULATORY GUIDE 1.3 REVISION 2 DATED 6/74 Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Boiling Water Reactors DISCUSSION:

The recommendations of this regulatory guide are not applicable to a PWR.

REGULATORY GUIDE 1.4 REVISION 2 DATED 6/74 Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors Use of this regulatory guide in Chapter 15 has been replaced by DISCUSSION: Regulatory Guide 1.183 for Alternative Source Term (AST) application The recommendations of this regulatory guide are met as described in Table 15.6-7.

ULNRC-06636 Page 30 of 374 DISCUSSION:

The recommendations of this regulatory guide are met as described in Table 15.7-1.

REGULATORY GUIDE 1.25 REVISION 0 DATED 3/72 Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors (Safety Guide 25)

Use of this regulatory guide in Chapter 15 has been replaced by DISCUSSION: Regulatory Guide 1.183 for alternative source term application The recommendations of this regulatory guide are met as described in Table 15.7-2.

REGULATORY GUIDE 1.26 REVISION 3 DATED 2/76 Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants DISCUSSION:

The recommendations of this regulatory guide are met as described in Table 3.2-4. As described in Section 3.2, Westinghouse utilizes the safety classes defined in ANSI N18.2a-1975.

REGULATORY GUIDE 1.27 Ultimate Heat Sink for Nuclear Power Plants DISCUSSION:

Refer to Appendix 3A in the Site Addendum.

REGULATORY GUIDE 1.28 Quality Assurance Program Requirements (Design and Construction)

DISCUSSION:

Refer to the Union Electric Company Operational Quality Assurance Manual.

REGULATORY GUIDE 1.29 REVISION 3 DATED 9/78 Seismic Design Classification DISCUSSION:

ULNRC-06636 Enclosure 6 Page 31 of 374 DISCUSSION:

UE complies with the requirements of this Regulatory Guide.

REGULATORY GUIDE 1.163 REVISION 0 DATED 9/95 Performance-Based Containment Leak-Test Program DISCUSSION:

UE complies with the recommendations of this Regulatory Guide as discussed in the Leak Rate Test Program (ESP-SM-1001).

REGULATORY GUIDE 1.181 Initial Issue DATED 9/99 Content of the Updated Final Safety Analysis Report in Accordance with 10 CFR 50.71(e) (Endorses NEI 98-03)

REGULATORY GUIDE 1.183 Initial Issue DATED 7/00 DISCUSSION:

Alternative radiological source terms for evaluating design basis accidents at nuclear power reactors AmerenUE will comply with the recommendations of this Regulatory Guide as needed to improve the FSAR. Major changes, (for example removal/archiving of information and DISCUSSION relocation of information) will be identified as exceptions to Regulatory Guide 1.70 as applicable. SeeofTable The recommendations 3A-1.

this regulatory guide are met as described in Chapter 15 REGULATORY GUIDE 1.182 Initial Issue DATED 5/00 Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants DISCUSSION:

AmerenUE complies with the requirements of this Regulatory Guide.

REGULATORY GUIDE 1.187 Initial Issue DATED 11/00 Guidance for Implementation of 10 CFR 50.59, Changes Tests and Experiments.

DISCUSSION:

AmerenUE complies with Regulatory Guide 1.187 with the following clarifications to NEI 96-07 Guidelines for 10 CFR 50.59 Implementation, dated November 2000:

1. With regard to Regulatory Position C.1 of Regulatory Guide 1.187, AmerenUE substitutes the word, "Implementation" for "Evaluations" to reflect title of NEI 96-07, dated November 2000.

ULNRC-06636 Enclosure 6 Page 32 of 374 Training and qualification of personnel performing Safety Analysis calculations will be accomplished in accordance with AmerenUE's Engineering Support Personnel training program.

4. 2.4 Comparison Calculations Comparison and benchmark calculations will be performed in accordance with approved procedural controls. Computer codes used for safety analysis will be controlled in accordance with AmerenUE's software control procedures.

5. 2.5 Quality Assurance and Change Control Safety Analysis calculations will be performed in accordance with the AmerenUE OQAP, which implements 10CFR50, Appendix B Criterion III. Computer codes used for safety analysis will be controlled in accordance with AmerenUE's software control procedures.

REGULATORY GUIDE 1.195 REVISION 0 DATED 5/03 Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors DISCUSSION:

The recommendations of this regulatory guide are met as described in the analysis of FSAR design basis accidents and their radiological consequences.

REGULATORY GUIDE 1.205 REVISION 01 DATED 12/2009 Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants DISCUSSION:

The recommendations of this regulatory guide are met. Refer to the FSAR Section 9.5.1.

REGULATORY GUIDE 1.194 Initial Issue DATED 6/03 Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants DISCUSSION The recommendations of this regulatory guide are met as described in Site Addendum 2, Section 2.3.4.4.

ULNRC-06636 Page 33 of 374 4.2 FUEL SYSTEM DESIGN The plant design conditions are divided into four categories in accordance with their anticipated frequency of occurrence and risk to the public: Condition I - Normal Operation; Condition II - Incidents of Moderate Frequency; Condition III - Infrequent Incidents; and Condition IV - Limiting Faults. Chapter 15.0 describes bases and plant operation and events involving each condition.

The reactor is designed so that its components meet the following performance and safety criteria:

a. The mechanical design of the reactor core components and their physical arrangement, together with corrective actions of the reactor control, protection, and emergency cooling systems (when applicable) ensure that:

1. Fuel damage* is not expected during Condition I and Condition II events. It is not possible, however, to preclude a very small number of rod failures. These are within the capability of the plant cleanup system and are consistent with plant design bases.**

2. The reactor can be brought to a safe state following a Condition III event with only a small fraction of fuel rods damaged** although sufficient fuel damage might occur to preclude immediate resumption of operation.

3. The reactor can be brought to a safe state and the core can be kept subcritical with acceptable heat transfer geometry following transients arising from Condition IV events.

b. The fuel assemblies are designed to withstand loads induced during shipping, handling, and core loading without exceeding the criteria of Section 4.2.1.5. Fuel assemblies can withstand loads introduced by a postulated reactor vessel head drop as evaluated in Section 9.1.4.3 for Westinghouse fuel.

c. The fuel assemblies are designed to accept control rod insertions in order to provide the required reactivity control for power operations and reactivity shutdown conditions (if in such core locations).

guidelines

• Fuel damage as used here is defined as penetration of the fission product barrier (i.e., the fuel rod cladding).

• • In any case, the fraction of fuel rods damaged must be limited so as to meet the dose guideline of 10 CFR 100.

50.67 or Regulatory Guide 1.183

ULNRC-06636 Page 34 of 374 4.3 NUCLEAR DESIGN 4.3.1 DESIGN BASES This section describes the design bases and functional requirements used in the nuclear design of the fuel and reactivity control system and relates these design bases to the General Design Criteria (GDC) presented in 10 CFR 50, Appendix A. Where applicable, supplemental criteria such as the "Final Acceptance Criteria for Emergency Core Cooling Systems" are addressed. Before discussing the nuclear design bases, it is appropriate to briefly review the four major categories ascribed to conditions of plant operation.

The full spectrum of plant conditions is divided into four categories, in accordance with the anticipated frequency of occurrence and risk to the public:

a. Condition I Chapter - Normal 15.0.9 Operation addresses the Fuel System Nuclear Design

b. Condition II Bases

- Incidents that affectof Moderate the Frequency fuel fission product inventory used in

c. Condition III the- Accident DoseFaults Infrequent Assessment.

d. Condition IV - Limiting Faults In general, the Condition I occurrences are accommodated with margin between any plant parameter and the value of that parameter which would require either automatic or manual protective action. Condition II incidents are accommodated with, at most, a shutdown of the reactor with the plant capable of returning to operation after corrective action. Fuel damage (fuel damage as used here is defined as penetration of the fission product barrier, i.e., the fuel rod cladding) is not expected during Condition I and Condition II events. It is not possible, however, to preclude a very small number of rod failures. These are within the capability of the CVCS and are consistent with the plant design basis.

Condition III incidents do not cause more than a small fraction of the fuel elements in the reactor to be damaged, although sufficient fuel element damage might occur to preclude immediate resumption of operation. The release of radioactive material due to Condition III incidents is not sufficient to interrupt or restrict public use of these areas beyond the exclusion radius. Furthermore, a Condition III incident does not by itself generate a Condition IV fault or result in a consequential loss of function of the reactor coolant or reactor containment barriers. 50.67 or Regulatory Guide 1.183 Condition IV occurrences are faults that are not expected to occur but are defined as limiting faults which must be designed against. Condition IV faults do not cause a release of radioactive material that results in exceeding the limits of 10 CFR 100.

The core design power distribution limits related to fuel integrity are met for Condition I occurrences through conservative design and maintained by the action of the control system. The requirements for Condition II occurrences are met by providing an adequate protection system which monitors reactor parameters. The control and

ULNRC-06636 Enclosure 6 Page 35 of 374 TABLE OF CONTENTS (Continued)

Section Page 6.5.2 CONTAINMENT SPRAY SYSTEM .................................................................. 6.5-3 6.5.2.1 Design Bases ................................................................................................. 6.5-4 6.5.2.1.1 Safety Design Bases ............................................................................... 6.5-4 6.5.2.1.2 Power Generation Design Basis ............................................................ 6.5-4 6.5.2.2 System Design ............................................................................................... 6.5-4 6.5.2.2.1 General Description................................................................................. 6.5-4 6.5.2.2.2 Component Description .......................................................................... 6.5-5 6.5.2.2.3 System Operation .................................................................................... 6.5-5 6.5.2.3 Safety Evaluation .......................................................................................... 6.5-6 6.5.2.4 Tests and Inspections................................................................................... 6.5-8 6.5.2.5 Instrumentation Requirements .................................................................... 6.5-8 6.5.2.6 Materials ......................................................................................................... 6.5-8 6.5.3 FISSION PRODUCT CONTROL SYSTEMS .................................................. 6.5-9 6.5.3.1 Primary Containment .................................................................................... 6.5-9 6.5.3.26.5A-4 PARTICULATE MODEL FOR OFFSITE AND Secondary Containment.......................................... ........................... 6.5-10 CONTROL ROOM DOSE CALCULATIONS 6.5.4 ICE CONDENSER AS A FISSION PRODUCT CLEANUP SYSTEM ....... 6.5-10 6.

5.5 REFERENCES

.................................................................................................. 6.5-10 FOR EQ DOSE CONSEQUENCES App. 6.5A IODINE REMOVAL MODELS FOR THE CONTAINMENT SPRAY SYSTEM ................................................... 6.5A-1 6.5A.1 PARTICULATE IODINE MODEL............................................................. 6.5A-2 6.5A.2 ELEMENTAL IODINE MODEL FOR EQ DOSE CALCULATIONS.......... 6.5A-3 6.5 A.3 ELEMENTAL IODINE MODEL FOR OFFSITE AND CONTROL ROOM DOSE CALCULATIONS ........................................................................ 6.5A-10 6.5A.4 REFERENCES ...................................................................................... 6.5A-11 5

6.6 INSERVICE INSPECTION OF CLASS 2 AND 3 COMPONENTS .................... 6.6-1 6.6.1 COMPONENTS SUBJECT TO INSPECTION ............................................... 6.6-1 6.6.2 ACCESSIBILITY .................................................................................................. 6.6-1 6.6.3 EXAMINATION TECHNIQUES AND PROCEDURES .................................. 6.6-2 6.0-vi

ULNRC-06636 Page 36 of 374 CHAPTER 6.0 ENGINEERED SAFETY FEATURES Engineered safety features (ESF) are those safety-related systems and components designed to directly mitigate the consequences of a design basis accident by:

a. Protecting the fuel cladding and Control Room

b. Ensuring the containment integrity

c. Limiting fission product releases to the environment within the guideline values of 10 CFR, Part 100 50.67 and Regulatory Guide 1.183 The limiting design basis accidents which are discussed and analyzed in C hapter 15.0 and Section 6.3 are:

a. Loss-of-coolant accident (LOCA)

b. Main steam line break (MSLB)

c. Steam generator tube rupture

d. Fuel handling accident The engineered safety features consist of the following systems:

a. Containment (Section 6.2.1)

b. Containment heat removal (Section 6.2.2)

c. Containment isolation (Sections 6.2.4 and 6.2.6)

d. Containment combustible gas control (Section 6.2.5)

e. Emergency core cooling (Section 6.3)

f. Fission product removal and control systems (Section 6.5)

g. Emergency HVAC and filtration (Section 9.4)

h. Control room habitability (Section 6.4)

i. Auxiliary feedwater (Section 10.4.9)

The containment is provided to contain radioactivity following a LOCA.

ULNRC-06636 Page 37 of 374 6.2 CONTAINMENT SYSTEMS The containment systems include the containment, the containment heat removal systems, the containment isolation system, and the containment combustible gas control system.

The design basis accident (DBA) is defined as the most severe of a spectrum of hypothetical loss-of-coolant accidents (LOCA). The ability of the containment systems to mitigate the consequences of a DBA depends upon the high reliability of these systems.

This section provides the design criteria and evaluations to demonstrate that these systems function within the specified limits throughout the unit operating lifetime.

6.2.1 CONTAINMENT FUNCTIONAL DESIGN A physical description of the containment and the design criteria relating to construction techniques, static loads, and seismic loads is provided in Section 3.8. This section pertains to those aspects of containment design, testing, and evaluation that relate to the accident mitigation function.

6.2.1.1 Containment Structure 6.2.1.1.1 Design Bases The safety design basis for the containment is that the containment must withstand the pressures and temperatures of the DBA without exceeding the design leakage rate, as required by 10 CFR 50, Appendix A, General Design Criterion 50, and that, in conjunction with the other containment systems and the other engineered safety features, the release of radioactive material subsequent to a DBA does not result in doses in excess of the guideline values specified in 10 CFR 100. The radiological consequences of the DBA are presented in Section 15.6.

50.67 and Regulatory

a. Assumed Accident Conditions Guide 1.183 For the purpose of determining the design pressure requirements for the containment structure and the containment internal structures, the following simultaneous occurrences are assumed:

1. The postulated reactor coolant system pipe rupture, as listed in Table 6.2.1-1, is assumed to be concurrent with the loss of offsite power and the worst single active failure. No two pipe breaks are assumed to occur simultaneously or consecutively. For design loadings on the systems used to mitigate the consequences of a postulated reactor coolant system pipe rupture, a safe shutdown earthquake is assumed.

ULNRC-06636 Page 38 of 374 The component cooling water system is described in Section 9.2.2, the essential service water system is described in Section 9.2.1, and the ultimate heat sink is described in Section 9.2.5.

Single failures in systems which remove energy from the containment are considered to be consistent with the single failures assumed in the development of the mass and energy release data. The energy removal capability of the containment air coolers, the containment spray system, and the residual heat removal system consider the parameters provided in Table 6.2.1-3.

50.67 and Regulatory Guide 1.183

f. Bases for Containment Depressurization Rate To meet the containment safety design basis of limiting the release of radioactive material subsequent to a DBA so that the doses are within the guideline values specified in 10 CFR 100, the containment pressure is reduced to less than 50 percent of the containment design pressure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after an accident. Chapter 15.0 contains the assumptions used in the analysis of the offsite radiological consequences of the accident.

g. Bases for Minimum Containment Pressure Used in ECCS Performance Studies The minimum containment pressure transient used in the analysis of the emergency core cooling system's capability is based on the conservative overestimated heat removal capability and pressure reduction capability of the containment structures and the containment systems and on the conservative reactor coolant system thermal analysis provided in Section 15.6. The determination and evaluation of the minimum containment pressure transient are provided in Section 6.2.1.5.

6.2.1.1.2 Design Features The principal containment and containment subcompartment design parameters are provided in Table 6.2.1-2. General arrangement drawings for the reactor containment are provided in Figures 1.2-9 through 1.2-18. Simplified arrangement drawings illustrating the nodalization model used for the containment subcompartment analyses are provided in Figures 6.2.1-43 through 6.2.1-46, 6.2.1-51 through 6.2.1-55, and 6.2.1-76.

a. Missile and Pipe Whip Protection Missile shield considerations are described in Section 3.5. The structural design of the containment and the containment subcompartments is discussed in Section 3.8. The designed structural strength considers the effects of pipe whip and jet forces, as discussed in Section 3.6.

ULNRC-06636 Page 39 of 374 The analysis has shown that even assuming a stuck RCCA with or without offsite power, and assuming a single failure in the engineered safeguards, the core remains in place and intact. Radiation doses will not exceed 10 CFR 100 guidelines.

DECREASE IN HEAT REMOVED BY THE SECONDARY SYSTEM Feedwater System Pipe Break 50.67 and Regulatory Guide 1.183 A major feedwater line rupture is defined as a break in a feedwater line large enough to prevent the addition of sufficient feedwater to the steam generators to maintain shell side fluid inventory in the steam generators. If the break is postulated in a feedwater line between the check valve and the steam generator, fluid from the steam generator may also be discharged through the break. Further, a break in this location could preclude the subsequent addition of auxiliary feedwater to the affected steam generator. (A break upstream of the feedwater line check valve would affect the NSSS only as a loss of feedwater. This case is covered by the evaluation in Sections 15.2.6 and 15.2.7).

Depending upon the size of the break and the plant operating conditions at the time of the break, the break could cause either an RCS cooldown (by excessive energy discharge through the break) or an RCS heatup. Potential RCS cooldown resulting from a secondary pipe rupture is evaluated in Section 15.1.5. Therefore, only the RCS heatup effects are evaluated for a feedwater line rupture.

A feedwater line rupture reduces the ability to remove heat generated by the core from the RCS for the following reasons:

a. Feedwater flow to the steam generators is reduced. Since feedwater is subcooled, its loss may cause reactor coolant temperatures to increase prior to reactor trip.

b. Fluid in the steam generator may be discharged through the break, and would then not be available for decay heat removal after trip.

c. The break may be large enough to prevent the addition of any main feedwater after trip.

An auxiliary feedwater system functions to ensure the availability of adequate feedwater so that:

a. No substantial overpressurization of the RCS occurs (less than 110 percent of design pressures); and

b. Sufficient liquid in the RCS is maintained so that the core remains inplace and geometrically intact with no loss of core cooling capability.

ULNRC-06636 Page 40 of 374 operators to achieve and/or maintain the plant in a safe shutdown condition. The following safety design bases are met:

SAFETY DESIGN BASIS ONE - The habitability systems are housed within a structure capable of withstanding the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles (GDC-2).

SAFETY DESIGN BASIS TWO - The habitability systems are designed to remain functional after an SSE and to perform their intended function following a postulated hazard, such as a fire, internal missiles, or pipe break (GDC-3 and 4).

SAFETY DESIGN BASIS THREE - Habitability system redundancy is provided so that safety functions can be performed, assuming a single active component failure coincident with a loss of offsite power.

SAFETY DESIGN BASIS FOUR - The habitability systems are designed so that the active components are capable of being tested during plant operation. Provisions are made to allow for inservice inspection of appropriate components of the control room air-conditioning system.

SAFETY DESIGN BASIS FIVE - The habitability systems are designed and fabricated according to codes consistent with the quality group classification assigned by Regulatory Guide 1.26 and the seismic category assigned by Regulatory Guide 1.29.

The power supply and control functions are in accordance with Regulatory Guide 1.32.

SAFETY DESIGN BASIS SIX - The capability to isolate all nonsafety-related HVAC system penetrations of the control building boundary is provided, if required, so that the occupation and habitability of the control room will not be compromised.

SAFETY DESIGN BASIS SEVEN - The radiation exposure of control room personnel throughout the duration of any one of the postulated DBAs discussed in Chapter 15.0 does not exceed the guideline values of GDC-19.

SAFETY DESIGN BASIS EIGHT - Throughout the duration of any one of the postulated hazardous chemical releases discussed in Section 2.2 of the Site Addendum or DBAs discussed in Chapter 15.0 of the FSAR, the habitability systems maintain the control room atmosphere at environmental conditions suitable for occupancy per GDC-19. The habitability systems comply with Regulatory Guides 1.78 and 1.95.

SAFETY DESIGN BASIS NINE - The control room ventilation system is capable of automatic transfer from its normal operational mode to its emergency mode upon detection of airborne radiation resulting in exposure of control room personnel in excess of GDC-19 limits.

order to prevent an event or the release of

ULNRC-06636 Page 41 of 374 SAFETY EVALUATION SIX - Section 9.4.1.2.3 describes the provisions made to assure the isolation of the control room.

SAFETY EVALUATION SEVEN - The direct radiation exposure rate of a control room occupant throughout the duration of any one of the postulated DBAs discussed in Chapter 15.0 does not exceed 0.5 mr/hr whole-body, and thus will not exceed GDC-19 requirements. A detailed discussion of the dose calculation model for control room operators is discussed in Appendix 15A. Control room shielding design, based on the most limiting design basis LOCA fission product release, is discussed in Section 12.3.

SAFETY EVALUATION EIG HT will- not exceed requirements.

Throughout the duration ofInclusion of postu any of the the lated resulting contribution hazardous chemical releasesdiscussed withor in Section 2.2 the Control DBAs Room i n Chapter discussed operator inhalation, 15.0, the habitability system maintains the control immersion, and transit room environmental condi tions below dose those established by Regul atory Guides 1.78 and 1.95 and GDC-19. Compliance with Requlatory Guides 1.78 and 1.95 is provided in Tables 6.4-1 and 6.4-2, respectively.

SAFETY EVALUATION NINE - Upon detection of high radiation in the induction trunk, the control room ventilation system is capable of automatic transfer from normal to emergency mode to minimize the exposure of control room personnel.

6.4.5 TESTS AND INSPECTIONS Testing and inspection of control room HVAC systems are described in Section 9.4.1.4.

The emergency mode of the control room HVAC system will undergo an acceptance test to verify that the system will maintain a 1/8-inch w.g. positive pressure in the emergency zone. Testing complies with Regulatory Guide 1.95, as described in Table 6.4-2.

The control room is classified as Type B per Regulatory Guide 1.78. Since the air exchange rate exceeds 0.06 air exchanges per hour for the control room, periodic testing of the control room pressurization system is not required per the exclusion provisions of Regulatory Guides 1.78 and 1.95. This periodic testing is not required for the Callaway plant based on the adequacy of a 400 cfm (nominal with tolerance of (+) 40 cfm, (-) 40 cfm) pressurization flow rate (Reference 1).

6.4.6 INSTRUMENTATION REQUIREMENTS Safety-related instrumentation and isolation signals are discussed in Sections 9.4.1.2.3 and 7.3.

Indication of all fan operational status is provided in the control room.

An indication of the position of all isolation dampers is provided in the control room.

All instrumentation associated with filtration units complies with Regulatory Guide 1.52, as described in Table 9.4-2.

ULNRC-06636 Page 42 of 374 10 CFR 100. The safety evaluations which demonstrate the design and construction of the ESF filtration systems are provided in Sections 9.4.2 and 9.4.3.

SAFETY EVALUATION TWO - The results of the analyses described in Chapter 15.0 demonstrate that the50.67 and control Regulatory building HVACGuide 1.183 systems reduce and control fission product release to the control room following a LOCA, such that radiation exposures of control room personnel are within the requirements of GDC-19. The safety evaluations which demonstrate the design and construction of these control building HVAC systems are provided in Sections 9.4.1 and 6.4.

6.5.1.4 Tests and Inspections Tests and inspections for ESF filter systems are described in Sections 9.4.1.4, 9.4.2.4, and 9.4.3.4.

6.5.1.5 Instrumentation Requirements Instrumentation and controls are provided to facilitate automatic operation and remote control of the system and to provide continuous indication of system parameters.

Further descriptions are provided in Sections 9.4.1.5, 9.4.2.5, and 9.4.3.5.

6.5.1.6 Materials The materials used for ESF filtration systems were chosen considering the environmental conditions and are commensurate with acceptable construction practices.

6.5.2 CONTAINMENT SPRAY SYSTEM The containment spray system (CSS) is an ESF, the functions of which are to reduce pressure and temperature in the containment atmosphere following a postulated LOCA or MSLB inside containment and to remove radioactive fission products from the containment atmosphere. These functions are performed by spraying a chemical solution into the containment atmosphere through a large number of nozzles on spray headers located in the containment dome. Reduction of pressure and temperature in the containment with the CSS is discussed in Section 6.2.2.1.

Radioiodine in its various forms is the fission product of primary concern in the evaluation of a LOCA. It is absorbed by the containment spray from the containment atmosphere.

To enhance this iodine absorption capacity of the spray, the spray solution is adjusted to an alkaline pH which promotes iodine hydrolysis, in which iodine is converted to nonvolatile forms tending to plate out on containment structures or to be retained in the containment recirculation sumps.

The physical characteristics of the CSS are discussed in Section 6.2.2.1. Discussed herein is the containment spray system's fission product removal capability following a LOCA.

ULNRC-06636 Page 43 of 374 6.5.2.1 Design Bases 6.5.2.1.1 Safety Design Bases SAFETY DESIGN BASIS ONE - The CSS is designed to provide an equilibrium sump solution pH of greater than or equal to 7.1 following the complete dissolution of the trisodium phosphate stored in baskets adjacent to the containment recirculation sumps.

SAFETY DESIGN BASIS TWO - The CSS is capable of reducing the iodine and particulate fission product inventories in the containment atmosphere such that the offsite radiation exposures resulting from a design basis LOCA are within the plant siting dose guidelines of 10 CFR 100.

50.67 Additional safety design bases are included in Section 6.2.2.1, in which the capability of the spray system to remove heat from the containment atmosphere is discussed.

6.5.2.1.2 Power Generation Design Basis The CSS has no power generation design basis.

6.5.2.2 System Design 6.5.2.2.1 General Description The spray additive tank has been retired in place and associated lines have been capped, as shown schematically in Figure 6.2.2-1.

Initially, water from the refueling water storage tank (RWST) is used for containment spraying followed by water from the containment recirculation sumps.

Those parts of the system in contact with containment spray fluids, are stainless steel or an equivalent corrosion-resistant material.

The trisodium phosphate (TSP-C) baskets constructed of stainless steel mounted to carbon steel supports contain sufficient TSP-C to bring the equilibrium sump fluid to a minimum pH of 7.1 upon mixing with the borated water from the refueling water storage tank, the accumulators, and reactor coolant. This assures continued iodine retention effectiveness of the sump water during the recirculation phase.

The spray header design, including the number of nozzles per header, nozzle spacing, and nozzle orientation, is provided in Section 6.2.2.1 and shown in Figures 6.2.2-2 and 6.2.2-4. Each spray header layout is oriented to provide more than 90-percent area coverage at the operating deck of the reactor building.

Total containment free volume, unsprayed containment free volume, specific unsprayed regions and volumes, and post-accident ventilation between sprayed and unsprayed

ULNRC-06636 Page 44 of 374 On actuation, approximately 5 percent of each spray pump's discharge flow is recirculated.

When the refueling water storage tank has reached its specified low-low-2 level limit, recirculation spray flow is manually initiated. The operator can remotely initiate recirculation flow by use of either or both of the spray pumps. Sections 6.2.2.1.5 and 6.5.2.5 address the instrumentation and information displays available to the operator, in order for manual switchover of the CSS to take place.

System flow rates and the duration of operational modes are presented in Section 6.2.2.1.2.3.

Design operation of the CSS is such that LOCA iodine removal requirements are fulfilled during the injection phase and the amount of TSP-C provided is sufficient to ensure long-term iodine retention. Following a large break LOCA, the containment spray during the injection phase will be a boric acid solution having a pH of about 4.5. The desired pH level is greater than 7.0 to assure iodine retention in the sumps, to limit corrosion and the associated production of hydrogen, and to limit chloride induced stress-corrosion cracking of austenitic stainless steels. To adjust the sump solution pH into the desired range, a minimum of 9000 pounds of trisodium phosphate dodecahydrate (NA3PO4

• 12 H2O

• 1/4 NaOH) is stored in two baskets, one adjacent to each containment recirculation sump, at an elevation to assure TSP-C disolution. This amount of trisodium phosphate is sufficient to assure that the equilibrium sump solution pH will be greater than or equal to 7.1. The containment iodine removal credit assumed in the calculation of offsite doses following a LOCA is provided in Table 15.6-6.

6.5.2.3 Safety Evaluation The safety evaluations are numbered to correspond to the safety design bases.

SAFETY EVALUATION ONE - The system's capability to reduce the airborne fission product inventory is based on the surface area of the spray solution for removal during injection and on sump solution pH for retention during recirculation, and on the system's capability to provide spray for essentially all regions of the containment, considering post-accident conditions.

During injection, the effectiveness of the spray against elemental iodine vapor is chiefly determined by the rate at which fresh solution surface area is introduced into the containment atmosphere, as discussed in Reference 3. The first-order spray removal coefficient calculated per Reference 3, as discussed in Section 6.5A.3, is 37 hr-1. Thus, the elemental iodine removal coefficient of 10 hr-1 used in Section 15.6.5 is conservative.

The minimum equilibrium sump pH of 7.1 assures iodine retention in the recirculated spray liquid.

4 20

ULNRC-06636 Page 45 of 374 CALLAWAY - SP at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> The system is designed to provide a spray solution during the recirculation phase with a minimum equilibrium pH of 7.1. The mass of TSP-C in the baskets results in this minimum pH level in the sumps.

The worst case concentration during the injection phase would be greater than or equal to 4.0 but less than 7.0 when water from the refueling water storage tank is sprayed directly to the containment. The injection phase is the only time that this pH = 4.0 condition could exist. The injection phase is short (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) relative to the entire spray duration (approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />). During the spray recirculation phase, the equilibrium pH range is 7.1-8.1. This spray is directed through the same spray headers and, therefore, should rinse all of the previously sprayed components (for a period of approximately 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />). the sump pH increases to and remains higher than the The minimum equilibrium sump pH of7.1 7.1equilibrium is based onvalue.

the Technical Specification minimum of 9000 lbm of TSP-C in the baskets and the maximum sump solution boric acid concentration of 2500 ppm boron. With the Technical Specification maximum of 14,300 lbm of TSP-C in the baskets and the minimum sump solution boric acid concentration of 1971 ppm boron, the maximum equilibrium sump pH would be less than 8.1. approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> The previously evaluated upper bound for containment spray pH of 11.0 will continue to be cited, consistent with Section 3.11(B).1.2.2, for the purpose of performing EQ reviews.

Another issue that has been reviewed is the unlikely, but possible, event in which an initially concentrated solution of TSP-C occupies the stagnant volume of an inoperable sump. This situation would not last for long since, as the recirculated sump fluid is cooled in the RHR heat exchangers, sufficient buoyancy-driven circulation within containment will result to displace the stagnant solution and eventually yield a uniform, equilibrium solution.

SAFETY EVALUATION TWO - The spray iodine removal analysis is based on the assumptions that:

a. Only one out of two spray pumps is operating

b. The ECCS is operating at its maximum capacity The spray system is assumed to spray approximately 85 percent of the total containment net free volume. This volume consists of those areas directly sprayed plus those volumes which have good communication with the directly sprayed volumes. The remaining 15 percent of the containment free volume has restricted communication with the sprayed volumes and is assumed to be unsprayed. A description of the unsprayed volumes is presented in Table 6.5-2.

The performance of the spray system was evaluated at the containment post-LOCA calculated saturation temperature corresponding to the calculated peak pressures and

ULNRC-06636 Enclosure 6 Page 46 of 374 3,086 CALLAWAY - SP containment design pressure provided in Table 6.2.1-2. The net spray flow rate of 3,131 gpm (see Table 6.5-2) per train was used in the calculations described in Appendix 6.5A. (and 0.646 hr-1 thereafter) 1.183 Based on Regulatory Guide 1.4, three species of airborne iodine are postulated to exist in the containment atmosphere following a LOCA. These are elemental, particulate, and organic species.

200 6.46 It has been assumed in these evaluations of spray removal effectiveness that organic iodine

,which formsboth bounds are injection and by the containment spray. A limited credit for the removal not removed 20 of airborne particulates and elemental iodine has been taken in the offsite and control recirculation spray design flow rates, room dose calculation, assuming that the spray removal rate is 0.45 hr-1 until a decontamination factor of 50 is attained for particulates and that spray removal rate is 10/

hr-1 until a decontamination factor (DF) of 28.7 is attained for elemental iodine. These assumptions underestimate the actual amounts of iodine removed and thus result in calculated accident doses higher than could realistically be expected.

Utilizing the dose analysis input parameters indicated above, in Table 6.5-2, and in Appendix 15A, the dose analysis of Section 15.6.5 demonstrates that offsite radiation exposures resulting from a design basis LOCA are within the plant siting dose guidelines of 10 CFR 100.

50.67 Appendix 6.5A provides the model used to calculate the iodine removal coefficients provided in Table 6.5-2.

6.5.2.4 Tests and Inspections CSS tests and inspections are discussed in Section 6.2.2.1.4, including spray nozzle tests and inspections.

6.5.2.5 Instrumentation Requirements Containment spray instrumentation is discussed in Section 6.2.2.1.5.

6.5.2.6 Materials The chemical compositions of the containment spray fluid entering the spray header during the injection phase of containment spray and the containment spray fluid in the system during the recirculation phase of containment spray (containment recirculation sump solution) are provided in Table 6.5-5.

None of the materials used is subject to decomposition by the radiation or thermal environment.

ULNRC-06636 Page 47 of 374 The corrosion of materials in the NSSS and the containment building, resulting from the spray solution used for iodine absorption, has been tested by the Reactor Division at ORNL (Ref. 2). The spray solutions provided in Table 6.5-5 result in negligible corrosion, based on these studies.

TSP-C does not undergo radiolytic decomposition in the post-LOCA environment.

Sodium has a low neutron absorption cross section and will not undergo significant activation.

With respect to the potential for decomposition, TSP-C is stable to at least 158°F.

Temperatures 158°F may result in the loss of H2O from the TSP-C but will not affect its caustic properties.

6.5.3 FISSION PRODUCT CONTROL SYSTEMS 6.5.3.1 Primary Containment 50.67 or Regulatory Guide 1.183 The containment consists of a prestressed post-tensioned, reinforced concrete structure with cylindrical walls, hemispherical dome, and base slab lined with a welded quarter-inch carbon steel liner plate, which forms a continuous, leaktight membrane.

Details of the containment structural design are discussed in Section 3.8. Layout drawings of the containment structure and the related items are given in the general arrangement drawings of Section 1.2.

The containment walls, liner plate, penetrations, and isolation valves function to limit the release of radioactive materials, subsequent to postulated accidents, such that the resulting offsite doses are less than the guideline values of 10 CFR 100. Containment parameters affecting fission product release accident analyses are given in Appendix 15A. sometime after 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Long-term containment pressure response to the design basis LOCA is shown in Figure 6.2.1-1. Relative to this time period, the CSS is operated to reduce iodine concentrations and containment atmospheric temperature and pressure commencing with system initiation, at approximately 60 seconds, as shown in Table 6.2.2-3 and ending when containment pressure has returned to normal. For the purpose of post-LOCA dose calculations discussed in Chapter 15.0, two dose models have been assumed, the 0-2 hour case and the 0-30 day case, as shown in Appendix 15A.

The containment minipurge system may be operated for personnel access to the containment when the reactor is at power, as discussed in Section 9.4.6.

Redundant, safety-related hydrogen recombiners are provided in the containment as the primary means of controlling postaccident hydrogen concentrations. A hydrogen purge system is provided for backup hydrogen control. See Section 6.2.5.3 (Safety Evaluation Eight).

ULNRC-06636 Page 48 of 374 Containment combustible gas control systems are discussed in detail in Section 6.2.5.

6.5.3.2 Secondary Containment This section is not applicable to SNUPPS.

6.5.4 ICE CONDENSER AS A FISSION PRODUCT CLEANUP SYSTEM This section is not applicable to SNUPPS.

6.

5.5 REFERENCES

1. Spraying Systems Company Topical Report No. SSCO-15215-1C-304SS-6.3-NP, April 1977, "Containment Spray Nozzles for Nuclear Power Plants"

2. "Design Considerations of Reactor Containment Spray Systems, The Corrosion of Materials in Spray Solutions," ORNL-TM-2412Part III, December 1969

3. NUREG-0800, Standard Review Plan Section 6.5.2, Revision 2, Containment Spray as a Fission Product Cleanup System, December 1988.

Insert 6.5.5

ULNRC-06636 Page 49 of 374 TABLE 6.5-1 ESF FILTRATION SYSTEMS INPUT PARAMETERS TO CHAPTER 15.0 ACCIDENT ANALYSIS Emergency exhaust filterReplace adsorberwith unitInsert T6.5-1(percent) efficiencies 90 Emergency exhaust system flowrate (SCFM) 9,000 Control room filter adsorber unit efficiency (percent) 95 Control room air conditioning system flowrate (SCFM) per train Filtered intake from control building 440 Filtered recirculation from control room 1,360

ULNRC-06636 Page 50 of 374 TABLE 6.5-2 INPUT PARAMETERS AND RESULTS OF SPRAY IODINE REMOVAL ANALYSIS

,maximum Core power rating 2.70 3,565 MWt Total containment free volume 2.50 x 106 ft3 **

Unsprayed containment free volume <15.0 percent Area coverage at the operating deck 13,500 cfm ***

Design >90 percent Average fall height Calculated of spray drops to >93 percent operating deck 131.4 feet Mixing rate between sprayed and unsprayed volumes 85,000 cfm Dose model One region Minimum vertical distance to operating deck from lowest 118 feet - 2 in.

spray header Net spray flow rate per train, injection Phase 3,131 gpm 3086 Number of spray pumps operating 1 Spray solution pH 4.0 - 7.0 (injection phase) 6.59 7.1 (recirculation phase at equilibrium) 20 Elemental iodine absorption coefficient, s, used In LOCA -1 10 hr (1) offsite and control room dose calculations Calculated s 25.7 hr-1 (2) 6.46 initially 37 hr-1 0.646 (for DF > 50)

Particulate iodine absorption coefficient, p, used in LOCA 0.45 hr-1 (3) offsite and control room dose calculations Calculated p 0.73 hr-1 (4)

Spray drop size, design See Figure 6.5-2 Schmidt number (See Section 6.5A.2) 11.58 Gas diffusivity (See Section 6.5A.2) 0.064 /sec Partition coefficient (See Section 6.5A.2) 5,000

ULNRC-06636 Enclosure 6 Page 51 of 374 TABLE 6.5-2 (Sheet 2)

Gas phase mass transfer coefficient (See Section 6.5A.3) 9.5 ft/min Terminal mass-mean drop velocity (See Section 6.5A.3) 790 ft/min Partition coefficient (See Section 6.5A.3) 1100 200 (1) Until DF = 28.7.

(2) s of 25.7 hr-1 was calculated in Section 6.5A.2 and used in the EQ dose calculations discussed in Section 3.11(B).1.2.2 s of 37 hr-1 was calculated in Section 6.5A.3 but 10 hr-1 was used in the offsite and control room dose calculations discussed in Section 15.6.5.

(3) Until DF = 50.

(4) p of 0.73 hr-1 was calculated in Section 6.5A.1 and used in the EQ dose calculations.

s of 25.7 hr-1 was calculated in Section 6.5A.2 and used in the EQ dose calculations discussed in Section 3.11(B).1.2.2 s of 20 hr-1 was calculated in Section 6.5A.3 and was used in the offsite and control room dose calculations discussed in Section 15.6.5

• • The maximum volume is conservative for the calculation of containment spray removal coefficients in Appendix 6.5A.3 and 6.5A.4 since the equation used includes the containment volume as the denominator.

• • • Per Regulatory Guide 1.183, the mixing rate attributed to natural convection between sprayed and unsprayed regions of the containment building, provided that adequate flow exists between these regions, is assumed to be two turnovers of the unsprayed regions per hour, unless other rates are justified. The unsprayed air volume is 405,000ft3, this results in a mixing rate of 405,000/30 minutes = 13,500 cfm

ULNRC-06636 Page 52 of 374 APPENDIX 6.5A - IODINE REMOVAL MODELS FOR THE CONTAINMENT SPRAY SYSTEM

ULNRC-06636 Page 53 of 374 6.5A.1 PARTICULATE IODINE MODEL FOR EQ DOSE CALCULATIONS The spray washout model for aerosol particles is represented in equation form as follows:

3hEF (6.5A-1)

P = ---------------

2dV Where:

P = spray removal constant for particles h = drop fall height E = total collection efficiency for a single drop F = spray volumetric flow rate d = mean drop diameter V = volume of sprayed region The capture of particles by falling drops results from Brownian diffusion, diffusiophoresis, interception, and impaction. Early in the injection phase, particles are removed mainly by impaction. Following injection, when the larger particles have already been removed, the removal rate is controlled by diffusiophoresis, which is the collection of particulates by steam condensing on the spray drops. The single drop collection efficiency, E, is taken as 0.0015, the minimum value observed in experimental tests (Ref. 1). The minimum collection efficiency, 0.0015, was only attained after the major fraction of airborne particles was removed. For early time periods, the removal rates were much higher than the minimum values ultimately reached. Per Reference 11, it is conservative to assume that E/D is 10 per meter initially (i.e., 1% efficiency for spray drops of one millimeter in diameter), changing abruptly to one per meter after the aerosol mass has been depleted by a DF of 50 (i.e., 98% of the particulate mass is ten times more readily removed than the remaining 2%). Using the 831 micron mean drop diameter identified in Table 6.5-2 and the minimum collection efficiency of 0.0015 from Reference 1, E/D would be 1.8 per meter which is consistent with the value from Reference 11 after a DF of 50 is attained.

The spray removal constant (P) for particulate iodine has been calculated to be 0.73/hr, based on equation 6.5A-1, and used in Section 3.11(B).1.2.2.

A limited and conservative credit for spray removal of airborne particulates containing iodine has been taken in Section 15.6.5, assuming the spray removal constant is 0.45/hr, until a decontamination factor of 50 is reached, following the postulated LOCA (see Table 6.5-2).

ULNRC-06636 Page 54 of 374 Where:

vf = the specific volume of liquid at saturation, ft3/lb v = the specific volume of the drop before condensation, ft3/lb hfg = the latent heat of evaporation, Btu/lb hg = the enthalpy of steam at saturation, Btu/lb d and d' = the drop diameter before and after condensation, cm Postma and Pasedag (Ref. 6) conclude that condensation will tend to increase the iodine washout rate due to the increased volume of the spray. Their effect has been conservatively ignored.

The drop exposure time calculated is based on the assumption that the drops were sprayed in such a manner that the initial downward velocity of the drops at the spray ring header elevation was zero. The drops fall under the effect of gravity from the spray ring header to the operating deck. The minimum height is given in Table 6.5-2. As the drop size increases, the average exposure time decreases from about 20 to 5 seconds.

Incorporating the above parameters into equation 6.5A-16 with the sprayed containment volume, V, and assuming a single spray header flow rate, the value of the spray removal coefficient calculated (25.7 hr-1) is presented in Table 6.5-2.

The resulting elemental iodine spray removal constant is greater than 10/hr. A conservative removal constant of 10/hr is assumed and used in the design basis LOCA evaluations presented in Section 15.6.5.

6.5A.3 ELEMENTAL IODINE MODEL FOR OFFSITE AND CONTROL ROOM DOSE CALCULATIONS As discussed in Reference 11, the effectiveness of the spray during the injection phase against elemental iodine vapor is chiefly determined by the rate at which fresh solution surface area is introduced into the containment atmosphere. The rate of solution created per unit gas volume in the containment atmosphere may be estimated as (6F/VD), where F is the spray volumetric flow rate, V is the volume of the sprayed region, and D is the mean diameter of the spray drops. The first-order spray removal constant for elemental iodine, s, may be taken to be:

6kgTF s = -----------------

VD where kg is the gas phase mass transfer coefficient and T is the drop fall time (or drop exposure time), which may be estimated by the ratio of the average fall height to the

ULNRC-06636 Enclosure 6 Page 55 of 374 terminal velocity of the average drop. The above expression represents a first-order approximation if a well-mixed droplet model is used for spray absorption efficiency. This expression is valid for s values equal to or greater than 10 per hour but less than 20 per hour. Using this expression and the values contained in Table 6.5-2 a value of 37 hr-1 is calculated. A value of 10 per hour will continue to be used in the dose calculations of Section 15.6.5. Insert 6.5A.3 Spray removal of elemental iodine continues until the DF of Equation 6.5A-15 is reached.

Although the VL term in Equation 6. 5A-15 Deleterepresents the volume of the sumps plus any overflow from the sumps, it is conservative to just use the volume of the sumps for VL since a lower DF will result. The value for the partition coefficient, H, in Equation 6.5A-15 was taken from Figure 6 of Reference 13 using the 323°K plot at 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> (representative of the average conditions during a LOCA). The value of 1100 used is considered to be conservative since the sump fluid temperature at 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> would be greater than 323°K per Figure 6.2.1-7 and Figure 6 of Reference 13 shows that higher temperatures would be associated with higher partition coefficients. The resulting DF is calculated to be 28.7 6.5A.4 REFERENCES Insert 6.5A.4

1. Hilliard, R. K., Coleman L. F., "Natural Transport Effects of Fission Product Behavior in the Containment System Experiment," BNWL-1457, Battelle Pacific Northwest Laboratories, Richland, Washington, December 1970.

2. Hilliard, R. K., et al, "Removal of Iodine and Particulates from Containment Atmospheres by Sprays - Containment Systems Experiment Interim Report,"

6.5A.5 BNWL-1244, 1970.

3. Perkins, J. F., "Decay of U235 Fission Products," Physical Science Laboratory, RR-TR-63-11, U.S. Army Missile Command Redstone Arsenal, Alabama, July 25, 1963.

4. Parsley, Jr., L. F., "Design Considerations of Reactor Containment Spray Systems - Part VII," ORNL TM 2412, Part 7, 1970.

5. Ranz, W.E., and Marshall, Jr., W.R., "Evaporation from Drops," Chemical Engineering Progress 48, 141-46, 173-80, 1952.

6. Postma, A. K., and Pasedag, W. F., "A Review of Mathematical Models for Predicting Spray Removal of Fission Products in Reactor Containment Vessels,"

WASH-1329, U.S. Atomic Energy Commission, June 1974.

7. Griffiths, V., "The Removal of Iodine from the Atmosphere by Sprays," Report No.

AHSB(S)R45, United Kingdom Atomic Energy Authority, London, 1963.

ULNRC-06636 Page 56 of 374

8. Eggleton, A. E. J., "A Theoretical Examination of Iodine-Water Partition Coefficient," AERE (R)-4887, 1967.

9. Postma, A. K., Coleman, L. F., and Hilliard, R. K., "Iodine Removal from Containment Atmospheres by Boric Acid Spray," BNP-100, Battelle-Northwest, Richland, Washington, 1970.

10. Coleman, L. F., "Iodine Gas-Liquid Partition," Nuclear Safety Quarterly Report, February, March, April 1970, BNWL-1315-2, Battelle-Northwest, Richland, Washington, p. 2.12-2.19, 1970.

11. NUREG-0800, Standard Review Plan Section 6.5.2, Revision 2, Containment Spray as a Fission Product Cleanup System, December 1988.

12. ANSI/ANS-56.5-1979, PWR and BWR Containment Spray System Design Criteria.

13. E.C. Beahm, W. E. Shockley, C. F. Weber, S. J. Wisbey, and Y. M. Wang, Chemistry and Transport of Iodine in Containment, NUREG/CR-4697, October 1986.

Insert 6.5A.5 here

ULNRC-06636 Page 57 of 374 CALLAWAY - SP CHAPTER 7.0 INSTRUMENTATION AND CONTROLS

7.1 INTRODUCTION

This section describes the various plant instrumentation and control systems and the functional performance requirements, design bases, system descriptions, design evaluations, and tests and inspections for each. The information provided in this chapter emphasizes those instruments and associated equipment which constitute the protection system, as defined in IEEE Standard 279-1971, "IEEE Standard: Criteria for Protection Systems for Nuclear Power Generating Stations."

The instrumentation and control systems provide automatic protection and exercise proper control against unsafe and improper reactor operation during steady state and transient power operations (Conditions I and II) and to provide initiating signals to mitigate the consequences of emergency and faulted conditions (Conditions III and IV).

ANS conditions are discussed in Chapter 15.0.

Applicable criteria and codes are listed in Table 7.1-2.

7.1.1 IDENTIFICATION OF SAFETY-RELATED SYSTEMS Safety-related instrumentation and control systems and their supporting systems are those systems required to ensure:

a. The integrity of the reactor coolant or Control Roompressure boundary.

b. The capability to shut down the reactor and maintain it in a safe shutdown condition.

c. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline exposures of 10 CFR 100.

The definitions provi ded below 50.67, are used Guide Regulatory to classify theorinstrumentation 1.183, 10 CFR 100, assys tems into the categories defined forappropriate.

Chapter 7.0 by Regulatory Guide 1.70.

A listing of these systems, by categories, that are comparable to those of nuclear power plants of similar design is given in Table 7.1-1. Table 7.1-1 also identifies the systems that are different with references to discussions of those differences.

The plant's control and instrumentation systems are grouped into the following categories:

a. Reactor trip system (RTS) 7.1-1 Rev. OL-24 11/19

ULNRC-06636 Page 58 of 374 9.1.4.1 Design Bases 9.1.4.1.1 Safety Design Bases The portions of the FHS that are safety related are the containment isolation features of the fuel transfer tube and the crane structural components which prevent falling of major crane components onto fuel assemblies or safe shutdown equipment.

SAFETY DESIGN BASIS ONE - The FHS is protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles (GDC-2).

SAFETY DESIGN BASIS TWO - The FHS is designed to remain intact after an SSE or following the postulated hazards of fire, internal missiles, or pipe breaks (GDC-3 and 4).

SAFETY DESIGN BASIS THREE - The FHS components are capable of being tested during plant operation. Provisions are made to allow for inservice inspection and testing of components at appropriate times.

SAFETY DESIGN BASIS FOUR - The FHS is designed and fabricated to codes consistent with the seismic category assigned by Regulatory Guide 1.29 and industry standard specifications.

SAFETY DESIGN BASIS FIVE - The containment isolation provisions for the system are selected, tested, and located in accordance with the requirements of GDC-54 and 10 CFR 50, Appendix J, Type B testing.

50.67 and Regulatory Guide 1.183 SAFETY DESIGN BASIS SIX - The FHS is designed and arranged so that there are no loads which, if dropped, could result in damage, leading to the release of radioactivity in excess of 10 CFR 100 guidelines, or impair the capability to safely shut down the plant.

Specific administrative controls for handling of the spent fuel pool transfer gates are addressed in Section 9.1.4.3.

This meets the requirements of Regulatory Guide 1.13, as described in Table 9.1-3.

9.1.4.2 System Description 9.1.4.2.1 General Description The fuel handling system consists of the equipment needed to refuel the reactor core.

Basically, this equipment is composed of cranes, handling equipment, and a fuel transfer system.

The associated fuel handling structures are divided into seven areas. In general, these areas are:

ULNRC-06636 Page 59 of 374 SAFETY EVALUATION ONE - The safety-related portions of the FHS are located in the reactor and fuel buildings. These buildings are designed to withstand the effects of earthquakes, tornadoes, hurricanes, floods, external missiles, and other appropriate natural phenomena. Sections 3.3, 3.4, 3.5, 3.7(B), and 3.8 provide the bases for the adequacy of the structural design of these buildings.

SAFETY EVALUATION TWO - The safety-related portions of the FHS are designed to remain intact after an SSE. Section 3.7(B) provides the design loading conditions that were considered. Sections 3.5, 3.6, and 9.5.1 provide the required hazards analysis.

SAFETY EVALUATION THREE - The FHS is initially tested with the program given in Chapter 14.0. Periodic inservice functional testing is done in accordance with Section 9.1.4.4. The fuel transfer tube is inspected in accordance with the technical requirements of ASME Section XI.

SAFETY EVALUATION FOUR - Section 3.2 delineates the seismic category applicable to the safety-related portions of this system.

50.67 and Regulatory Guide 1.183 SAFETY EVALUATION FIVE - Sections 6.2.4 and 6.2.6 provide the safety evaluation for the system containment isolation arrangement and testability.

SAFETY EVALUATION SIX - In the event of a fuel handling accident in the fuel building, the radiological consequences analyzed in Chapter 15.0 demonstrate that the 10 CFR Part 100 guideline values are not exceeded. The circumstances resulting in a handling accident are limited to the following conditions.

a. Fuel drop from a lifting device

b. Improper operation of the transfer equipment and cranes

c. DELETED

d. Drop of the RV head The fuel handling equipment is designed to prevent a fuel assembly drop by providing special gripping devices which are locked in a manner which will not allow the release of the fuel assembly during transfer. The special features are described in Section 9.1.4.2.2.

Improper operation of the fuel transfer system is prevented by the location of special limit switches and interlocks which will not allow the movement of fuel assemblies unless they are properly oriented, thus avoiding a fuel handling accident. Further description of these devices is given in Section 9.1.4.2.2.

Limit switches and interlocks located on the fuel handling cranes in conjunction with administrative controls prevent any improper operations which may result in a fuel

ULNRC-06636 Page 60 of 374 TABLE 9.1-3 DESIGN COMPARISION TO REGULATORY POSITIONS OF REGULATORY GUIDE 1.13 REVISION 1, DATED DECEMBER 1975, TITLED SPENT FUEL STORAGE FACILITY DESIGN BASIS Regulatory Guide 1.13 Position Union Electric

1. The spent fuel storage facility (including its 1. Complies as structures and equipment, except as noted in Paragraph 6 described in below) should be designed to Category I seismic Section 9.1.2.1.1 requirements.

2. The facility should be designed (a) to keep tornadic 2. Complies as winds and missiles generated by these winds from the fuel described in storage pool and (b) to keep missiles generated by Section 3.5, and tornadic winds from contacting fuel within the pool. 3.8.

3. Interlocks should be provided to prevent cranes 3. Complies as from passing over stored fuel (or near stored fuel in a described in manner such that if a crane failed the load could tip over on Section 9.1.4.

stored fuel) when fuel handling is not in progress. During fuel handling operations, the interlocks may be bypassed and administrative control used to prevent the crane from carrying loads that are not necessary for fuel handling over the stored fuel or other prohibited areas. The facility should be designed to minimize the need for bypassing such interlocks.

4. A controlled leakage building should enclose the 4. Complies as fuel pool. The building should be equipped with an described in appropriate ventilation and filtration system to limit the Section 9.4.2 and potential release of radioactive iodine and other radioactive 15.7.4.

materials. The building need not be designed to withstand extremely high winds, but leakage should be suitably controlled during refueling operations. The design of the ventilation and filtration system should be based on the assumption that the cladding of all of the fuel rods in one fuel bundle might be breached. The inventory of radioactive materials available for leakage from the building should be based on the assumptions given in Regulatory Guide 1.25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors" (Safety Guide 25).

Replace with:

USNRC, Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Plants, July 2000 Rev. OL-13

ULNRC-06636 Enclosure 6 Page 61 of 374 CALLAWAY - SP building ventilation system is designed to provide fresh air ventilation at a minimum rate of 0.1 cfm per square foot of floor area.

POWER GENERATION DESIGN BASIS TWO - The control building exhaust system serves to remove from the control building the hydrogen generated by the batteries during normal operation.

POWER GENERATION DESIGN BASIS THREE - The access control air-conditioning system provides RP access control areas, and the nonvital electric equipment areas of the electrical and mechanical equipment level with an environment suitable for personnel comfort and electrical equipment operation.

POWER GENERATION DESIGN BASIS FOUR - The access control exhaust system collects and processes the effluents from the potentially contaminated regions of the access control area. The exhaust system is designed to meet the requirements of the discharge concentration limits of 10 CFR 20 and the as-low-as-reasonably-achievable dose objective of 10 CFR 50, Appendix I. The access control exhaust system charcoal adsorption train complies with Regulatory Guide 1.140, to the extent discussed in Table 9.4-3.

POWER GENERATION DESIGN BASIS FIVE - The counting room recirculation system provides adequate cooling, humidity control, and filtering of the counting room environment for personnel and equipment.

POWER GENERATION DESIGN BASIS SIX - A supplemental cooling train is provided for each Class 1E electrical equipment air-conditioning system train. During periods when one Class 1E electrical equipment air-conditioning system train is unavailable or removed from service for a limited period of time (such as for online maintenance), the supplemental cooling train provides additional ventilation via forced ventilation flowpaths that enable the other functional Class 1E electrical equipment air conditioning train to provide cooling to the rooms and areas for both trains of Class 1E electrical equipment.

9.4.1.2 System Description 9.4.1.2.1 General Description The control building HVAC systems are shown in Figure 9.4-1. The systems consist of the control building supply system, control room air-conditioning system with supplemental filtration and pressurization systems, Class 1E electrical equipment air-conditioning system, access control air-conditioning system, counting room recirculation system, control building exhaust system, and the access control exhaust system. The design conditions for these systems are presented in Table 3.11(B)-1.

Potential radiation doses in the control room are discussed in Chapter 15.0.

The control building is serviced by an outside-air-supply system which provides fresh cooled or heated air to each of the various levels of the building. Self-contained The model for control room dose analysis is presented in Appendix 15A.

9.4-3 Rev. OL-24 11/19

ULNRC-06636 Page 62 of 374 The indication of the amount of filter loading for all filters associated with the essential and nonessential air handlers is provided at each of the air handlers.

Alarms are provided in the control room to indicate high charcoal bed temperatures in the control room filtration, control room pressurization and access control filtration units and high room temperature in the ESF switchgear and dc switchgear rooms.

An alarm is provided in the control room to indicate high hydrogen concentrations in a battery room.

Alarms are provided in the control room to indicate high carbon monoxide/carbon dioxide concentrations, high radiation, and smoke in the control building intake.

All instrumentation provided with the filtration units is as required by Regulatory Guide 1.52 or 1.140, as applicable.

9.4.2 FUEL BUILDING HVAC The fuel building ventilation system consists of the fuel building supply system which includes the fuel building heating coil, the fuel building supply air unit, and the fuel handling area cooling coil; the emergency exhaust system, including the emergency exhaust heating coil; the auxiliary/fuel building normal exhaust system; the spent fuel pool cooling pump room coolers; and the unit heaters. Since both the emergency exhaust system and the auxiliary/fuel building normal exhaust system also serve the release auxiliary building, their operation in t he auxiliary building is discussed in Section 9.4.3.

The fuel building supply system provides conditioned outside air for ventilation and cooling or heating, as required, to all areas of the fuel building. The auxiliary/fuel building normal exhaust system exhausts air from the area above the spent fuel pool during normal operation and provides a means of purging smoke following a postulated fire.

In the event of a fuel handling accident, the emergency exhaust system collects and processes the airborne particulates in the fuel building. In the event of a LOCA, the emergency exhaust system processes the atmosphere of the auxiliary building.

The fuel storage pool cooling pump room coolers provide a suitable ambient temperature for the electric motor drives of the safety-related pumps.

The fuel building unit heaters provide supplemental heating for the fuel building, when required.

and is also credited (based upon manual actuation) to insure any release from containment to the auxiliary building via an open personnel hatch is transported to the plant vent for release.

ULNRC-06636 Enclosure 6 Page 63 of 374 SAFETY EVALUATION FOUR - The fuel storage pool cooling pump room coolers, the emergency exhaust system, and the fuel building HVAC boundary penetration isolation provisions are initially tested with the program given in Chapter 14.0. Periodic inservice functional testing is done in accordance with Section 9.4.2.4.

The exhaust system is also credited (based upon manual actuation)

Section during a fuel6.6 provides handling the ASME accident Boiler and Pressure in containment Vessel Code,Section XI to insure any requirements that are appropriate for the spent fuel release from containment to the auxiliary building via an open pool cooling pump room coolers.

personnel hatch is transported to the plant vent for release.

SAFETY EVALUATION FIVE - Section 3.2 delineates the quality group classification and seismic category applicable to the safety-related portion of this system and supporting system. All the power supplies and control function necessary for safe function of the fuel storage pool cooling pump room coolers, emergency exhaust system, and the fuel building HVAC boundary penetration isolation provisions are Class 1E, as described in Chapters 7.0 and 8.0.

SAFETY EVALUATION SIX - Section 9.4.2.2.3 describes the provisions made to assure the isolation of the auxiliary building following a DBA.

SAFETY EVALUATION SEVEN - The emergency exhaust system maintains a negative pressure of no less than 1/4 in. w.g. in the fuel building to prevent unprocessed exfiltration following a fuel handling accident which releases radioactivity. The emergency 50.67 andexhaust systemGuide Regulatory 1.183 for radioactivity downstream of the filter is monitored adsorber unit prior to release to the site. The filter adsorber unit limits the radiological consequences of a fuel handling accident to less than 10 CFR 100 limits.

SAFETY EVALUATION EIGHT - Room coolers are installed in each fuel storage pool cooling pump room and are designed to maintain these rooms below 122°F (50°C),

based on maximum heat load within the room.

9.4.2.4 Tests and Inspections Preoperational testing is described in Chapter 14.0.

Filters and adsorbers for the emergency exhaust system are tested in the manufacturer's shop, after initial installation and subsequent to each filter or adsorber change. After installation, interim tests and inspections will be performed after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of operation and once per 18 months in accordance with the requirements of Regulatory Guide 1.52 and ASTM D3803-1989 as discussed in Table 9.4-2, to detect any deterioration of components that may develop under service or standby conditions.

Prefilters will not undergo factory or inplace testing since no credit is taken for removal of particulates in meeting permissible dose rates.

HEPA filters will be factory tested with monodispersed DOP aerosol to demonstrate a minimum particulate removal efficiency of no less than 99.97 percent for 0.3 micron

ULNRC-06636 Page 64 of 374 Depending on the equipment in the compartments, the access varies from Zones B through E. Corridors are shielded to allow Zone B access, and operator areas for valve compartments are limited to Zone C access.

Removable sections of block shield walls and concrete plugs are utilized to replace worn-out equipment and spent filter cartridges, respectively. Partial shield walls are placed between equipment in compartments with more than one piece of equipment to permit maintenance access.

12.3.2.2.5 Turbine Building Shielding Design Radiation shielding is not required for any process equipment located in the turbine building. All areas in the turbine building are classified Zone A.

TEDE 12.3.2.2.6 Control Room Shielding Design The design basis LOCA dictates the shielding requirements for the control room.

Shielding is provided to permit access and occupancy of the control room under LOCA conditions with radiation doses limited to 5 rem whole body from all contributing modes of exposure for the duration of the accident, in accordance with GDC-19. A complete discussion of control room habitability during a LOCA is provided in Section 6.4. Figure 12.3-3 provides an isometric view of the control room shielding.

12.3.2.2.7 Diesel Generator Building Shielding Design There are no radiation sources in the diesel generator building. Therefore, no shielding is required within the building.

12.3.2.2.8 Miscellaneous Plant Areas and Plant Yard Areas Sufficient shielding is provided for all plant buildings containing radiation sources so that radiation levels at the accessible outside surfaces of the buildings are maintained below Zone A levels. Plant yard areas which are frequently occupied by plant personnel are fully accessible during normal operation and shutdown. These areas are surrounded by a security fence and closed off from areas accessible to the general public. Access to outside storage tanks which have a contact dose rate greater than 0.5 mrem/hr is restricted by a fence located at a distance at which the dose rate is less than 0.5 mrem/

hr.

12.3.2.2.9 Independent Spent Fuel Storage Installation (ISFSI) Shielding Design The ISFSI is designed for interim dry storage of spent nuclear fuel. The shielding is sufficient to comply with the requirements of 10 CFR 72.104 and 10 CFR 72.106. The dose rate on the VVM closure lid is calculated to be 0.25 mrem/hr neutron and 0.69 mrem/hr gamma for a total dose rate of 0.94 mrem/hr. The dose rate on the outlet duct screen is calculated to be 0.80 mrem/hr neutron and 1.32 mrem/hr gamma for a total

ULNRC-06636 Page 65 of 374 12.3.3.1 Design Objectives The plant HVAC systems for normal plant operation and anticipated operational occurrences are designed to meet the requirements of 10 CFR 20, "Standards for Protection Against Radiation," and 10 CFR 50, "Licensing of Production and Utilization Facilities."

12.3.3.2 Design Criteria ,10 CFR 50.67 and Regulatory Guide 1.183, as appropriate Design criteria for the plant HVAC systems include the following:

a. During normal operation and anticipated operational occurrences, the average and maximum airborne radioactivity levels to which plant personnel are exposed in the restricted areas of the plant are ALARA and within the limits specified in 10 CFR 20.

b. During normal operation and anticipated operational occurrences, the dose from concentrations of airborne radioactive material in unrestricted areas beyond the site boundary will be ALARA and within the limits specified in 10 CFR 20 and 10 CFR 50.

c. The plant siting dose guidelines of 10 CFR 100 will be satisfied, following those hypothetical accidents described in Chapter 15.

d. The dose to control room personnel shall not exceed the limits specified in GDC-19, following those hypothetical accidents described in Chapter 15.0 and Section 6.4.

12.3.3.3 Design Guidelines In order to accomplish the design objectives, the following guidelines are followed, wherever practicable.

12.3.3.3.1 Guidelines to Minimize Airborne Radioactivity

a. Access control and traffic patterns are considered in the basic plant layout to minimize the spread of contamination.

b. Equipment vents and drains are piped directly to a collection device connected to the collection system, instead of allowing any contaminated fluid to flow across the floor to the floor drain.

c. All-welded piping systems are employed on contaminated systems, to the maximum extent practicable, to reduce system leakage. If welded piping systems are not employed, drip trays are provided at the points of potential leakage. Drains from drip trays are piped directly to the collection system.

ULNRC-06636 Page 66 of 374

e. The clear space for doors is a minimum of 3 feet by 7 feet.

f. The filters are designed with replaceable 2 feet by 2 feet units that are clamped in place against compression seals. The filter housing is designed, tested, and proven to be airtight with bulkhead type doors that are closed against compression seals.

, 10 CFR 50.67 and Regulatory Guide 1.183, as 12.3.4 AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING appropriate INSTRUMENTATION 12.3.4.1 Area Radiation Monitoring The area radiation monitoring system (ARMS) is provided to supplement the personnel and area radiation survey provisions of the plant radiation protection program described in Section 12.5 and to ensure compliance with the personnel radiation protection guidelines of 10 CFR 20, 10 CRF 50, 10 CFR 70, and Regulatory Guides 8.2, 8.8, and 8.12.

12.3.4.1.1 Design Bases The principal objectives and criteria of the ARMS are provided below.

12.3.4.1.1.1 Safety Design Bases The area radiation monitoring system has no function related to the safe shutdown of the plant or the capability to mitigate the consequences of accidents that could result in offsite exposures comparable to the guideline exposure of 10 CFR 100 and, therefore, has no safety design bases. See Appendix 7A for a discussion of Regulatory Guide 1.97.

12.3.4.1.1.2 Power Generation Design Bases POWER GENERATION DESIGN BASIS ONE - The ARMS functions continuously to immediately alert plant personnel entering or working in nonradiation or low-radiation areas of increasing or abnormally high radiation levels which, if unnoticed, could possibly result in inadvertent overexposures.

POWER GENERATION DESIGN BASIS TWO - The ARMS serves to inform the control room operator of the occurrence and approximate location of an abnormal radiation increase in nonradiation or low-radiation areas.

POWER GENERATION DESIGN BASIS THREE - The ARMS complies with the requirements of 10 CFR 50, Appendix A, General Design Criterion 63 for monitoring fuel and waste storage and handling areas.

ULNRC-06636 Page 67 of 374 TABLE OF CONTENTS (Continued)

Section Page 15.6.6 A NUMBER OF BWR TRANSIENTS .................................................... 15.6-50 15.

6.7 REFERENCES

...................................................................................... 15.6-50 15.7 RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT........ 15.7-1 15.7.1 RADIOACTIVE WASTE GAS DECAY TANK FAILURE .......................... 15.7-1 15.7.1.1 Identification of Causes ...................................................................... 15.7-1 15.7.1.2 Sequence of Events and System Operations .................................... 15.7-1 15.7.1.3 Core and System Performance .......................................................... 15.7-1 15.7.1.4 Barrier Performance........................................................................... 15.7-1 15.7.1.5 Radiological Consequences .............................................................. 15.7-2 15.7.2 RADIOACTIVE LIQUID WASTE SYSTEM LEAK OR FAILURE ............. 15.7-5 15.7.2.1 Identification of Causes ...................................................................... 15.7-5 15.7.2.2 Sequence of Events and System Operation ...................................... 15.7-5 15.7.2.3 Core and System Performance .......................................................... 15.7-5 15.7.2.4 Barrier Performance........................................................................... 15.7-5 15.7.2.5 Radiological Consequences .............................................................. 15.7-5 15.7.3 POSTULATED RADIOACTIVE RELEASE DUE TO LIQUID TANK FAILURES............................................................................................... 15.7-8 15.7.4 FUEL HANDLING ACCIDENTS .............................................................. 15.7-8 15.7.4.1 Identification of Causes and Accident Description ............................. 15.7-8 15.7.4.2 Sequence of Events and Systems Operations ................................... 15.7-9 15.7.4.3 Core and System Performance .......................................................... 15.7-9 15.7.4.4 Barrier Performance........................................................................... 15.7-9 15.7.4.5 Radiological Consequences .............................................................. 15.7-9 15.8 ANTICIPATED TRANSIENTS WITHOUT SCRAM....................................... 15.8-1 15A.1 GENERAL ACCIDENT PARAMETERS .................................................. 15.A-1 15A.2 OFFSITE RADIOLOGICAL CONSEQUENCES CALCULATIONAL MODELS.................................................................................................. 15.A-1 15A.2.1 ACCIDENT RELEASE PATHWAYS .................................................. 15.A-1 15A.2.3 TWO - REGION SPRAY MODEL IN CONTAINMENT (LOCA) ......... 15.A-5 Insert Chapter15TOC-A 15.0-viii

ULNRC-06636 Enclosure 6 Page 68 of 374 TABLE OF CONTENTS (Continued)

Section Page 15A.2.4 OFFSITE THYROID DOSE CALCULATION MODEL........................ 15.A-8 15A.2.5 OFFSITE TOTAL-BODY DOSE CALCULATIONAL MODEL............. 15.A-9 15A.3 CONTROL ROOM RADIOLOGICAL CONSEQUENCES CALCULATIONAL MODELS ................................................................... 15.A-9 15A.3.1 INTEGRATED ACTIVITY IN CONTROL ROOM.............................. 15.A-10 15A.3.2 CONTROL ROOM THYROID DOSE CALCULATIONAL MODEL............................................................................................. 15.A-12 15A.3.3 CONTROL ROOM BETA-SKIN DOSE CALCULATIONAL MODEL............................................................................................. 15.A-13 15A.3.4 CONTROL ROOM TOTAL-BODY DOSE CALCULATION .............. 15.A-13 15A.4 REFERENCES ...................................................................................... 15.A-14 6

App 15B Regulatory Guide 1.183, Revision 0 Conformance Tables 15B-1 15.0-ix

ULNRC-06636 Page 69 of 374 LIST OF TABLES (Continued) Replace with "Deleted" Number Title 15.6-6 Parameters Used in Evaluating the Radiological Consequences of a Loss-Of-Coolant-Accident 15.6-7 Design Comparison to the Regulatory Positions of Regulatory Guide 1.4 Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors, Revision 2, June 1974 15.6-8 Radiological Consequences of a Loss-of-Coolant-Accident 15.6-9 Summary of Large Break LOCA Analysis Assumptions 15.6-10 Minimum and Maximum Safety Injection Flows For LBLOCA 15.6-10a Deleted 15.6-10b Deleted 15.6-11 Input Parameters Used in the Small Break LOCA Analysis 15.6-12 Safety Injection Flows vs. Pressure for SBLOCA - Minimum Safeguards, Spill to RCS Pressure 15.6-13 Large Break LOCA Results 15.6-14 Large Break LOCA Time Sequence of Events 15.6-15 NOTRUMP Transient Results 15.6-16 Beginning of Life (BOL) Rod Heatup Results 15.7-1 Design Comparison to the Regulatory Positions of Regulatory Guide 1.24 Assumptions Used for Evaluating the Potential Consequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure Revision 0, Dated March 23, 1972 15.7-2 Design Comparison to the Regulatory Positions of Regulatory Guide 1.25 Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors Revision 0, Dated March 23, 1972 Replace with "Deleted"

ULNRC-06636 Page 70 of 374 LIST OF TABLES (Continued)

Number Title 15.7-3 Parameters used in evaluating the radiological consequences of a waste gas decay tank rupture 15.7-4 Radialogical Consequences of a Waste Gas Decay Tank Rupture 15.7-5 Parameters used in Evaluating the Radiological Consequences of a Liquid Radwaste Tank Failure 15.7-6 Radiological Consequences of a Liquid Radwaste Tank Failure 15.7-7 Parameters used in Evaluating the Radiological Consequences of a Fuel-Handling Accident 15A-3 Core Inventory (Ci) and WGDTR Fuel 15.7-8 Radiological Consequenc esand of aRod FuelGap Activties Handling Accident 15A-1 Parameters Used in Accident Analysis 15A-2 Limiting Short-Term Atmospheric Dispersion Factors (/Qs) for Accident analysis 15A-3 Fuel and Rod Gap Inventories - Core (Ci) 15A-4 Dose Conversion Factors Used in Accident Analysis 15A-5 Initial Radioactivity For Accidents That Use the Primary-to-Secondary Leakage Release Pathway Insert Chapter15List_Of_Tables

ULNRC-06636 Enclosure 6 Page 71 of 374 Replace with new CALLAWAY - SP Figures LIST OF FIGURES (Continued)

Number Title 15.6-3a Pressurizer and Steam Generator (Ruptured and Intact Generators)

Pressure Transients for Steam Generator Tube Rupture Event 15.6-3b Reactor Coolant System Temperature (Ruptured Loop) Transient for Steam Generator Tube Rupture Event 15.6-3c Reactor Coolant System Temperature (Intact Loops) Transient for Steam Generator Tube Rupture Event 15.6-3d Steam Flow Rate (Intact Generators) Transient for Steam Generator Tube Rupture Event 15.6-3e Steam Flow Rate (Ruptured Generator) Transient for Steam Generator Tube Rupture Event 15.6-3f Steam Generator Temperature (Faulted and Intact Generators)

Transients for Steam Generator Tube Rupture Event 15.6-3g Steam Generator Atmospheric Relief Valve Flow Rate (Ruptured Generator) Transient for Steam Generator Tube Rupture Event 15.6-3h Steam Generator Atmospheric Relief Valve Flow Rate (Intact Generators) Transient for Steam Generator Tube Rupture Event 15.6-3i Faulted Steam Generator Break Flow Rate Transient for Steam Generator Tube Rupture Event 15.6-3j Auxiliary Feedwater Flow Rate and Narrow Range Level (Ruptured Generator) Transients for Steam Generator Tube Rupture Event 15.6-3k Auxiliary Feedwater Flow Rate and Narrow Range Level (Intact Generators) Transients for Steam Generator Tube Rupture Event 15.6-3l Steam Generator Liquid Volume (Ruptured Generator) Transient for Steam Generator Tube Rupture Event 15.6-3m Pressurizer PORV Flow Rate Transient for Steam Generator Tube Rupture Event 15.6-3n Pressurizer Liquid Volume Transient for Steam Generator Tube Rupture Event

ULNRC-06636 Enclosure 6 Page 72 of 374 CALLAWAY - SP Replace with new Figures LIST OF FIGURES (Continued)

Number Title 15.6-3o Feedwater Flow Rate (Ruptured Generator) Transient for Steam Generator Tube Rupture Event 15.6-3p Feedwater Flow Rate (Intact Generators) Transient for Steam Generator Tube Rupture Event 15.6-4 Sequence of Events for Large Break LOCA Analysis 15.6-4a Deleted 15.6-4b Deleted 15.6-5 Deleted 15.6-5a Cladding Temperature at PCT and Burst Elevations (CD=0.4, Low Tavg, MIN SI, Cosine Power Shape, non-IFBA) 15.6-5b Cladding Temperature at PCT and Burst Elevations (CD=0.6, Low Tavg, MIN SI, Cosine Power Shape, non-IFBA) 15.6-5c Cladding Temperature at PCT and Burst Elevations (CD=0.8, Low Tavg, MIN SI, Cosine Power Shape, non-IFBA) 15.6-5d Cladding Temperature at PCT and Burst Elevations (CD=1.0, Low Tavg, MIN SI, Cosine Power Shape, non-IFBA) 15.6-5e Cladding Temperature at PCT and Burst Elevations (CD=0.6, High Tavg, MIN SI, Cosine Power Shape, non-IFBA) 15.6-5f Cladding Tempertaure at PCT and Burst Elevations (CD=0.6, High Tavg, MAX SI, Cosine Power Shape, non-IFBA) 15.6-5g Cladding Temperature at PCT and Burst Elevations (CD=0.6, High Tavg, MIN SI, 8.5 Power Shape, non-IFBA) 15.6-5h Cladding Temperature at PCT and Burst Elevations (CD=0.6, High Tavg, MIN SI, Cosine Power Shape, IFBA) 15.6-6 Deleted 15.6-6a Core Pressure During Blowdown (CD=0.4, Low TAVG, MIN SI, Cosine Power Shape, non-IFBA)

ULNRC-06636 Enclosure 6 Page 73 of 374 LIST OF FIGURES (Continued)

Number Title 15.6-29 4 Inch Cold Leg Break Core Exit Steam Flow vs Time Replace with new figures 15.6-30 4 Inch Cold Leg Break Core Heat Transfer Coefficient vs Time 15.6-31 4 Inch Cold Leg Break Hot Spot Fluid Temperature vs Time 15.6-32 4 Inch BL and IL Pumped SI Flow Rate vs Time 15.6-33 Deleted 15.6-33a Pressurizer and Steam Generator (Ruptured and Intact Generators)

Pressure Transients for SGTR Event with Overfill 15.6-33b Reactor Coolant System Temperature (Ruptured Loop) Transient for SGTR Event with Overfill 15.6-33c Reactor Coolant System Temperature (Intact Loops) Transient for SGTR Event with Overfill 15.6-33d Reactor Coolant System and Steam Generator (Ruptured and Intact Generators) Water Mass Transient for SGTR Event with Overfill 15.6-33e Ruptured Steam Generator Break Flow Flashing Fraction Transient for SGTR Event with Overfill 15.6-33f Steam Generator Temperature (Ruptured and Intact Generators)

Transient for SGTR Event with Overfill 15.6-33g Steam Generator Atmospheric Release Flow Rate (Ruptured Generator) Transient for SGTR Event with Overfill 15.6-33h Steam Generator Atmospheric Release Flow Rate (Intact Generators) Transient for SGTR Event with Overfill 15.6-33i Ruptured Steam Generator Break Flow Rate Transient for SGTR Event with Overfill 15.6-33j Auxiliary Feedwater Flow Rate and Narrow Range Level (Ruptured Generator) Transients for SGTR Event with Overfill 15.6-33k Auxiliary Feedwater Flow Rate and Narrow Range Level (Intact Generators) Transients for SGTR Event with Overfill

ULNRC-06636 Page 74 of 374 Replace with new LIST OF FIGURES (Continued) figures Number Title 15.6-33l Ruptured Steam Generator Liquid Volume Transient for SGTR Event with Overfill 15.6-33m Pressurizer PORV Flow Rate Transient for SGTR Event with Overfill 15.6-33n Pressurizer Liquid Volume Transient for SGTR Event with Overfill 15.6-34 Deleted 15.6-35 Deleted 15.6-36 Deleted 15.6-37 Deleted 15.6-38 Deleted 15.6-39 Deleted 15.6-40 Deleted 15.6-41 Deleted 15.6-42 Deleted 15.6-43 Deleted 15.6-44 Deleted 15.6-45 Deleted 15.6-46 Deleted 15.6-47 Deleted 15.6-48 Deleted 15.6-49 Deleted 15.6-50 Deleted 15.6-51 Deleted

ULNRC-06636 Enclosure 6 Page 75 of 374 LIST OF FIGURES (Continued)

Number Title 15.6-61m Deleted Replace with new Figures 15.6-61n Deleted 15A-1 Release Pathways 15A-2 Allowable Inleakage Values 15A-3 Control Room Model

ULNRC-06636 Page 76 of 374 Spent fuel cask drop accidents are not applicable to Callaway Plant. The spent fuel cask handling equipment has been upgraded to single-failure-proof status to provide the maximum practical defense in depth in accordance with NUREG-0612 and to allow the use of the spent fuel cask handling equipment and lifting devices to handle heavy loads in the vicinity of spent fuel without the need for load drop analyses. This is supported by NRC Information Notice 99-15 which stated in general that for cask movements with single-failure-proof cranes, cask drops or tipping accidents need not be considered.

Since the cask cannot drop, no cask rupture can occur and thus no radioactivity can be released.

15.0.1.4 Condition IV - Limiting Faults Condition IV occurrences are faults which are not expected to take place, but are postulated because their consequences would include the potential for the release of significant amounts of radioactive material. They are the most drastic which must be designed against and represent limiting design cases. Condition IV faults are not to cause a fission product release to the environment resulting in an undue risk to public health and safety in excess of guideline values of 10 CFR 100. A single Condition IV fault is not to cause a consequential loss of required functions of systems needed to cope with the fault, including those of the emergency core cooling system and the containment. For the purposes of this report the following faults have been classified in this category:

a. Steam system pipe break. 50.67 or Regulatory Guide 1.183

b. Feedwater system pipe break.

c. Reactor coolant pump shaft seizure (locked rotor).

d. Reactor coolant pump shaft break.

e. Spectrum of rod cluster control assembly ejection accidents.

f. Steam generator tube rupture.

g. Loss-of-coolant accidents, resulting from a spectrum of postulated piping breaks within the reactor coolant pressure boundary (large break).

h. Design basis fuel handling accidents.

15.0.2 OPTIMIZATION OF CONTROL SYSTEMS 15.0.2.1 Setpoint Study A control system setpoint study is performed in order to simulate performance of the reactor control and protection systems. In this study, emphasis is placed on the

ULNRC-06636 Enclosure 6 Page 77 of 374 The secondary power is obtained from the measurement of steam or feedwater flow, feedwater inlet temperature to the steam generators, and steam pressure. High accuracy instrumentation is provided for use during these measurements. Accuracy tolerances meet or exceed requirements established by the safety analysis.

15.0.8 PLANT SYSTEMS AND COMPONENTS AVAILABLE FOR MITIGATION OF ACCIDENT EFFECTS The plant is designed to afford protection against the possible effects of natural phenomena, postulated environmental conditions, and dynamic effects of the postulated accidents. In addition, the design incorporates features which minimize the probability and effects of fires and explosions. The Operating Quality Assurance Manual discusses the quality assurance program which has been implemented to assure that the NSSS will satisfactorily perform its assigned safety functions. The incorporation of these features in the plant, coupled with the reliability of the design, ensures that the normally operating systems and components listed in Table 15.0-6 will be available for mitigation of the events discussed in Chapter 15.0. In determining which systems are necessary to mitigate the effects of these postulated events, the classification system of ANSI-N18.2-1973 is utilized. The design of safety-related systems (including protection systems) is consistent with IEEE Standard 379-1972 and Regulatory Guide 1.53, in the application of the single failure criterion.

In the analysis of the Chapter 15.0 events, control system action is considered only if that action results in more severe accident results. No credit is taken for control system operation if that operation mitigates the results of an accident. For some accidents, the analysis is performed both with and without control system operation to determine the worst case. The pressurizer heaters are generally assumed not to be energized for the analysis of the Chapter 15 events. Operation of the pressurizer heaters as a result of normal control action or a single failure will be less conservative or have negligible effects for most analyses. Therefore, unless it is shown that such a control action results in more limiting results or more severe consequences, the control action of the pressurizer heaters is not modeled for the analyses performed in Chapter 15. Any exceptions are noted in the text describing the individual analysis assumptions.

15.0.9 FISSION PRODUCT INVENTORIES The calculation of the core fission product inventory employs the ORIGEN 2 computer code modelling a three region enveloping cycle core with a core power level of 3636 MWt (3565 MWt plus 2% postulated calorimetric error). Of the 96 assemblies in core Region 1, 32 have operated at a specific power of 50.7 MW/MTU for 474 days and 64 have operated at a specific power of 57.0 MW/MTU for 474 days. Of the 88 assemblies in core Region 2, 24 have operated at a specific power of 48.5 MW/MTU for 474 days and at 44.3 MW/MTU for 474 days and 64 have operated at a specific power of 52.7 MW/MTU for 474 days and at 33.8 MW/MTU for 474 days. The 9 assemblies in core Region 3 have operated at a specific power of 50.6 MW/MTU for 474 days, at 38.0 MW/

MTU for 474 days, and at 21.1 MW/MTU for 474 days. The average burnups in Regions Insert 15.0.9

ULNRC-06636 Page 78 of 374 1, 2, and 3 at the end of a cycle (MWD/MTU) are 26,000, 42,000, and 52,000, respectively. The isotopic yields utilize data for fissioning of U-235, U-238, and Pu-239 and account for the depletion of U-235. Radiological consequences are evaluated with source terms based on the 3636 MWt core rating (Table 15A-3), Callaway-specific meteorology based on three years of combined meteorological data (Table 15A-2), and appropriate dose conversion factors (Table 15A-4).

15.0.10 RESIDUAL DECAY HEAT 15.0.10.1 Total Residual Heat Residual heat in a subcritical core is calculated for the LOCA per the requirements of Appendix K of 10 CFR 50.46, as described in References 11 and 12. These requirements include assuming infinite irradiation time before the core goes subcritical to determine fission product decay energy. For all other accidents, the same models are used, except that fission product decay energy is based on core average exposure at the end of the equilibrium cycle.

15.0.10.2 Distribution of Decay Heat Following Loss-of-Coolant Accident During a LOCA, the core is rapidly shut down by void formation or RCCA insertion, or both, and a large fraction of the heat generation to be considered comes from fission product decay gamma rays. This heat is not distributed in the same manner as steady state fission power. Local peaking effects which are important for the neutron dependent part of the heat generation do not apply to the gamma ray contribution. The steady state factor of 97.4 percent, which represents the fraction of heat generated within the clad and pellet, drops to 95 percent for the hot rod in a LOCA.

For example, consider the transient resulting from the postulated double ended break of the largest reactor coolant system pipe; 1/2 second after the rupture about 30 percent of the heat generated in the fuel rods is from gamma ray absorption. The gamma power shape is less peaked than the steady state fission power shape, reducing the energy deposited in the hot rod at the expense of adjacent colder rods. A conservative estimate of this effect is a reduction of 10 percent of the gamma ray contribution or 3 percent of the total. Since the water density is considerably reduced at this time, an average of 98 percent of the available heat is deposited in the fuel rods, the remaining 2 percent being absorbed by water, thimbles, sleeves, and grids. The net effect is a factor of 0.95, rather than 0.974, to be applied to the heat production in the hot rod.

15.0.11 COMPUTER CODES UTILIZED Summaries of some of the principal computer codes used in transient analyses are given below. The codes used in the analyses of each transient have been listed in Table 15.0-2.

ULNRC-06636 Page 79 of 374 local condition heat transfer. Component models include a two region nonequilibrium pressurizer, centrifugal and jet pumps, valves, non-conducting heat exchangers, steam separators, and turbine. An automatic steady state initialization procedure is also available. Insert 15.0.11.8 References 18 and 23 The RETRAN code is discussed in Reference 18.

15.0.11.9 VIPRE RETRAN-02 and RETRAN-3D codes are The VIPRE computer program performs thermal-hydraulic calculations. The code calculates coolant density, mass velocity, enthalpy, void fractions, static pressure and DNBR distributions along flow channels within a reactor core.

The VIPRE code is described in Reference 19.

15.0.11.10 ANC ANC is an advanced nodal code capable of two-dimensional and three-dimensional neutronics calculations. ANC is the reference model for certain safety analysis calculations, power distributions, peaking factors, critical boron concentrations, control rod worths, reactivity coefficients, etc. In addition, three-dimensional ANC validates one-dimensional and two-dimensional results and provides information about radial (x-y) peaking factors as a function of axial position. It can calculate discrete pin powers from nodal information as well.

The ANC code is described in Reference 20.

15.0.12 LIMITING SINGLE FAILURES The most limiting single failure as described in Section 3.1 of safety-related equipment, where one exists, is identified in each analysis description, and the consequences of this failure are described therein. In some instances, because of redundancy in protection equipment, no single failure which could adversely affect the consequences of the transient has been identified. The failure assumed in each analysis is listed in Table 15.0-7.

15.0.13 OPERATOR ACTIONS For most of the events analyzed in Chapter 15.0 the plant will be in a safe and stable hot standby condition following the automatic actuation of reactor trip. This condition will, in fact, be similar to plant conditions following any normal, orderly shutdown of the reactor.

At this point, the actions taken by the operator would be no different than normal operating procedures. The exact actions taken, and the time at which these actions would occur, will depend on what systems are available (e.g., turbine bypass system, main feedwater system, etc.) and the plans for further plant operation. As a minimum, to maintain the hot stabilized condition, decay heat must be removed via the steam

ULNRC-06636 Page 80 of 374

1) Each Level channel is tested one-at-a-time during the level channel testing with zero time delay as describd in the WCAP.

2) The TTD function and timers discussed in Reference 15 are no longer applicable in Callaway.

3) Section 3.6.2.2 is titled OUTAGE TESTING. The PROM logic modules and EAM testing described under this section may be performed on-line and not restricted to performance during outages.

16. RETRAN-02 -- A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems," Electric Power Research Institute, EPRI NP-1850-CCM-A, Rev. 2, 1984.

17. Letter from Cecil O. Thomas (NRC) to Dr. Thomas W. Schnatz, Utility Group for Regulatory Applications (UGRA), "Acceptance for Referencing of Licensing Topical Reports EPRI CCM-5, 'RETRAN - A Program for One Dimensional Transient Thermal Hydraulic Analysis of Complex Fluid Flow Systems,' and EPRI NP-1850-CCM, 'RETRAN A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems,'" dated September 2, 1984.

18. D.S. Huegel, et. al., WCAP-14882-P-A (Proprietary)/WCAP-15234-A (Non-proprietary), RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analysis, April 1999.

19. Y.X. Sung, et. al., WCAP-14565-P-A (Proprietary)/WCAP-15306-A (Non-proprietary), VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis, October 1999.

20. Y.S. Liu, et. al., WCAP-10965-P-A, ANC: A Westinghouse Advanced Nodal Compter Code, September 1986.

21. Westinghouse Letter SCP-07-17, Callaway Plant Engineering Report and Guidelines in Support of End of Cycle 15 Tavg Coastdown, Revision 1, dated February 9, 2007.

22. ORNL/TM-2005/39, Oak Ridge National Lab, SCALE 6.1 Package Manual, June 2011.

23. M. P. Paulsen et. al., NP--7450(A), "RETRAN-3D -- A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems," , Research Project 889-10, EPRI, Rev. 11, May 2017

ULNRC-06636 Enclosure 6 Page 81 of 374 CALLAWAY - SP TABLE 15.0-2 (Sheet 6)

REACTOR REACTOR PRESSURIZER EQUIVALENT FULL COOLANT VESSEL PRESSURIZER WATER S/G TUBE POWER PUMP HEAT COOLANT VESSEL PRESSURE LEVEL FEEDWATER PLUGGING STEADY EVENT (MWt) FLOW (gpm) T-AVG (°F) (PSIA) (% span) TEMP (°F) LEVEL STATE FH FQ 15.5 Increase in coolant inventory Inadvertent ECCS 20 382,630 588.4 2250 65 446 5% NA NA operation at power (DNB Case)

(Pzr. Filling Case) 20 374,400 567.2 2190 43 446 5% NA NA CVCS malfunction See Section 15.5.2 for all asumptions 15.6 Decrease in coolant inventory Inadvertent RCS 14 382,630 588.4 2250 60 446 5% NA NA Overfill depressurization S/G tube rupture ASD faillure case 14 374,400 592.7 2280 60 446 0% NA NA Overall case 14 374,400 567.7 2280 38 390 5% NA NA Loss of coolant accidents See Section 15.6.5 for all asumptions

++++RETRAN Option 1 Film Boiling Correlation NOTES: 1. Deleted

2. 2250 psia used in offsite dose evaluation Rev. OL-22 11/16

ULNRC-06636 Enclosure 6 Page 82 of 374 TABLE 15.0-7 SINGLE FAILURES ASSUMED IN ACCIDENT ANALYSES Event Description Worst Failure Assumed Feedwater temperature reduction One protection train Excessive feedwater flow One protection train Excessive steam flow (1)(3)

Inadvertent secondary depressurization (3)

Steam system piping failure One safety injection train Steam pressure regulator malfunction (2)

Loss of external load One protection train Turbine trip One protection train Inadvertent closure of MSIV One protection train Loss of condenser vacuum One protection train Loss of ac power Turbine driven auxiliary feedwater pump Loss of normal feedwater Turbine driven auxiliary feedwater pump Feedwater system pipe break Turbine driven auxiliary feedwater pump Partial loss of forced reactor coolant flow One protection train Complete loss of forced reactor coolant flow One protection train RCP locked rotor One protection train RCP shaft break One protection train RCCA bank withdrawal from subcritical One protection train RCCA bank withdrawal at power One protection train Dropped RCCA, dropped RCCA bank (1)

Statically misaligned RCCA (3)

Single RCCA withdrawal One protection train Inactive RC pump startup (3)

Flow controller malfunction (2)

Uncontrolled boron dilution Standby ECCS charging pump is operating (Modes 1 and 2), one source range NIS channel (Modes 3-5)

Improper fuel loading (3)

RCCA ejection One protection train

, ASD failure nadvertent ECCS operation at power One protection train I

Increase in RCS inventory One pressurizer level channel BWR transients (2)

Inadvertent RCS depressurization One protection train Failure of small lines carrying primary (3) coolant outside containment SGTR One SG atmospheric steam dump valve BWR piping failures (2)

SGTR, Overfill Auxiliary Feedwater Flow Control Valve Rev. OL-2 1 5/15

ULNRC-06636 Page 83 of 374 15.1.5 STEAM SYSTEM PIPING FAILURE 15.1.5.1 Identification of Causes and Accident Description The steam release arising from a rupture of a main steamline would result in an initial increase in steam flow that decreases during the accident as the steam pressure falls.

The energy removal from the RCS causes a reduction of coolant temperature and pressure. In the presence of a negative moderator temperature coefficient, the cooldown results in an insertion of positive reactivity. If the most reactive RCCA is assumed stuck in its fully withdrawn position after reactor trip, there is possibility that the core will become critical and return to power. A return to power following a steamline rupture is a potential problem mainly because of the high power peaking factors which exist, assuming the most reactive RCCA to be stuck in its fully withdrawn position. The core is ultimately shut down by50.67 andacid the boric Regulatory solutionGuide 1.183 delivered by the ECCS.

The analysis of a main steamline rupture is performed to demonstrate that the following criteria are satisfied:

Assuming a stuck RCCA with or without offsite power, and assuming a single failure in the engineered safety features, the core remains in place and intact.

Radiation doses do not exceed the guidelines of 10 CFR 100.

Although DNB and possible clad perforation following a steam pipe rupture are not necessarily unacceptable, the following analysis, in fact, shows that no DNB occurs for any rupture, assuming the most reactive RCCA stuck in its fully withdrawn position. The DNBR design basis is discussed in Section 4.4.

A major steamline rupture is classified as an ANS Condition IV event. See Section 15.0.1 for a discussion of Condition IV events.

Effects of minor secondary system pipe breaks are bounded by the analysis presented in this section. Minor secondary system pipe breaks are classified as Condition III events, as described in Section 15.0.1.3.

The major rupture of a steamline is the most limiting cooldown transient, and is analyzed at zero power with no decay heat. Decay heat would retard the cooldown, thereby reducing the return to power. A detailed analysis of this transient with the most limiting break size, a double ended rupture, is presented here. The assumptions used in this analysis are discussed in Reference 3. Reference 3 also contains a discussion of the spectrum of break sizes and power levels analyzed.

During startup or shutdown evolutions when safety injection on low pressurizer pressure or low steamline pressure is blocked and steamline isolation on low steamline pressure is blocked below P-11 (pressurizer pressure less than 1970 psig), the high negative steamline pressure rate (HNPR) signal is enabled by P-11 to provide steamline isolation. A series of steamline break sensitivities in Mode 3 conditions has been

ULNRC-06636 Page 84 of 374 Margin to Critical Heat Flux A DNB analysis was performed for both of these cases. It was found that both cases had a minimum DNBR greater than the safety analysis limit value as discussed in Section 4.4.1.1. The WLOP DNB correlation was used in this analysis (Reference 12).

Historically, the W-3 DNB correlation had been used; see Reference 5 for the justification discussing the use of the W-3 correlation for low pressure applications, accepted by the NRC in Reference 6.

15.1.5.3 Radiological Consequences ;however, the Reactor Coolant Pumps are assumed to remain on for the steam 15.1.5.3.1 Method Of Analysis release input to radiological consequence analysis.

15.1.5.3.1.1 Physical Model The radiological consequences of a MSLB inside the containment are less severe than the one outside the containment because the radioactivity released will be held up inside the containment, allowing decay and plateout of the radionuclides. To evaluate the radiological consequences due to a postulated MSLB (outside the containment), it is assumed that there is a complete severance of a main steamline outside the containment.

It is also assumed that there is a simultaneous loss of offsite power, resulting in reactor coolant pump coastdown. The ECCS is actuated and the reactor trips.

The main steam isolation valves, their bypass valves, and the steamline drain valves isolate the steam generators and the main steamlines upon a signal initiated by the engineered safety features actuation system under the conditions of high negative steamline pressure rates, low steamline pressure, or high containment pressures. The main steam isolation valves are installed in the main steamlines from each steam generator downstream from the safety and atmospheric relief valves outside the containment. The break in the main steamline is assumed to occur outside of the containment. The affected steam generator (steam generator connected to a broken steamline) blows down completely. The steam is vented directly to the atmosphere.

Each of the steam generators incorporates integral flow restrictors, which are designed to limit the rate of steam blowdown from the steam generators following a rupture of the main steamline. This, in turn, reduces the cooling rate of the reactor coolant system thereby reducing the return to power.

In case of loss of offsite power, the remaining steam generators are available for dissipation of core decay heat by venting steam to the atmosphere via the atmospheric relief and safety valves. Venting continues until the reactor coolant temperature and pressure have decreased sufficiently so that the RHR system can be utilized to cool the reactor. The MSSVs release steam at their individual set pressures and are not capable of supporting a controlled plant cooldown to RHR entry conditions. Although the MSSVs

ULNRC-06636 Enclosure 6 Page 85 of 374 would be available following a MSLB for SG over-pressurization protection if needed, they do not have a safety function to mitigate a MSLB or to cool down the plant. Plant cooldown to RHR entry conditions is supported by the ASDs. Section 10.3.3 SAFETY EVALUATION SEVEN provides more details.

15.1.5.3.1.2 Assumptions and Conditions The major assumptions and parameters assumed in the analysis are itemized in Tables 15.1-3 and 15A-1. Tables 15B-1 and 15B-4 provide a comparison of the analysis to the guidelines in Regulatory Guide 1.183.

The assumptions used to determine the concentrations of radioactive isotopes within the secondary system for this accident are as follows:

a. The initial secondary side radio-iodine concentrations are assumed tobe 10% of the initial Case 1 primary sideconcentrations.

b. A primary-to-secondary leakage rate of 1 gpm is assumed to exist and is assumed to be in the affected steam generator.

faulted

c. The reactor coolant initial iodine activity is determined by two methods, and both cases are analyzed. These are:

Insert 15.1.5.3.1.2A Case 1 - The Case 1 initial radio-iodine concentrations are the same as the Case 1 concentrations used for the Steam Generator Tube Rupture accident sequence. Refer to Table 15.6-4.

Case 2 - The Case 2 initial radio-iodine concentrations are the same as the Case 2 concentrations used for the Steam Generator Tube Insert 15.1.5.3.1.2B Rupture accident sequence. Refer to Table 15.6-4.

d. The initial reactor coolant concentrations of noble gas correspond to 1-percent fuel defects. 225 µCi/gm DOSE EQUIVALENTXE-133).

e. Partition factors used to determine the secondary system activities are given in Table 15.1-3.

for iodines and alkali metals The following specific assumptions and parameters are used to calculate the activity release:

a. Offsite power is lost, resulting in reactor coolant pump coastdown.

b. No condenser air removal system release and no normal operating steam Approximately generator blowdown is assumed to occur during the course of the accident.

eight (7.899)

c. Eight hours after the occurrence of the accident, the residual heat-removal system (RHRS) starts operation to cool down the plant.

however, the Reactor Coolant Pumps are assumed to remain on for the steam release input to radiological consequence analysis.

ULNRC-06636 Enclosure 6 Page 86 of 374 CALLAWAY - SP 22 cooled to 212oF such that there would be no flashing of the leaked fluid

d. After the accident, the primary-to-secondary leakage continues for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, at which time the reactor coolant system is depressurized.

faulted

e. The affected steam generator (steam generator connected to the broken steamline) is allowed to blow down completely. approximately 8 on the intact steam generators

f. Steam release to the atmosphere and the associated activity release from the safety and relief valves and the broken steamline is terminated 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the accident, when the RHRS is activated to complete cooldown.

and alkali metal faulted

g. The amount of noble gas activity released is equal to the amount present in the reactor coolant, which leaks to the secondary during the accident. The amount of iodine activity released is based on the activity present in the secondary system and the amount of leaked reactor coolant which is entrained in the steam that is discharged to the environment via the safety and relief valves and the broken steamline. Partition factors used for the intact unaffected steam generators after the accident occurs are given in Table 15.1-3. An iodine partition factor of 1 is used for the affected steam generator. After the break, all primary-to-secondary leakage is assumed to be through the faulted steam generator.

h. The activity released from the broken steamline and the safety and relief valves during the 8-hour duration of the accident is immediately vented to the atmosphere.

15.1.5.3.1.3 Mathematical Models Used in the Analysis Mathematical models used in the analysis are described in the following sections:

a. The mathematical models used to analyze the activity released during the course of the accident are based on the assumptions listed above.

b. The atmospheric dispersion factors used in the analysis were calculated based on the onsite meteorological measurement programs described in Section 2.3 of the Site Addendum. TEDE ,

c. The thyroid inhalation dose and total-body gamma immersion doses to a receptor at the exclusion area boundary and outer boundary of the low-population zone were analyzed, using the models described in Appendix 15A. , and in the control room 15.1.5.3.1.4 Identification of Leakage Pathways and Resultant Leakage Activity For evaluating the radiological consequences due to a postulated MSLB, the activity released from the affected steam generator (steam generator connected to the broken steamline) is released directly to the environment. The unaffected steam generators are faulted intact

ULNRC-06636 Enclosure 6 Page 87 of 374 assumed to continually discharge steam and entrained activity via the safety and relief valves up to the time initiation of the RHRS can be accomplished.

Since the activity is released directly to the environment with no credit for plateout or retention, the results of the analysis are based on the most direct leakage pathway available. Therefore, the resultant radiological consequences represent the most conservative estimate of the potential integrated dose due to the postulated MSLB.

15.1.5.3.2 Identification of Uncertainties and Conservatisms in the Analysis iodine and noble gas

a. Reactor coolant activities are based on an initial radio-iodine spectrum that would conservatively bound those found in either open or tight type fuel defects. Tight fuel defects tend to produce limiting results for thyroid does, while open fuel defects tend to produce limiting results for whole body dose. The assumed concentrations of longer-lived isotopes represent the values that would be reached in the presence of tight fuel defects. The assumed concentractions of shorter-lived isotopes represent the values that would be reached in the presence of open fuel defects. Since the assumed iodine spectrum represents bounding values for different types of fuel defects, the initial radio-iodine inventory would exceed the Technical Specification limit of 1.0 µCi/gm. Additionally, large spiking factors are assumed in the analysis.

b. A 1-gpm steam generator primary-to-secondary leakage is assumed, which is significantly greater than that anticipated during normal operation.

Furthermore, it was conservatively assumed that all leakage is to the faulted affected steam generator only.

1 gpm

c. The meteorological conditions which may be present at the site during the course of the accident are uncertain. However, it is highly unlikely that the Insert 15.1.5.3.2 assumed meteorological conditions would be present during the course of the accident for any extended period of time. Therefore, the radiological consequences evaluated, based on the meteorological conditions assumed, are conservative.

15.1.5.3.3 Conclusions 15.1.5.3.3.1 Filter Loadings The only ESF filtration system considered in the analysis which limits the consequences of the MSLB is the control room filtration system. Activity loadings on the control room charcoal filter are based on flow rate through the filter, the concentration of activity at the filter inlet, and the filter efficiency.

Activity in the control room filter as a function of time has been evaluated for the more limiting LOCA analysis, as discussed in Section 15.6.5.4.3.1. Since the control room

ULNRC-06636 Enclosure 6 Page 88 of 374 filters are capable of accommodating the potential design basis LOCA fission product iodine loadings, more than adequate design margin is available with respect to postulated MSLB releases. ,

15.1.5.3.3.2 Dose to Receptor at the Exclusion Area Boundaryand Low-Population Zone Outer Boundary and in the Control Room worst 2-hour time period TEDE doses , assumptions The potential radiological consequences resulting from the occurrence of a postulated MSLB have been conservatively analyzed, using asusmptions and models described.

The total-body gamma doses due to immersion from direct radiation and the thyroid dose due to inhalation have been analyzed for the 0-2 hour dose at the exclusion area 22 boundary and for the duration of the accident (0 to 8 hrs) at the low-population zone outer boundary. The results are listed in Table 15.1-4. The resultant doses are well within the guideline values of 10 CFR 100. 50.67 and Regulatory Guide 1.183 for offsite

, and in the Control and the GDC 19 limit for the Control Room 15.1.5.4 Conclusions Room The analysis has shown that the criteria stated earlier in Section 15.1.5.1 are satisfied.

Although DNB and possible clad perforation following a steam pipe rupture are not necessarily unacceptable and not precluded by the criteria, the above analysis shows that the DNB design basis is met for any rupture, assuming the most reactive RCCA stuck in its fully withdrawn position.

A safety evaluation was performed to determine the impact of a potential increase in the stroke time of the feedwater isolation valves beyond the value assumed in the analyses (15 seconds) due to the installation of new valve actuators. It was concluded that the results presented in this section for the zero power steamline break event are not adversely affected by this plant modification. As such, the reported results and conclusions remain valid.

15.1.5.5 Steam Line Break with Coincident Control Rod Withdrawal This accident is no longer applicable to Callaway since automatic rod withdrawal is no longer available.

15.1.5.6 Steam System Piping Failure at Full Power 15.1.5.6.1 Identification of Causes and Accident Description A Steamline Rupture - Full Power Core Response transient is defined as a break that results in an increase in steam flow from one or more steam generators. A Steamline Rupture can result from:

• An inadvertent opening of a steam generator dump, safety or relief valve

ULNRC-06636 Enclosure 6 Page 89 of 374 TABLE 15.1-3 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A MAIN STEAM LINE BREAK 34 I. Source Data:

a. Core power level, Mwt 3636

b. Steam generator tube 1 leakage, gpm

c. Reactor coolant initial iodine activity:

1) Case1 The MSLB Case 1 initial radio-iodine inventory is the same as the Case 1 SGTR initial radio-iodine inventory. Refer to Table Insert T15.1-3A 15.6-4.

2) Case 2 The MSLB Case 2 initial radio-iodine inventory is the same as the Case 2 SGTR initial radio-iodine inventory. Refer to Table 15.6-4.

d. Reactor coolant initial noble gas activity:

1) Case1 Based on 1-percent fuel defects as provided in Table 15A-5 (225 µCi/gm DOSE EQUIVALENT XE-133)

2) Case 2 Based on 1-percent fuel defects as Insert T15.1-3B provided in Table 15A-5 (225 µCi/gm DOSE EQUIVALENT XE-133) f.e. Secondary system initial 10% of Case 1 primary side activity iodine activity 5.51E+5 g.

f. Iodine partition factors

1) Affected steam generator Faulted 1.0

2) Unaffected steam Intact 0.01 generator

i. g. Reactor coolant mass, lbs 5.50E+5

j. h. Steam generator mass

1) Affected steam generator, 1.555E+5 Faulted lbs Insert T15.1-3C

2) Each unaffected steam 1.555E+5 generator, lbs II. Atmospheric Dispersion Factors See Table 15A-2 Intact

ULNRC-06636 Page 90 of 374 TABLE 15.1-4 RADIOLOGICAL CONSEQUENCES OF A MAIN STEAM LINE BREAK (TEDE)

Doses (rem)

Control Room (30 days)

CASE 1, accident initiated iodine spike Exclusion area boundary (0-2 hr) 5.54E-1 Thyroid 6-8 hours 4.9E00 Whole body 1.3E-01 Low-population zone outer boundary (duration) 4.97E-1 Thyroid 6.1E00 Whole body 8.8E-02 2.81 CASE 2, pre-accident iodine spike 1.11E-1 Exclusion area boundary (0-2 hr)

Thyroid 4.0E00 Whole body 2.5E-02 1.03E-1 Low population zone outer boundary (duration)

Thyroid 1.3E00 Whole body 1.3E-02 2.16 Control Room (30 days)

ULNRC-06636 Page 91 of 374 As such, the assumptions used in this analysis are designed to minimize the energy removal capability of the system and to maximize the possibility of water relief from the coolant system by maximizing the coolant system expansion, as noted in the assumptions listed above.

The assumptions used in the analysis are essentially identical to the loss of normal feedwater flow incident (Section 15.2.7), except that power is assumed to be lost to the reactor coolant pumps at the time of reactor trip.

Results The transient response of the RCS following a loss of ac power with pressurizer PORVs unavailable is shown in Figures 15.2-9 through 15.2-11. The calculated sequence of events for this transient is listed in Table 15.2-1. The first few seconds after the loss of power to the reactor coolant pumps will closely resemble a simulation of the complete loss of flow incident (see Section 15.3.2); i.e., core damage due to rapidly increasing core temperatures is prevented by promptly tripping the reactor. After the reactor trip, stored and residual decay heat must be removed to prevent damage to either the RCS or the core.

The RETRAN code results show that the natural circulation flow available is sufficient to provide adequate core decay heat removal following reactor trip and RCP coast-down.

A separate case was run with high head ECCS charging pumps initiated on a loss of offsite power signal (see assumption (l) above). This case did result in the filling of the pressurizer. However, this occurred sufficiently late in the transient such that operator action to unblock both pressurizer power-operated relief valves could be credited to preclude water relief through the pressurizer safety valves. This action was assumed to occur 9 minutes following the loss of offsite power while pressurizer filling occurred well after this time. This case is analyzed similar to the Inadvertent ECCS at Power event, discussed in Section 15.5.1, where operator action is required to unblock the pressurizer power-operated relief valves thereby precluding water relief through the pressurizer safety valves.

15.2.6.3 Radiological Consequences 15.2.6.3.1 Method of Analysis 15.2.6.3.1.1 Physical Model The dose calculation for loss of ac power is based on the sequence of events described in Table 15.2-1. It is assumed that heat removal from the nuclear steam supply system is achieved by venting the steam for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

7.3 The reactor coolant is assumed to be contaminated by radioactive fission products introduced through fuel cladding defects. The secondary system is contaminated by the inleakage of reactor coolant through postulated steam generator tube leaks.

ULNRC-06636 Enclosure 6 Page 92 of 374 The radioactivity in the vented steam is dispersed in the atmosphere without any reduction due to plateout, fallout, filtering, etc.

15.2.6.3.1.2 Assumptions and Conditions The major assumptions and parameters assumed in the analysis are found in Tables 15.2-2 and 15.A-1. The assumptions used to determine the activity released are as follows: Insert 15.2.6.3.1.2A

a. The reactor coolant initial iodine activity assumed is the Technical Specification limit of 1.0 µCi/gm I-131 dose equivalent (adjusted consistent with Table 15.6-4 item I.c.1). 10%

Insert 15.2.6.3.1.2B d.b. The initial secondary system iodine activity assumed is 1/10 of the initial reactor coolant iodine activity.

f. c. A 1-gpm steam generator primary-to-secondary leakage is assumed for the duration of steam venting.

gd. For noble gases, the activity released is taken to be the activity introduced Insert by reactor coolant 15.2.6.3.1.2C inleakage without holdup in the steam system.

h.e. The partition factor for iodine in the steam generators is taken as 0.01 for secondary side releases and 1.0 for iodine in primary-to-secondary leakage. This assumption is conservatively based on a leak in the upper tubes which are assumed to be uncovered for the accident duration.

j.f. The atmospheric dispersion factors are given in Table 15A-2.

15.2.6.3.1.3 Mathematical Models Used in the Analysis Mathematical models used in the analysis are described in the following sections:

i. The alkali metal activity present in the primary-to-secondary leakage is assumed a.

to be homogeneously The mathematical modelsGenerator mixed in the Steam used to analyze the activity released durin g the inventory.

course of the accident are described in Appendix 15A.

b. The atmospheric dispersion factors used in the analysis were calculated using the onsite meteorological measurement programs described in Section 2.3 of the Site Addendum. and in the control room

c. The thyroid inhalation and total-body immersion doses to a receptor at the exclusion area boundary or outer boundary of the low population zone were analyzed using the models described in Appendix 15A.

TEDE ,

ULNRC-06636 Enclosure 6 Page 93 of 374 CALLAWAY - SP 15.2.6.3.1.4 Identification of Leakage Pathways and Resultant Leakage Activities Normal activity paths from the secondary system, such as the condenser air removal system and steam generator blowdown, cease during loss of ac power. The steam is released to the atmosphere through the:

a. Power-operated atmospheric relief valves

b. Reactor coolant activities based on extreme

b. Main steam safety valves iodine spiking effects are conservatively high.

Since all these paths are taken as direct to the atmosphere without any form of decontamination, they are all radiologically equivalent and need not be distinguished.

15.2.6.3.2 Identification of Uncertainties in, and Conservative Aspects of, the Analysis iodine and noble gas The principal uncertainties in the dose calculation arise from the uncertainties in the accident circumstances, particularly the extent of steam contamination, the weather at the time, and delay before preferred ac power is restored. Each of these uncertainties is s handled by making very conservative or worst-case assumptions.

are

a. Reactor coolant activities are based on the Technical Specification limit, which is significantly higher than the activities associated with normal operating conditions, based on 0.12-percent failed fuel.

c.

b. A 1-gpm steam generator primary-to-secondary leakage is assumed, which is significantly greater than that anticipated during normal operation.

c.

d. The meteorological conditions which may be present at the site during the course of the accident are uncertain. However, it is highly unlikely that the assumed meteorological conditions would be present during the course of the accident for any extended period of time. Therefore, the evaluated radiological consequences, based on the meteorological conditions assumed, will be conservative.

e. Steam dump to the condenser is not available 15.2.6.3.3 Conclusions 15.2.6.3.3.1 Filter Loadings No filter serves to limit the release of radioactivity in this accident. There is no significant activity buildup oninany and thefilters as Room Control a consequence of loss of ac power. ,

15.2.6.3.3.2 Doses to Receptor at Exclusion Area Boundary and Low Population Zone Outer Boundary worst The maximum doses to an individual who spends the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after loss of ac power at the exclusion area boundary, and the maximum doses for a long-term exposure (8

ULNRC-06636 Page 94 of 374 and in the control CALLAWAY - SP room for 30 days hours or longer) at the outer boundary of the low-population zone, are given in Table 15.2-3. These doses are very small compared with the guideline values of 10 CFR 100.

15.2.6.4 Conclusions Results of the analysis show that, for the loss of non-emergency ac power to plant auxiliaries event, all safety criteria are met. Auxiliary feedwater capacity is sufficient to prevent water relief through the pressurizer relief and safety valves; this assures that the RCS is not overpressurized.

50.67 for offsite dose and the GDC 19 Analysis of the natural circulation capability of thefor limit RCS the demonstrates Control Room that g sufficient lon term heat removal capability exists followin g reactor coolant pump coastdown to preve nt fuel or clad damage.

15.2.7 LOSS OF NORMAL FEEDWATER FLOW 15.2.7.1 Identification of Causes and Accident Description A loss of normal feedwater, from pump failures, valve malfunctions, or loss of offsite ac power, or feedwater control system failure, results in a reduction in the capability of the secondary system to remove the heat generated in the reactor core. If an alternative supply of feedwater were not supplied to the plant, core residual heat following reactor trip would heat the primary system water to the point where water relief from the pressurizer would occur, resulting in a substantial loss of water from the RCS. Since the plant is tripped well before the steam generator heat transfer capability is reduced, the primary system variables never approach a DNB condition.

The reactor trip on low-low water level in one or more steam generators provides the necessary protection against a loss of normal feedwater.

The following occur upon loss of normal feedwater (assuming main feedwater pump failures or valve malfunctions):

a. As the steam system pressure rises following the trip, the steam generator power-operated relief valves are automatically opened to the atmosphere.

Steam dump to the condenser is assumed not to be available. If the steam flow through the power-operated relief valves is not available, the steam generator safety valves may lift to dissipate the stored thermal energy of the reactor coolant system and fuel plus the residual decay heat produced in the reactor.

b. As the no-load temperature is approached, the steam generator power-operated relief valves (or safety valves, if the power-operated relief valves are not available) are used to dissipate the residual decay heat and to maintain the plant at the hot shutdown condition.

ULNRC-06636 Enclosure 6 Page 95 of 374

a. Core Power level, MWt 3636 (includes CALLAWAY - SP 2% uncertainty)

TABLE 15.2-2 PARAMETERS USED IN EVALUATING RADIOLOGICAL CONSEQUENCES OF LOSS OF NONEMERGENCY AC POWER I. Source Data b.a. Steam generator tube leakage, gpm 1

b. Reactor coolant initial iodine activity 1.0 µCi/gm of I-131 dose equivalent (adjusted consistent with Insert T15.2-2A Table 15.6-4 item I.c.1) c.

d. Secondary system initial iodine activity Equivalent to 1/10 of the 225 Ci/gm dose equivalent initial RCS iodine activity of Xe-133 10% dose e.d. Reactor coolant initial noble gas activity Based on 1 percent fuel equivalent Portion of primary to secondary leakage defects, as provided in flashed directly to the environment Table 11.1-5 f.e. 5% for the first 2.667 hours0.00772 days <br />0.185 hours <br />0.0011 weeks <br />2.537935e-4 months <br /> Iodine partition factor in the steam 0.01 generators for secondary side releases for liquid primary-to-f.

g. Iodine partition factor in the steam 1.0 secondary leakage in the generators for primary-to-secondary Steam Generators leakage 0.01 i.g. Each steam generator water mass, lb 1.032E+5 92.5E+3 II. Atmospheric Dispersion Factors See Table 15A-2

h. Iodine and alkali metal partition factor for steam release from Steam Generator

ULNRC-06636 Page 96 of 374 CALLAWAY CASE 1, Accident initiated iodine spike - SP TABLE 15.2-3 RADIOLOGICAL CONSEQUENCES OF LOSS OF NONEMERGENCY AC POWER TEDE 0.7 - 2.7 hrs Doses (rem)

Exclusion area boundary (0-2 hr) 1.10E-02 Thyroid 6.2E-02 Whole body 4.9E-04 Low-population zone, outer boundary (duration) 5.56E-03 Thyroid 2.9E-02 Whole body 30 day 1.4E-04 1.50 Control Room (30 days)

CASE 2, Pre-accident iodine spike Exclusion area boundary (0.6 - 2.6 hrs) 5.31E-03 Low-population zone, outer boundary (duration) 2.95E-03 Control Room (30 days) 0.82

ULNRC-06636 Enclosure 6 Page 97 of 374 CALLAWAY - SP The transient results with and without offsite power available are shown in Figures 15.3-9 through 15.3-12. The results of these calculations are also summarized in Table 15.3-2.

The peak RCS pressure reached during the transient is less than that which would cause stresses to exceed the faulted condition stress limits. Also, the peak clad temperature is considerably less than 2,700°F. The clad temperature was conservatively calculated, assuming that DNB occurs at the initiation of the transient.

The calculated sequence of events is shown on Table 15.3-1. Figure 15.3-9 shows that the core flow rapidly reaches a new equilibrium value. With the reactor tripped, a stable plant condition will eventually be attained. Normal plant shutdown may then proceed.

15.3.3.3 Radiological Consequences The release of radioactivity is assumed to continue until 15.3.3.3.1 Method of Analysis shutdown cooling is in operation and releases from the steam generators have been terminated 15.3.3.3.1.1 Physical Model The instantaneous seizure of a reactor coolant pump rotor results in a reactor trip on a low coolant flow signal. With the coincident loss of offsite power, the condensers are not available, so the excess heat is removed from the secondary system by steam relief through the steam generator safety and relief valves. Steam generator tube leakage is assumed to continue until the pressures in the reactor coolant and secondary systems are equalized. The reactor coolant will contain the gap activities of the fraction of the fuel which undergoes DNB in addition to its assumed equilibrium activity.

Tables 15B-1 and 15B-6 provide a comparison of the analysis to the guidelines in 15.3.3.3.1.2 Regulatory Assumptions and Conditions Guide 1.183.

The major assumptions and parameters used in the analysis are itemized in Tables 15.3-3 and 15A-1 and summarized below.

Insert 15.3.3.3.1.1 The assumption used to determine the initial concentrations of isotopes in the reactor coolant and secondary coolant prior to the accident are as follows:

a. The reactor coolant iodine activity is based on the dose equivalent of 1.0 µCi/gm of I-131 (adjusted consistent with Table 15.6-4 item I.c.1).

b. The noble gas activity in the reactor coolant is based on 1-percent fuel defects.

c. The initial secondary system iodine activity assumed is 1/10 of the initial reactor coolant iodine activity.

is assumed to experience DNB such that The its following conditions are u sed to calculate the activity released.

a. 5 percent of the core gap activity is released to the reactor coolant at the beginning of the accident.

, with adjustment for high burnup fuel and radial power peaking as presented in Table 15.3-3.

None of the fuel is predicted to experience Fuel Centerline Melt 15.3-9 Rev. OL-18 12/10

ULNRC-06636 Enclosure 6 Page 98 of 374 CALLAWAY - SP shutdown cooling is in operation and releases from the steam generators have been terminated. The

b. Offsite power is lost.

c. Following the incident, steam is released to the environment for heat 7.26 removal.

d. Primary-to-secondary leakage continues after the accident for a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. At that time, reactor coolant and secondary system pressures are equalized. Until the pressure equalizes, the leakage rate is assumed to be constant and equal to the rate existing prior to the incident of 1 gpm (500 lbs/hr).

e. Fission products released from the fuel-cladding gap of the damaged fuel rods are assumed to be instantaneously and homogeneously mixed with Insert 15.3.3.3.1.2 the reactor coolant.

f. The noble gas activity released is equal to the amount present in the reactor coolant which leaks into the secondary system after the accident.

g. The partition factor for iodine in the steam generators is taken as 0.01 for secondary side releases and 0.161 for iodine in primary-to-secondary leakage. This assumption is conservatively based on a leak in the upper tubes which are assumed to be uncovered for the accident duration.

i. h. The activity released from the steam generators is immediately vented to the environment. or the control room k.i. No credit is taken for radioactive decay or ground deposition during radioactivity transport to offsite locations.

l.j. Breathing rates, short-term accident atmospheric dispersion factors corresponding to ground level releases, and dose conversion factors are given in Tables 15A-1, 15A-2, and 15A-4.

j. The alkali particulates are conservatively combined with, and treated as, halogens for transport 15.3.3.3.1.3trough theMathematical steam generators.

Models Used in the Analysis Mathematical models used in the analysis are described in the following sections:

a. The mathematical models used to analyze the activity released during the course of the accident are described in Appendix 15A.

b. The atmospheric dispersion factors used in the analysis were calculated based on the onsite meteorological measurement programs described in Section 2.3 of the Site Addendum, and are provided in Table 15A-2.

ULNRC-06636 Page 99 of 374 CALLAWAY - SP

, and in the control room TEDE

c. The thyroid inhalation and total-body immersion doses to a receptor at the exclusion area boundary or outer boundary of the low-population zone were analyzed using the models described in Appendix 15A.

15.3.3.3.1.4 Identification of Leakage Pathways and Resultant Leakage Activity

. and Main Steam Safety The leakage pathways are: Valves (MSSVs)

a. Direct steam relief to the atmosphere through the S/G PORVs.

b. Primary-to-secondary steam generator tube leakage and subsequent steam relief to the atmosphere through the S/G PORVs.

15.3.3.3.2 and MSSVs Identification of Uncertainties and Conse rvative Elements in the Analysis

a. The initial reactor coolant and secondary coolant iodine activities are based 1 gpm on the assumptions stated in Section 15.3.3.3.1.2.

b. A 1-gpm steam generator primary-to-secondary leakage, which is significantly greater than that anticipated during normal operation, is assumed.

c. The coincident loss of offsite power with the occurrence of a reactor coolant pump locked rotor is a highly conservative assumption. In the event of the availability of offsite power, the condenser steam dump valves will open, permitting steam dump to the condenser. Thus there is no direct release to the environment.

d. The meteorological conditions which may be present at the site during the course of the accident are uncertain. However, it is highly unlikely that the meteorological conditions assumed will be present during the course of the accident for any extended period of time. Therefore, the radiological consequences evaluated, based on the meteorological conditions assumed, are conservative.

15.3.3.3.3 Conclusions 15.3.3.3.3.1 Filter Loadings The only ESF filtration system considered in the analysis which limits the consequences of the reactor coolant pump locked rotor accident is the control room filtration system.

Activity loadings on the control room charcoal filter are based on the flow rate through the filter, the concentration of activity at the filter inlet, and the filter efficiency.

The activity in the control room filter as a function of time has been evaluated for the loss-of-coolant accident, Section 15.6.5. Since the control room filters are capable of

ULNRC-06636 Enclosure 6 Page 100 of 374 CALLAWAY - SP

, and in the Control Room accommodating the potential design-basis loss-of-coolant accident fission product iodine loadings, more than adequate design margin is available with respect to postulated reactor coolant pump locked rotor accident releases. ,

15.3.3.3.3.2 Doses to Receptor at the Exclusion Area Boundary and Low-Population TEDE doses Zone Outer Boundary The potential radiological consequences resulting from the occurrence of a postulated reactor coolant pump locked rotor have been conservatively analyzed, using assumptions and models described in previous sections. ,

worst 2 The total-body doses due to immersion from direct radiation and the thyroid dose due to inhalation have been analyzed for the 0-2 hour dose at the exclusion area boundary and for the duration of the accident (0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) at the low-population zone outer boundary.

The results are listed in Table 15.3-4. The resultant doses are well within the guideline values of 10 CFR 100.

Regulatory Guide 1.183 for offsite locations and the 15.3.3.4 Conclusions full GDC 19 limit in the control room

a. Since the peak RCS pressure reached during this transient is less than that which would cause stresses to exceed the faulted condition stress limits,

, and in the control the integrity of the primary coolant system is not endangered.room for 30 days

b. Since the peak clad surface temperature calculated for the hot spot during the transient remains considerably less than 2,700°F, the core will remain in place and intact with no loss of core cooling capability.

15.3.4 REACTOR COOLANT PUMP SHAFT BREAK 15.3.4.1 Identification of Causes and Accident Description The accident is postulated as an instantaneous failure of a reactor coolant pump shaft, as discussed in Section 5.4. Flow through the affected reactor coolant loop is rapidly reduced, though the initial rate of reduction of coolant flow is greater for the reactor coolant pump rotor seizure event. Reactor trip is initiated on a low flow signal in the affected loop.

Following initiation of the reactor trip, heat stored in the fuel rods continues to be transferred to the coolant, causing the coolant to expand. At the same time, heat transfer to the shell side of the steam generators 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 reduced heat transfer in the steam generators, causes an insurge into the pressurizer and a pressure increase throughout the RCS. The insurge into the pressurizer compresses the steam volume, actuates the automatic spray

ULNRC-06636 Page 101 of 374 Insert T15.3-3 CALLAWAY - SP TABLE 15.3-3 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A LOCKED ROTOR ACCIDENT I. Source Data

a. Power level, MWt 3,636 (includes 2% uncertainty)

b. Steam generator tube leakage, gpm 1

c. Reactor coolant initial iodine activity Dose equivalent of 1.0

µCi/gm of I-131 (adjusted consistent with Table 15.6-4 item I.c.1)

d. Reactor coolant initial noble gas activity Based on 1-percent fuel defects, as provided in Table 15A-5

e. Secondary system initial iodine activity Equivalent to 1/10 of the initial RCS activity

f. Activity released to reactor coolant from failed fuel

1. Noble gas, percent of gap inventory 5

2. Iodine, percent of gap inventory 5

3. Gap inventory Table 15A-3

g. Iodine partition factor in the steam 0.01 generators for secondary side releases

h. Iodine partition factor in the steam 0.161 generators for primary-to-secondary leakage

j. i. Reactor coolant mass, lbs 5.5E+5 k.j. Steam generator mass, per generator 9.25E+5 5.51E+5 II. Atmospheric Dispersion Factors See Table 15A-2 9.25E+4

ULNRC-06636 Page 102 of 374 TABLE 15.3-4 RADIOLOGICAL CONSEQUENCES OF A LOCKED ROTOR ACCIDENT 0.48 - 2.48 hrs Doses (rem)

Exclusion Area Boundary (0-2 hr) 4.0E-01 Thyroid 2.0E01 Whole body 3.8E-01 Low Population Zone Outer Boundary (duration) 2.1E-01 Thyroid 8.1E00 Whole body 8.8E-02 Control Room (0-30 days) 1.34

ULNRC-06636 Page 103 of 374 CALLAWAY - SP hotter side of the rod. Calculations have indicated that this bowing would result in a negative reactivity effect at the hot spot since Westinghouse cores are under-moderated, and bowing will tend to increase the under-moderation at the hot spot. Since the 17 x 17 fuel design is also under-moderated, the same effect would be observed. In practice, no significant bowing is anticipated, since the structural rigidity of the core is more than sufficient to withstand the forces produced. Boiling in the hot spot region would produce a net flow of coolant away from that region. However, the heat from the fuel is released to the water relatively slowly, and it is considered inconceivable that cross flow will be sufficient to produce significant lattice forces. Even if massive and rapid boiling, sufficient to distort the lattice, is hypothetically postulated, the large void fraction in the hot spot region would produce a reduction in the total core moderator to fuel ratio, and a large reduction in this ratio at the hot spot. The net effect would therefore be a negative feedback. It can be concluded that no conceivable mechanism exists for a net positive feedback resulting from lattice deformation. In fact, a small negative feedback may result. The effect is conservatively ignored in the analysis.

15.4.8.3 Radiological Consequences 15.4.8.3.1 Method of Analysis 15.4.8.3.1.1 Physical Model Prior to the accident, it is assumed that the plant has been operating with simultaneous fuel defects and steam generator tube leakage for a time sufficient to establish equilibrium levels of activity in the reactor coolant and secondary systems.

The RCCA ejection results in reactivity being inserted to the core which causes the local power to rise. In a conservative analysis, it is assumed that partial cladding failure and fuel melting occurs. The fuel pellet and gap activities are assumed to be immediately and uniformly released within the reactor coolant. Two release paths to the environment exist which are analyzed separately and conservatively, as if all the activity is available for release from each path.

The activity released to the containment from the reactor coolant through the ruptured control rod mechanism pressure housing is assumed to be mixed instantaneously throughout the containment and is available for leakage to the atmosphere. The only removal processes considered in the containment are iodine plateout, radioactive decay, and leakage from the containment.

The model for the activity available for release to the atmosphere from the S/G relief valves assumes that the release consists of the activity in the secondary system plus that fraction of the activity leaking from the reactor coolant through the steam generator tubes. The leakage of reactor coolant to the secondary side of the steam generator continues until the pressures in the reactor coolant and secondary systems equalize.

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

15.4-43 Rev. OL-22

ULNRC-06636 Enclosure 6 104 of 374 Page 104 375 and MSSVs CALLAWAY - SP Thereafter, no mass transfer from the reactor coolant system to the secondary system due to the steam generator tube leakage is assumed. Thus, in the case of coincident loss of offsite power, activity is released to the atmosphere from steam relief through the S/G PORVs.

1.183. Tables 15B-1 and 15B-7 provide a comparison of 15.4.8.3.1.2 the analysis to the guidelines of Regulatory Guide 1.183.

Assumptions and Conditions The major assumptions and parameters used in the analysis are itemized in Tables 15.4-3 and 15A-1 and summarized below. The assumptions are consistent with Regulatory Guide 1.77.

The assumption used to determine the initial concentrations of isotopes in the reactor coolant and secondary coolant prior to the accident are as follows:

a. The initial reactor coolant iodine activity corresponds to an isotope mixture that bounds Technical Specification allowable conditions for both tight and open fuel defects. The initial isotopic mix is based on the relative concentrations from Table 11.1-5. The concentrations are then changed to achieve a Dose Equivalent I-131 (DEI) of 1.0 µCi/gm, while maintaining the isotopic ratios from Table 11.1-5. This provides conservative values for the longer lived iodines which contribute the majority of the calculated thyroid dose. The initial concentration of the shorter lived iodines are then increased to bound the concentrations which would be observed in the presence of open fuel defects. The shorter lived iodine isotopes are not major contributors of thyroid dose, but may provide a noticeable contribution to calculated whole body dose. The initial reactor coolant iodine activity assumed for this sequence, as provided in Table 15A-5, to the TEDE. bounds allowable plant conditions for open or tight fuel defects, and the contributions of the longer and shorter lived isotopes to whole body and thyroid consequences.

b. The noble gas activity in the reactor coolant and secondary system is based on 1-percent fuel defects. 10%

c. The initial secondary side iodine activity to 1/10th of the initial assumed primary side iodine activity.

The noble gas concentrations for both Kr and Xe are then scaled to achieve a Dose Equivalent The following Xe-133 conditions equaltotocalculate are used the Technical Specification the activity releasedlimit andofthe 225 Ci/gm.

offsite doses following a RCCA ejection accident.

a. 10 percent of the fuel rod gap activity, except for Kr-85 which is 30 percent, is additionally released to the reactor coolant.

b. 0.25 percent of the fuel is assumed to melt. Insert 15.4.8.3.1.2A Insert 15.4.8.3.1.2B 15.4-44 Rev. OL-22

ULNRC-06636 Enclosure 6 Page 105 of 374 residual heat removal operation to take-over CALLAWAY SP decay heat removal at 7.29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br />

c. Following the incident until primary and secondary side pressures equalize, steam is released to the environment.

leakage

d. The 1-gpm primary-to-secondary leak to the steam generators is assumed.

e. All noble gas activity in the reactor coolant which is transported to the secondaryfor the primary system via thetoprimary-to-secondary secondary release pathway leakagecase and is assumed to be Insert 15.4.8.3.1.2C immediatelyinstantaneously released to the and homogeneously mixed within environment.

containment for the containment leakage pathway case

f. Fission products released from the fuel-cladding gap of the damaged fuel rods are assumed to be instantaneously and homogeneously mixed with the reactor coolant.

g. A partition factor of 0.1 between the vapor and liquid phases for radioiodine in the steam generators is used for secondary side releases and 0.1 for iodine in primary-to-secondary leakage.

i.h. The activity released from the steam generators is immediately relieved to the environment.

j.i. The containment is assumed to leak at 0.2 volume percent/day during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> immediately following the accident and 0.1 volume percent/

day thereafter. or control room k.j. No credit is taken for radioactive decay or ground deposition during radioactivity transport to offsite location.

s l.k. Short-term accident atmospheric dispersion factors corresponding to ground level releases, breathing rates, and dose conversion factors are given in Tables 15A-2, 15A-1, and 15A-4, respectively.

m.l. Offsite power is assumed lost.

15.4.8.3.1.3 Mathematical Models Used in the Analysis Mathematical models used in the analysis are described in the following sections:

a. The mathematical models used to analyze the activity released during the in thatare course of the accident thedescribed condenserinisAppendix not available 15A.to receive the steam released for decay heat removal in the fuel.

b. The atmosphericHowever, dispersion asfactors bias applied used intothe increase thewere analysis heatcalculated load based on the on (and site so steam release) meteorological all four reactor measurement coolant programs pumpsi n described continue to operate.

Section 2.3 of the Site Addendum, and are provided in Table 15A-2.

ULNRC-06636 Enclosure 6 Page 106 of 374 CALLAWAY - SP TEDE and in the control room

c. The thyroid inhalation and total-body immersion doses to a receptor at the exclusion area boundary or outer boundary of the low-population zone were analyzed, using the models described in Appendix 15A.

15.4.8.3.1.4 Identification of Leakage Pathways and Resultant Leakage Activity and MSSVs The leakage pathways are:

a. Direct steam relief to the atmosphere through the S/G PORVs

b. Primary-to-secondary steam generator tube leakage and subsequent steam relief to the atmosphere through the S/G PORVs. and MSSVs

c. The resultant activity released to the containment is assumed available for leakage directly to the environment.

15.4.8.3.2 Identification of Uncertainties and Conservative Elements in the Analysis

a. Reactor coolant and secondary coolant activities are many times greater than assumed for normal operation conditions.

a.b. A 1-gpm steam generator primary-to-secondary leakage, which is significantly greater than that anticipated during normal operation and greater are

d. The alkali particulates thanconservatively the Technical combined Specification limit with, andoftreated 600 gallons per day,for as, halogens is assumed.

transport trough the steam generators.

b.c. The coincident loss of offsite power with the occurrence of a RCCA ejection accident is a highly conservative assumption. In the event of the availability of offsite power, the condenser steam dump valves will open, permitting steam dump to the condenser. Thus there is no direct release via that path to the environment. none recognized c.

d. It is assumed that 50 percent of the iodines released to the containment atmosphere is adsorbed (i.e. plate out) onto the internal surfaces of the containment or adheres to internal components. However, it is estimated that the removal of airborne iodines by various physical phenomena such as adsorption, adherence, and settling could reduce the resultant doses by a factor of 3 to 10.

e.

d. The activity released to the containment atmosphere is assumed to leak to the environment at the containment leakage rate of 0.2-volume percent/

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 0.1-volume percent/day thereafter. The initial containment leakage rate is based on the peak calculated internal containment pressure anticipated after a LOCA. The pressures associated with a RCCA ejection accident are considerably lower than that calculated for a LOCA. The pressure inside the containment also decreases

ULNRC-06636 Page 107 of 374 a significant factor considerably with time, with an expected decrease in leakage rates.

Taking into account that the containment leak rate is a function of pressure, the resultant doses could be reduced by a factor of 5 to 10 (Ref. 12).

e.f. The meteorological conditions which may be present at the site during the course of the accident are uncertain. However, it is highly unlikely that the meteorological conditions assumed will be present during the course of the accident for any extended period of time. Therefore, the radiological consequences evaluated, based on the meteorological conditions assumed, are conservative.

15.4.8.3.3 Conclusions 15.4.8.3.3.1 Filter Loadings The only ESF filtration system considered in the analysis which limits the consequences of the RCCA ejection accident is the control room filtration system. Activity loadings on the control room charcoal filter are based on the flow rate through the filter, the concentration of activity at the filter inlet, and the filter efficiency.

The activity in the control room filter as a function of time has been evaluated for the loss-of-coolant accident, Section 15.6.5. Since the control room filters are capable of accommodating the potential design-basis loss-of-coolant accident fission product iodine loadings, more than adequate design margin is available with respect to postulated RCCA ejection accident releases. ,

15.4.8.3.3.2 Doses to Receptor at the Exclusion Area Boundary and Low-Population TEDE doses Zone Outer Boundary and in the Control Room The potential radiological consequences resulting from the occurrence of a postulated RCCA ejection accident have been conservatively analyzed, using assumptions and models described in previous sections.

worst 2 The total-body doses due to immersion from direct radiation and the thyroid ose d due to inhalation have been analyzed for the 0-2 hour dose at the exclusion area boundary and for the duration of the accident at the low-population zone outer boundary. The results are listed in Table 15.4-4. The resultant doses are well within the guideline values of 10 CFR 100.

15.4.8.4 Regulatory Guide 1.183 for offsite locations and Conclusions and in the control room for 30 days the full GDC 19 limit in the control room Even on a conservative basis, the analyses indicate that the described fuel and clad limits are not exceeded. It is concluded that there is no danger of sudden fuel dispersal into the coolant. Since the peak pressure does not exceed that which would cause stresses to exceed the faulted condition stress limits, it is concluded that there is no danger of further consequential damage to the RCS. The analyses have demonstrated

ULNRC-06636 Enclosure 6 Page 108 of 374 TABLE 15.4-3 PARAMETERS USED IN EVALUATING THE RCCA EJECTION ACCIDENT (includes 2% uncertainty)

I. Source Data

a. Core power level, MWT 3636

b. Core inventories Table 15A-3 225 Ci/gm of Xe-133

c. Steam generator tube leakage, gpm 1 dose equivalent

d. Reactor coolant initial noble gas Based on 1-percent fuel activity defects, as provided in Table

g. Radial peaking factor 1.65 15A-5

e. Reactor coolant initial iodine activity See Section 15.4.8.3.1.2.a.

f. Secondary system initial iodine activity See Section 15.4.8.3.1.2.c.

g.

h. Extent of core damage 10 percent of fuel rods experience cladding failure; 0.25 percent of fuel experiences melting i.h. Activity released to reactor coolant, percent

1. Cladding failure (a) Noble gas gap activity 100 25 for Case 1 to containment (b) 100 2 to reactor coolant Iodine gap activity 50 for Case

2. Fuel melting Insert T15.4-3 (a) Noble gas gap activity 100 (b) Iodine fuel activity 50 0.01

j. i. Iodine partition factor in the steam 0.1 generators for secondary side releases bulk boiling

j. Iodine partition factor in the steam 0.1 generators for primary-to-secondary leakage 5.51E+5

m. k. Reactor coolant mass, lbs 5.50 + 5 n.l. Total secondary side fluid mass 4.24E + 5 released to the environment, lbs 1.66E6 steam

ULNRC-06636 Page 109 of 374 TABLE 15.4-3 (Sheet 2)

II. Atmospheric Dispersion Factors Table 15A-2 2.7E+6 III. Activity Release Data

a. Containment volume, ft3 2.5 + 6

b. Containment leak rate, volume percent/day

1. 0-24 hours 0.20

2. 1-30 days 0.10

c. Percent of containment leakage that is 100 unfiltered 0

d. Plateout of iodine within containment, 50 percent

e. Offsite power Lost 3648

f. Mass of primary fluid leaked to the 167 secondary lbs

g. Duration of primary-to-secondary 1200 leakage, sec 26,244 (7.29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br />)

ULNRC-06636 Enclosure 6 Page 110 of 374 TEDE TABLE 15.4-4 RADIOLOGICAL CONSEQUENCES OF A ROD-EJECTION ACCIDENT Control Room (30 days) 4.05 Doses (rem)

CASE 1, Containment Leakage Release Exclusion Area Boundary (0-2 hr) 1.43 Thyroid 1.3E01 Whole body 6.6E-02 Low Population Zone Outer Boundary (duration) 3.24 Thyroid 1.3E01 Whole body 2.3E-02 CASE 2, Steam Generator Atmospheric Steam Dump Release Exclusion Area Boundary (0-2 hr) 1.34 Thyroid 4.9E00 Whole body 0.6 - 2.6 hrs 1.6E-01 7.2E-01 Low Population Zone Outer Boundary (duration)

Thyroid 4.9E-01 Whole body 1.6E-02 Control Room (30 days) 4.13 Rev. OL-15

ULNRC-06636 Enclosure 6 Page 111 of 374 of the reactor makeup system, it would not result in engineered safety features system actuation. Frequent operation of the automatic reactor makeup system will provide the operator some indication of the loss of reactor coolant.

15.6.2.1 Radiological Consequences

c. The alkali metal activity in the reactor coolant is based on 1% fuel defects, as provided in Table 11.1-5 15.6.2.1.1 Method of Analysis 15.6.2.1.1.1 Physical Model The volatile fractions of the spilled reactor coolant are assumed to be available for immediate release to the environment. 225 Ci/gm of Xe-133 dose equivalent 15.6.2.1.1.2 Assumptions and Conditions The major assumptions and parameters used in the analysis are provided in Table 15.6-2 and summarized below:

a. The reactor coolant initial iodine activity is based on the dose equivalent of 1.0 µCi/gm of I-131 (adjusted consistent with Table 15.6-4 item I.c.1).

Although no reactor trip or primary side depressurization is expected, an accident-initiated iodine spiking factor of 500 is assumed in Table 15.6-2 to conservatively address scenarios including a reactor trip.

b. The initial noble gas activity in the reactor coolant is based on 1-percent fuel defects.

d.c. A total of 39,958 pounds of reactor coolant is spilled (based on a release for 30 minutes followed by a 10-second valve closure) onto the auxiliary building floor.

e.d. All of the noble gases in the spilled reactor coolant are released to the environment.

f.e. Ten percent of the spill is assumed to flash. All of the iodine activity in the flashed fraction of the spill is assumed to be released.

Twenty g.f. No credit is taken for mixing and holdup of the releases within the auxiliary building, nor are the auxiliary building normal exhaust filters credited with reducing the release. That is, the release is modeled as being direct to the environment.

15.6.2.1.1.3 Mathematical Models Used in the Analysis Mathematical models used in the analysis are described in the following sections:

ULNRC-06636 Page 112 of 374

a. The mathematical models used to analyze the activity released during the TEDE course of the accident are described in Appendix 15A.

, and in the control

b. The atmospheric dispersion factors usedroom in the analysis were calculated based on the onsite meteorological measurement programs described in Section 2.3 of the Site Addendum, and are provided in Table 15A-2.

c. The thyroid inhalation and total body immersion doses to a receptor at the exclusion area boundary or outer boundary of the low-population zone were analyzed, using the models described in Appendix 15A.

15.6.2.1.1.4 Identification of Leakage Pathways and Resultant Leakage Activity The reactor coolant spilled in the auxiliary building will collect in the floor drain sumps.

From there, it will be pumped to the radwaste treatment system. Therefore, the only release paths that present a radiological hazard involve the volatile fraction of spilled coolant.

Normally, gases released in the auxiliary building mix with the building atmosphere and are gradually exhausted through the filtered building ventilation system. The charcoal filters normally remove a very large fraction of the airborne iodine in the building atmosphere. However, the ventilation system is not designed to mitigate the consequences of an accident (e.g., it might not survive an earthquake more severe than the operating-basis earthquake), nor can the possibility of unplanned leakages from the auxiliary building be eliminated; hence, no credit is taken for these effects reducing the released activity.

The evaporated radionuclides are assumed to be available immediately to the outside atmosphere.

15.6.2.1.2 Identification of Uncertainties and Conservatisms in the Analysis The principal uncertainties in the calculation of doses following a letdown line rupture arise from the unknown extent of reactor coolant contamination by radionuclides, the quantity of coolant spilled, the fraction of the spilled activity that escapes the auxiliary building, and the environmental conditions at the time. Each of these uncertainties is treated by taking worst-case or extremely conservative assumptions.

The extent of coolant contamination assumed greatly exceeds the levels expected in practice. The rupture is postulated in a seismic Category I, ASME Section III, Class 2 piping system. It is assumed that the leak goes undetected for 30 minutes. It is expected that considerable holdup and filtration occurs in the auxiliary building, but no credit is assumed.

The purpose of all these conservatisms is to place an upper bound on doses.

ULNRC-06636 Enclosure 6 Page 113 of 374 15.6.2.1.3 Conclusions 15.6.2.1.3.1 Filter Loadings No filter is credited with the collection of radionuclides in this accident analysis. The buildup on these filters (auxiliary building and control building charcoal filters) that may be expected due to the adsorption of some of the iodine is very small compared with the design capacity of these filters.

15.6.2.1.3.2 Dose to Receptor at the Exclusion Area Boundary and Low-Population Zone Outer Boundary and in the Control Room TEDE The radiological consequences resulting from the occurrence of a postulated letdown line rupture have been conservatively analyzed, using assumptions and models described in previous sections. worst 2 The thyroid inhalation total-body immersion doses have been analyzed for the 0-2 hour dose at the exclusion area boundary and for the duration of the accident at the low-population zone outer boundary. The results are listed in Table 15.6-3. The resultant doses are well within the guideline values of 10 CFR 100.

15.6.3 STEAM GENERATOR TUBE FAILURE The accident examined is the complete severance of a single steam generator tube. This event is considered an ANS Condition IV event, a limiting fault (see Section 15.0.1). The 50.67 and accident is assumed to take place at power withRegulatory the reactorGuide coolant1.183 contaminated with fission products

, and in the control corresponding to co ntinuous operation with a limited amo unt of defective roomfuel for rods.

30 daysThe accident leads to an increase in the contamination of the secondary system due to the leakage of radioactive coolant from the RCS. In the event of a coincident loss of offsite power or failure of the steam dump system, discharge of activity to the atmosphere takes place via the steam generator safety and/or power-operated atmospheric steam dump valves.

In view of the fact that the steam generator tube material is Inconel-600 and is a highly ductile material, it is considered that the assumption of a complete severance is somewhat conservative. The more probable mode of tube failure would be one or more minor leaks of undetermined origin. Activity in the steam and power conversion system is subject to continual surveillance, and an accumulation of minor leaks which exceed the limits established in the Technical Specifications is not permitted during plant operation.

In order to select the reference worst case, a spectrum of steam generator tube rupture (SGTR) events was anayzed. The letters of Reference 3 provide a detailed description of the selection process. Based on the selection process, two major SGTR accident scenarios are identified as the major concerns for radioactive releases to the environment.

ULNRC-06636 Page 114 of 374 15.6.3.1.2 Analysis of Effects and Consequences Method of Analysis -3D Mass and energy balance calculations are performed using RETRAN (Section 15.0.11.8) to determine primary-to-secondary mass release and to determine the amount of steam vented from each of the steam generators from the occurrence of the tube rupture until after the second primary-secondary pressure equalization. RETRAN provides time-dependent values of RCS mass, break flow, flashed fraction, steam generator liquid mass, and steam generator atmospheric steam dump valve flow for the calculation of radiological consequences. Conservatively high values of break flow rate and flashed fraction are assumed for the first hour of the transient to maximize radiological consequences. Supplementary mass and energy balance calculations, with conservative assumptions, are performed for the period from pressure equalization until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the accident, beyond the time of RHR initiation.

In estimating the mass transfer from the RCS through the broken tube, the following assumptions are made:

a. Reactor trip and safety injection occur coincidentally as a result of low pressurizer pressure. Overtemperature T trip is not considered. This allows more break flow. Loss of offsite power occurs at reactor trip.

b. The tube rupture is a double-ended guillotine break of a single hot leg tube at the tube sheet of the steam generator. This break location maximizes the flashed fraction of the RCS break flow.

c. As listed on Table 15.0-4, the low pressurizer pressure safety analysis limit (SAL) for reactor trip is 1845 psig. This reactor trip SAL is lower than the actual setpoint of 1885 psig, which thereby delays the trip and results in increased break flow. Safety injection is assumed concurrent with reactor trip which decreases the time for initiation of safety injection, again resulting in increased break flow. Safety injection occurs 15 seconds after the SI signal. The actual SI setpoint is 1849 psig with a lower SAL in Table 15.0-4. This minimum expected delay results in an early rise in RCS pressure due to SI and results in increased break flow.

d. Break flow is characterized by resistance-limited flow. An additional 5%

uncertainty is added to the flow.

e. The assumption of a loss of offsite power at reactor trip prevents steam dump to the condenser and steam is discharged to the atmosphere via the ASDs. With the condenser unavailable for retention of any leaked radioactivity, offsite doses are maximized.

f. Pressurizer heaters and spray are not modelled.

ULNRC-06636 Page 115 of 374

g. MSIV isolation is modeled at reactor trip and the assumed loss of offsite power, although it could be sig nificantly delayed based on the expected This encompasses operator response. Early isolat ion of the a 2.3 second MSIV causes the failed open ASD to have a greater impact on the signal ruptureddelay steam and gen aerator pressure, which maximizes steam flow and bre ak2.0 flow flashing.

second also valveSteamline isolationFeedwater maximizes the mass transferredstroke from thetime.ruptured steam generator to the atmosphere. 4.3

h. Prior to reactor trip, the normal feedwater matches the steam flow in the intact steam generators. For the ruptured steam generator, the total feed flow (including the break flow) matches the steam flow. The feedwater isolation signal occurs 2.3 seconds after reactor trip causes rod motion and the feedwater isolation valves stroke closed within 2.0 seconds. These are the minimum expected delay and stroke time, respectively, which tend to decrease heat removal from the RCS resulting in higher RCS temperatures and pressures. This results in maximum flashed fraction and break flow.

62%

i. The initial steam generator liquid level is 43.4% of the narrow range span.

This is the minimum expected level, minimizing the amount of secondary inventory available for decay heat removal. This increases the flashed fraction (the amount of leaked reactor coolant that is vaporized on the secondary side). Auxiliary feedwater (AFW) flow is maintained to achieve a narrow range level of at least 45% in all steam generators. AFW is initiated 60 seconds after reactor trip and attains a flow rate of 250 gpm to all steam generators.50% Thismaximum expected delay for AFW initiation maximizes break flow and maintains high RCS temperatures. This minimum expected AFW flow to the ruptured steam generator results in decreased RCS heat rem oval, until maintains the cooldown hightoRCS temperatures, and thereby maximizes the fl ashedRHRfraction conditionsof leaked is reactor coolant.

initiated when it is

j. The ruptured steam gene rators's loweredASD to 1% to at 1184.7 psia. This is 4%

is set tpoint which higher than the nominal semaximize the steam delays the release of pressure from the ruptured steam gener ator, resulting released duringin increas the ed valve discharge flow The ASD on the ruptu and integrated break flow. long term cooldown.red steam generator fails open for 20 minutes, begi nning on initial demand, shortly after reactor trip.

k. The initial reactor applying coolant a 1.2133 average multiplier to temperature is 4.3°F above the nominal value. This increases the flashed fracti on of the RCS break flow.

l. Core residual heat generation is based on the 1979 version of ANS 5.1.

ANSI/ANS-5.1-1979 is a conservative representation of the decay energy release rates. Long-term operation at the initial power level preceding the trip is assumed.

m. The narrow range level in all steam generators must be greater than 4%

and the ruptured steam generator pressure must be greater than 430 psig 15.6-11 Rev. OL-25

ULNRC-06636 Enclosure 6 Page 116 of 374 prior to initiating RCS cooldown. The cooldown is initiated 10 minutes after the failed ASD is isolated.

n. RCS depressurization is assumed to begin 3 minutes after completion of cooldown. When the ruptured steam generator pressure is higher than the RCS pressure, the pressurizer PORVs are closed.

o. Safety injection is terminated 5 minutes after completion of RCS depressurization.

Other initial conditions, given in Table 15.0-2, are chosen to maximize RCS temperatures, decay heat, flashed fraction of RCS leakage, and break flow, thereby maximizing radioactivity transfer to the secondary and, consequently, offsite doses.

The ASDs on the The above assumptions, suitably conservative for this case are made to maximize offsite intact SGs are doses.

conservatively assumed toPrior opento reactor trip, steam is dumped to the condenser from both the ruptured and intact only 90%. steam generators. After the condenser is lost, following assumed loss of offsite power at reactor trip, steam from all steam generators is released to the atmosphere.

that of Following isolation of the ruptured steam generator, one of the ASDs on the three intact steam generators isofconservatively the intact loops umed to fail closed. This additional failure is ass beyond single failure criteria. The effect of this assumption is to conservatively increase the time it takes to reduce the RCS temperature to below the ruptured steam generator saturation temperature using atmospheric steam dump from the intact steam generators.

From 2 to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, steam is assumed to be relieved from the intact steam generators to reduce the RCS temperature and pressure to RHRS conditions. The ruptured steam generator is depressurized to the RHRS cut-in pressure using the emergency recovery procedures. After 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, further plant cooldown is carried out with the RHRS. The 0 to 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 2 to 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> steam releases from the intact steam generators required to remove decay heat, metal heat, reactor coolant pump heat, and stored fluid energy in the RCS and steam generators are determined based on these assumptions.

5.86 5.86 Key Recovery Sequence The ruptured loop has no cooling to drive The recovery sequence to be followed consists circulation in the loop of the following and operator major thereforeactions:

remains at a higher temperature than the intact loops. After the

a. intactgenerator; Identification of the ruptured s team loops have reached RHR cut-in temperature the ruptured loop ASD block valve is reopened to

b. Isolation of the ruptured steamdepressurize the ruptured generator including closureSGoftothe RHR cut-in manual ASD block valve; pressures and to induce circulation and cool the ruptured loop to RHR-cut in conditions. This

c. Assuring After subcoolinghave the operators of the method putRCS was chosen fluid below for cooling saturation the ruptured SG as at the ruptured steam generator pressure; an end to the rupture flow by it is the most conservative with respect to steam lowering the primary release pressure to below that of the ruptured SG, 15.6-12 Rev. OL-25

ULNRC-06636 Page 117 of 374

d. Controlled depressurization of the RCS to a value equal to the ruptured steam generator pressure; intact loops to RHR cut-in temperature

e. Subsequent termination of safety injection flow; and

f. Further cooldown and depressurization of the RCS to conditions suitable for RHR initiation.

g. Depressurization of the ruptured SG and the RCS and cooldown of Results the ruptured loop until RHR cut-in conditions are reached.

In Table 15.6-1, the sequence of events is presented. These events include postulated operator action times. Loss of offsite power is assumed to occur at reactor trip. 220,243 The previously discussed assumptions lead to an estimate of 486,000 pounds for the total amount of reactor coolant transferred to the secondary side of the ruptured steam generator as a result of a tube rupture accident. The steam releases to the condenser and atmosphere from both the ruptured and intact steam generators are given in Table 15.6-4.

The following is a list of figures of pertinent This value time dependent is conservatively parameters:

biased to 300,000 pounds for use as the basis for offsite and control room dose.

Number Title 15.6-3a Pressurizer and Steam Generator (Ruptured and Intact Generators)

Pressure Transients for Steam Generator Tube Rupture Event 15.6-3b Reactor Coolant System Temperature (Ruptured Loop) Transient for Steam Generator Tube Rupture Event 15.6-3c Reactor Coolant System Temperature (Intact Loops) Transient for Steam Generator Tube Rupture Event 15.6-3d Steam Flow Rate (Intact Generators) Transient for Steam Generator Tube Rupture Event 15.6-3e Steam Flow Rate (Ruptured Generator) Transient for Steam Generator Tube Rupture Event 15.6-3f Steam Generator Temperature (Ruptured and Intact Generators)

Transients for Steam Generator Tube Rupture Event 15.6-3g Steam Generator Atmospheric Relief Valve Flow Rate (Ruptured Generator) Transient for Steam Generator Tube Rupture Event 15.6-3h Steam Generator Atmospheric Relief Valve Flow Rate (Intact Generators)

Transient for Steam Generator Tube Rupture Event

ULNRC-06636 Page 118 of 374 Number Title 15.6-3i Faulted Steam Generator Break Flow Rate Transient for Steam Generator Tube Rupture Event 15.6-3j Auxiliary Feedwater Flow Rate and Narrow Range Level (Ruptured Generator) Transients for Steam Generator Tube Rupture Event 15.6-3k Auxiliary Feedwater Flow Rate and Narrow Range Level (Intact Generators) Transients for Steam Generator Tube Rupture Event 15.6-3l Steam Generator Liquid Volume (Ruptured Generator) Transient for Steam Generator Tube Rupture Event 15.6-3m Pressurizer PORV Flow Rate Transient for Steam Generator Tube Rupture Event 15.6-3n Pressurizer Liquid Volume Transient for Steam Generator Tube Rupture Event 15.6-3o Feedwater Flow Rate (Ruptured Generator) Transient for Steam Tables 15B-1Generator Tubeprovide and 15B-5 Rupturea Event comparison of the analysis to the guidelines in Regulatory Guide 1.183.

15.6-3p Feedwater Flow Rate (Intact Generators) Transient for Steam Generator Tube Rupture Event 15.6.3.1.3 Radiological Consequences Method of Analysis The evaluation of the radiological consequences due to a postulated steam generator tube rupture (SGTR) with a stuck open atmospheric steam dump valve on the ruptured steam generator assumes a complete severance of a single steam generator tube while the reactor is operating at full rated power and a coincident loss of offsite power.

Occurrence of the accident leads to an increase in contamination of the secondary system due Theto intact reactorloop ASDsleakage coolant are ough thrthe tube break. A reactor trip occurs automatically, as a result assumed of only to open 90%. er pressure. The reactor trip will automatically low pressuriz trip the turbine.

Steam generator blowdown will automatically be terminated by the SGBSIS (AFAS) signal (refer to Section 10.4.8) which is initiated by the safety injection signal. The assumed coincident loss of offsite power will cause closure of the condenser steam dump valves to protect the condenser. The steam generator pressure will then increase rapidly, resulting in steam discharge as well as activity release through the steam generator atmospheric steam dump valves. An atmospheric steam dump valve on one of the unaffected steam generators is conservatively assumed not to open and will therefore be unavailable to support the RCS cooldown. This assumption has the effect of increasing the time it takes to reduce the RCS temperature to below the ruptured

ULNRC-06636 Enclosure 6 Page 119 of 374 steam generator saturation temperature. This additional failure is beyond the required single failure criteria. Venting from the affected steam generator, i.e., the steam generator which experiences the tube rupture, will continue until the manual block valve is closed, isolating the stuck open atmospheric steam dump valve on the ruptured steam generator. At this time, the affected steam generator is effectively isolated. The remaining unaffected steam generators remove core decay heat by venting steam through the atmospheric steam dump valves until the controlled cooldown is terminated.

alkali metals, The analysis of the radiological consequences of an SGTR considers the most severe release of secondary activity, as well as reactor activity leaked from the tube break. The inventory of iodine and noble gas fission product activity available for release to the environment depends on the primary-to-secondary break flow and coolant leakage rates, the percentage of defective fuel in the core, flashed fraction of reactor coolant, and the mass of steam discharged to the environment. Conservative assumptions were made for all these parameters.

activity initially in the primary and secondary Assumptions systems and Conditions The major assumptions and parameters assumed in the analysis are itemized in Tables 15.6-4 and 15A-1 and are summarized below.

The assumptions RHRused conditions are reached to determine in the intact loops.

the concentrations At thisinpoint of isotopes the ruptured the reactor coolant loop ASD block valve is reopened to and secondary systems prior to the accident are as follows:induce natural circulation and cooling in the ruptured loop. Steam release through the ASDs of all loops is

a. terminated The assumed when RHRcoolant reactor cut-in temperature is is iodine activity reached in thefor determined ruptured loop.

the following two cases:

Case 1 - The initial reactor coolant iodine activity corresponds to an isotope mixture that bounds Technical Specification allowable conditions for both tight and open fuel defects.

The isotopic mix is based on the initial RCS concentrations from Table 15A-5. This table provides conservative values for the iodine isotopic spectrum that bound the RCS concentrations which could be expected with either tight or open fuel defects. Case 1 then includes an accident initiated, spiked release rate that increases by a factor of 335 during the accident sequence.

Case 2 - The initial reator coolant iodine activity corresponds to an assumed pre-accident iodine spike which results in concentrations that are a factor of 60 higher than those used in Case 1.

b. The noble gas activity in the reactor coolant, as provided in Table 15A-5 (225 mCi/gm DOSE EQUIVALENT XE-133) .

ULNRC-06636 Enclosure 6 Page 120 of 374

c. The initial secondary side radio-iodine concentrations are assumed to be 10% of the initial Case 1 primary side concentrations.

The following assumptions and parameters are used to calculate the activity released and the offsite doses following an SGTR:

a. Break flow to the ruptured steam generator is conservatively assigned values that bound calculated break flow rate values. The assumed values bound the break flow rates calculated by the RETRAN code. Break flow rate values are discussed in Table 15.6-4 (225 µCi/gm DOSE EQUIVALENT XE-133).

b. The fraction of reactor coolant that flashes to steam after reaching the secondary side, as assumed in the accident analysis, varies over time.

Key events which trigger changes in the assumed flashed fraction are reactor trip and closure of the manual block valve to isolate the failed open SG atmospheric steam dump valve. Flashed fraction values assumed in the radiological analysis are described in Table 15.6-4.

c. A 1-gpm primary-to-secondary leak is assumed to occur to the unaffected steam genera tors.

occurs for the first 1.39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br />. Primary approximately

d. All noble gas activity in the reactor coolant which is transported to the secondary system via the tube rupture and the primary-to-secondary leakage is assumed to be immediately released to the environment.

e. At 80 minutes after the accident, it is assumed that the RCS and steam generator pressures are equalized and below the steam generator atmospheric relief valve set pressure. Break flow to the ruptured steam generator and primary-to-secondary leakage to the intact steam generators are conservatively assumed to continue until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the tube rupture.

f. The iodine partition faction between the liquid and steam in the steam is generator is assumed to be 0.01. 5.86

g. The steam releases from the steam generators to the atmosphere are given in Table 15.6-4.

At 5.86h. Offsite power is lost.

i. Five hours after the accident, the RHR system is assumed to be in operation to cool down the plant. Thus, no additional steam release is assumed.

j. Radioactive decay prior to the release of activity is considered. No decay during transit or ground deposition is considered.

ULNRC-06636 Page 121 of 374

k. Short-term accident atmospheric dispersion factor, breathing rates, and dose conversion factors are provided in Tables 15A-2, 15A-1, and 15A-4, respectively.

Mathematical Models Used in the Analysis Mathematical models used in the analysis are described in the following sections:

a. The mathematical models used to analyze the activity released during the course of the accident are based on the assumptions listed above.

and in the control room

b. The atmospheric dispersion factors used in the analysis were calculated based on the onsite meteorological measurements program, as described TEDE in Section 2.3 of the Site Addendum, and are provided in Table 15A-2.

c. The thyroid inhalation immersion doses to a receptor at the exclusion area boundary and outer boundary of the low-population zone were analyzed, using the models described in Appendix 15A.

Identification of Leakage Pathways and Resultant Leakage Activity The For the purposes of evaluating the radiological consequences due to a postulated SGTR, the activity released from the affected steam generator, prior to isolation, is released directly to the environment by the atmospheric steam dump valve. Two of the unaffected steam generators are assumed to continually discharge steam and entrained activity via the atmospheric steam dump valves up to the time initiation of the RHR system can be accomplished. Since the activity is released directly to the environment with no credit for plateout or retention, the results of the analysis are based on the most direct leakage pathway available. Therefore, the resultant radiological consequences represent the most conservative estimate of the potential integrated dose due to the postulated SGTR.

Identification of Uncertainties and Conservatisms in the Analysis

a. Reactor coolant activities based on extreme iodine spiking effects are orders of magnitude greater than that assumed for normal operating conditions.

5.86

b. A 1-gpm steam generator primary-to-secondary leakage, with a conservatively high density, is assumed which is significantly greater than that anticipated during normal operation. This leakage continues for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, even though RHR operation is assumed to begin at 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

until

c. Tube rupture of the steam generator is assumed to be a double-ended severance of a single steam generator tube. This is a conservative assumption, since the steam generator tubes are constructed of highly ductile materials. The more probable mode of tube failure is one or more

ULNRC-06636 Enclosure 6 Page 122 of 374 the leaked primary fluids which immediately flashes to steam after arriving in the secondary side. Release via the steaming pathway is terminated by the SG atmospheric steam dump block valve closure at 30 minutes.

Release via the flash pathway is conservatively continued following block valve closure. Release via this pathway is continued until the RETRAN TEDE results indicate that no further flashing will occur.

k. The steam release from the intact steam generator ASDs during the 5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> cooldown to RHR cut-in conditions is conservatively assumed to occur in its entirety during the 0-2 hour period of the transient.

k.l. Whole body doses from the intact steam generator ASDs during the cooldown to RHR cut-in conditions are calculated using conservative primary side activities.

Table 15.6-5 lists the offsite doses for the SGTR with a stuck-open ASD.

15.6.3.2 STEAM GENERATOR TUBE RUPTURE WITH FAILURE OF FAULTED STEAM GENERATOR AFW CONTROL VALVE 15.6.3.2.1 Identification of Causes and Accident Description As discussed in Reference 3, an SGTR case demonstrating the effects of steam generator overfill was performed. In this case the analysis assumes the failure of the AFW control valve on the discharge side of the motor-driven AFW pump feeding the ruptured steam generator. The ASD on the ruptured steam generator is not assumed to fail open. The ASD never opens and all liquid relief is considered through a main steam safety valve (MSSV). The AFW control valve is assumed failed in the wide-open position to maximize the flow to the ruptured steam generator. Failure of this valve coupled with the contribution from the turbine-driven AFW pump provides a greater potential for overfilling the ruptured steam generator. For this special overfill scenario, reactor trip and safety injection actuation were conservatively assumed at SGTR initiation (time zero) to maximize the AFW addition to the ruptured steam generator. Some of the assumptions which differ from the analysis described in Section 15.6.3.1.1 do so because the trip time sensitivity has been eliminated. The effect of these revised assumptions is an increase in break flow and ruptured steam generator AFW flow, which results in overfill and water relief.

The analysis scenario is outlined below. This analysis is consistent with the overfill scenario presented in Reference 3, but has been updated to match the current plant configuration. This includes revised (longer) operator action times that reflect recent simulator studies of this SGTR scenario.

An SGTR occurs while the plant is at 100% thermal power and while at steady state.

Concurrent with the SGTR a reactor trip occurs and a safety injection signal is generated. A loss of offsite power (LOOP) is assumed coincident with the reactor trip.

ULNRC-06636 Page 123 of 374 Following reactor trip, safety injection actuation, and the loss of offsite power, the feedwater flow stops and the Main Steam Isolation Valves close. The secondary pressure rises and approaches the setpoints of the secondary ASDs and MSSVs. In response to the reactor trip and LOOP, auxiliary feedwater is delivered to the secondary.

It is assumed that the AFW control valve fails full open on the ruptured SG and delivers excessive AFW to the ruptured steam generator. The excessive AFW flow quickly rebounds the ruptured steam generator water level and dr ives Thethe steam ASDs aregenerator conservatively toward overfill. assumed to open 90%.

In accordance with the emergency operating procedures (EOPs), the ruptured SG is isolated by ensuring that the MSIV, ASD, and blowdown isolation valves are closed on the ruptured loop. The final isolation step requires AFW termination to the SG. After isolation, the primary and ruptured secondary pressure rise in response to reduced heat removal. Following isolation of the ruptured steam generator, operators begin cooldown of the primary via the intact steam generators ASDs. An atmospheric steam dump valve on one of the unaffected steam generators is conservatively assumed not to open and will therefore be unavailable to support the RCS cooldown. This assumption has the effect of increasing the time it takes to reduce the RCS temperature to below the ruptured steam generator saturation temperture. This additional failure is beyond the required single failure criteria. Eventually proper subcooling limits are obtained and primary depressurization is initiated using a primary power operated relief valve (PORV).

Primary depressurization is performed until primary and secondary pressures equalize.

This stops break flow momentarily. In accordance with EOP procedures, the safety injection flow is terminated fairly soon after the depressurization step. Unfortunately, safety injection flow, in the interim, has re-pressurized the primary and a primary/

secondary pressure difference still exists. After SI termination, it is assumed that the operators minimize the primary/secondary pressure difference by opening a PORV. Any primary rise after this step is moderate and a function of decay heat.

Primary and secondary equilibrium does not occur before the ruptured steam generator overfills and water fills the steamline up to the MSIV. When the steam generator and steamline go water solid a pressure spike (on the secondary) occurs as the primary side (driven by SI) drives the secondary pressure toward equilibrium. Thus a safety valve opens and contaminated water is dumped to the atmosphere. Water continues to be relieved from the ruptured SG MSSV until equilibrium is reached between the primary and secondary pressures, effectively terminating flow into the ruptured steam generator.

To assure continued relief, an active failure of the SV is assumed to occur, i.e., after water relief the valve remains partially open (5%). Eventually, water relief depletes the secondary mass and creates a steam void. This steam void grows until water is no longer able to pass out the safety valve.

It is assumed that steam relief continues until RHR cut-in, since steam relief continues to shrink the ruptured SG mass via cooling and mass depletion. Following break flow termination it is assumed that the operators transition to the cooldown procedures and initiate cooldown via intact SG atmospheric steam dump. Cooldown to RHR cut-in

ULNRC-06636 Page 124 of 374 conditions requires approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from initiation of intact SG atmospheric steam dump.

15.6.3.2.2 6.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Analysis of Effects from the start of the and Consequences event. -3D Method of Analysis Mass and energy balance calculations are performed using RETRAN Section 15.0.11.8 to determine the plant response to the SGTR and calculate the break flow, break flow flashing, secondary releases, and system masses for the calculation of the radiological consequences.

In the calculation of the plant response for this scenario the following assumptions are made:

a. Single failure: The ruptured steam generators auxiliary feedwater control valve fails in the full open position.

b. Additional active failure: The ruptured steam generators safety valve fails partially open (5% effective area) after water relief.

c. The atmospheric steam dump (ASD) valve on the ruptured SG is assumed inoperable in the closed position for the duration of the accident sequence.

d. The tube rupture is modeled as a double-ended-guillotine break of a single tube at the cold leg tube sheet. An additional 5% uncertainty is added to the flow predicted for resistance limited flow.

e. Initial conditions

• Core power = 3565 MWt

• Pressurizer pressure = 2280 psia. This is the nominal pressure plus error allowance. The higher pressure maximizes the break flow.

• Pressurizer level = 38% of narrow range span (NRS)

• Vessel average temperature = 570.7oF - 3oF = 567.7oF. This is the minimum expected vessel average temperature. The lower temperature increases the density of the reactor coolant and thus increases the leakage.

• RCS flow = thermal design flow = 374,400 gpm

• Feedwater temperature = 390oF.

387.2

ULNRC-06636 Page 125 of 374

• Steam generator level = 57.5% NRS. This is the nominal level plus uncertainty to maximize the initial inventory.

• 80.9 tube plugging = 5%.

Steam ge nerator This level maximized the

f. Reactor trip occurs at time zero. initial inventory

g. Loss of offsite power (LOOP) occurs at reactor trip (i.e., at time zero)

h. MSIV isolation is modeled at reactor trip and the assumed loss of offsite power, although it could be significantly delayed based on the expected operator response. Early isolation of the MSIV allows the ruptured SG to depressurize due to the addition of the (maximum) AFW flow, while the intact SG pressure stays relatively high. This results in increased break flow to the ruptured SG, which is conservative. It also leads to higher AFW flow to the ruptured SG. If the MSIV would be left open, the ruptured and intact SGs would tend to be at the same pressure, which would be closer to that of the intact SGs (which are lumped together in the RETRAN model).

Also, with the MSIV open, overfilling the ruptured SG would not necessarily lead to water relief, since the water could go to the intact SGs. The secondary pressure would not spike and the safety valve would not lift.

i. The MSIV closes in 1.5 seconds. As noted above, early isolation is considered to be more limiting.

j. The main feedwater isolation valve (MFIV) closure is modeled as a step function after a 17 second delay. The SI signal generated at reactor trip initiates the isolation. A safety e valuation for 3840 was performed seconds determine the to manual until the impact of a potential increase in the stroke time of the feedwater isolatio n cooldown to RHR conditions is valves beyond the value assumed in the analyses (15 seconds) due to the initiated. Then ramped over 1000 isolation of new valve actuators. It was concluded that the results seconds to 1.2133 0.9706 presented in this section for a steam generator tube rupture with a failed or faulted steam generator AFW control valve are not adversely affected by this plant modification. As such, the reported results and conclusions remain valid.

k. Decay heat = 0.8 x 1979 ANS 2 model

l. The following maximum AFW flow rates are modeled prior to partial/full isolation of AFW flow to the ruptured SG:

• The AFW flow to the ruptured SG Decay heat before = 1.2133 isolation of after 4840 the turbine seconds.

driven AFW pump flow to the ruptured steam generator, at the intact SG pressure of 1235.7 psia is used as a base. As the intact SG

ULNRC-06636 Page 126 of 374 pressure drops the flow to the ruptured SG is reduced. This model is reflected in the table below:

Reduction in Ruptured AFW to Intact SG AFW to SG Pressure Ruptured SG Pressure Ruptured SG (psia) (gpm) (psia) (gpm) 414.7 1312.6 1317.0 414.7 72.6 614.7 1214.0 1209.4 614.7 55.4 814.7 1104.0 1099.1 814.7 37.8 1014.7 982.0 976.8 1014.7 20.0 1139.7 895.0 889.9 1139.7 8.6 1235.7 823.0 818.1 1235.7 0.0

• The AFW flow to the intact SGs (total for the 3) before isolation of the turbine driven AFW pump flow to the ruptured steam generator is provided in the table below.

Intact SG Pressure AFW to Intact (psia) SGs (gpm) 214.7 1691.0 1687.0 414.7 1576.0 1569.9 614.7 1448.7 1455.0 814.7 1321.1 1326.0 1014.7 1202.3 1186.0 1139.7 1085.9 1091.0 1235.7 1008.4 1013.0

m. The following maximum AFW flow rates are modeled after partial/full isolation of AFW flow to the ruptured SG:

ULNRC-06636 Page 127 of 374

• The AFW flow to the ruptured SG after isolation of the turbine driven AFW pump flow to the ruptured steam generator is provided in the table below:

Ruptured AFW to SG Pressure Ruptured SG (psia) (gpm) 414.7 767.8 770.0 614.7 709.3 712.0 814.7 651.0 648.0 1014.7 586.0 583.2 1139.7 537.0 534.4 1235.7 498.0 494.9

• The AFW flow to the intact SGs (total for the 3) after isolation of the turbine driven AFW pump flow to ruptured steam generator, and after complete isolation of AFW to the ruptured SG, is provided in the table below:

Intact AFW to Intact SG Pressure SGs (gpm)

(psia) 214.7 1758.0 1760.0 414.7 1649.5 1656.0 614.7 1546.0 1534.0 814.7 1425.0 1417.8 1014.7 1288.1 1295.0 1139.7 1205.0 1197.5 1235.7 1129.0 1122.1

n. AFW flow is initiated 5 seconds after reactor trip, with a 30-second ramp up to full flow. Quicker initiation of AFW flow provides more limiting results for this accident sequence.

1300 1192.7

ULNRC-06636 Page 128 of 374

o. Safety Injection modeling: High and intermediate injection pumps assumed with maximum expected flow. Injection starts 15 seconds after the SI signal (which is generated at the start of the event). Quicker initiation of AFW flow provides more limiting results for this accident sequence.

p. Only two of the intact SG ASDs are credited in the RCS cooldown. This conservatively assumes an additional failure beyond single failure criteria.

q. Operator actions modeled:

• Isolation of turbine-driven AFW flow to the ruptured SG at 10 minutes from the start of the event.

The intact loop

• ASDs are assumed Isolation to only of all AFW flowopen to the90%. This SG ruptured results at 20 minutes from the in conservatively increasing the start of the cooldown time.

event.

• Initiate cooldown by dumping steam from the lumped intact loop SG ASD after 30 minutes from reactor trip (which is at the start of the event).

• The cooldown is terminated when the core outlet temperature reaches the target temperature specified in the EOPs as a function of the ruptured SG pressure.

• Initiate RCS depressurization using the pressurizer power-operated relief valves 3 minutes after the end of the RCS cooldown.

• The depressurization is terminated when the pressurizer pressure and the faulted SG pressure are equal.

• SI flow is terminated 5 minutes after the depressurization is completed.

• Depressurize using pressurizer power-operated relief valve 15 minutes after SI termination to terminate break flow.

212oF includes

• Cooldown to RHR cut-in is initiated after break flow is terminated.

The RETRAN analysis does not include the complete cooldown to demonstrate RHR conditions. The initial part of the cooldown is shown to demonstrate that once the cooldown is initiated the pressure differential (and break flow) is minimal.

r. The break flow flashing fraction is conservatively determined assuming all break flow is at the ruptured loop hot leg temperature.

ULNRC-06636 Page 129 of 374 At 30 minutes the operators initiate RCS cooldown by opening two of the intact SG 42 ASDs. This cooldown continues until the subcooling margin appropriate to allow the primary depressurization is reached. The cooldown is completed approximately 45 minutes into the event.

45 66 46 At approximately 48 minutes operators depressurize the primary using pressurizer power operated relief valves (PORVs) until primary-secondary pressure equilibrium is reached, at approximately 49 minutes. Safety injection flow is terminated 5 minutes later. A secondary RCS depressurization is initiated at approximately 69 minutes from the start of the event, leading to break flow termination. Cooldown to RHR conditions using two of the intact SG ASDs is assumed to be initiated at approximately 69 minutes from the start of the event.

64 Eventually, the steam void resulting from continued water relief from the assumed stuck open MSSV on the ruptured steam generator grows to the extent that the valve no longer passes water. This occurs at approximately 89 minutes from the start of the event.

84 The following is a list of figures of pertinent time dependent parameters:

Number Title 15.6-33a Pressurizer and Steam Generator (Ruptured and Intact Generators)

Pressure Transients for SGTR Event with Overfill 15.6-33b Reactor Coolant System Temperature (Ruptured Loop) Transient for SGTR Event with Overfill Number Title 15.6-33c Reactor Coolant System Temperature (Intact Loops) Transient for SGTR Event with Overfill 15.6-33d Reactor Coolant System and Steam Generator (Ruptured and Intact Generators) Water Mass Transient for SGTR Event with Overfill 15.6-33e Ruptured Steam Generator Break Flow Flashing Fraction Transient for SGTR Event with Overfill 15.6-33f Steam Generator Temperature (Ruptured and Intact Generators)

Transient for SGTR Event with Overfill 15.6-33g Steam Generator Atmospheric Release Flow Rate (Ruptured Generator)

Transient for SGTR Event with Overfill 15.6-33h Steam Generator Atmospheric Release Flow Rate (Intact Generators)

Transient for SGTR Event with Overfill

ULNRC-06636 Page 130 of 374 15.6-33i Ruptured Steam Generator Break Flow Rate Transient for SGTR Event with Overfill 15.6-33j Auxilary Feedwater Flow Rate and Narrow Range Level (Ruptured Generator) Transients for SGTR Event with Overfill 15.6-33k Auxiliary Feedwater Flow Rate and Narrow Range Level (Intact Generators) Transients for SGTR Event with Overfill 15.6-33l Ruptured Steam Generator Liquid Volume Transient for SGTR Event with Overfill Tables 15B-1 and 15B-5 provide a comparison of the analysis to the guidelines in Regulatory Guide 1.183.

15.6-33m Pressurizer PORV Flow Rate Transient for SGTR Event with Overfill 15.6-33n Pressurizer Liquid Volume Transient for SGTR Event with Overfill 15.6.3.2.3 Radiological Consequences The analysis of the radiological consequences of the SGTR with overfill and water release is performed in a manner consistent with that presented in Section 15.6.3.1.3 for the SGTR with the postulated stuck open ARV. The assumptions are outlined below.

Unless otherwise noted, these assumptions are consistent with the Section 15.6.3.1.3 analysis assumptions.

a. Short-term accident atmospheric dispersion factors and breathing rates are provided in Tables 15A-2 and 15A-1, respectively.

b. Dose conversion factors are listed in Table 15A-4.

c. The initial reactor coolant system (RCS) iodine and noble gas concentrations are defined as in the Section 15.6.3.1.3 dose calculations.

d. Spike modeling

• The accident-initiated iodine spike is modeled as in the Section 15.6.3.1.3 dose calculations.

• The pre-accident iodine spike case spike is modeled as in the Section 15.6.3.1.3 dose calculations.

e. Initial secondary activity is 10% of the primary side activity modeled for the accident-initiated iodine spike.

f. Water/Steam Iodine Partitioning: Fluid released from the steam generators as steam retains a portion of the activity present in the fluid. The partition

ULNRC-06636 Enclosure 6 Page 131 of 374 factor is 0.01. All activity contained in break flow that flashes to steam upon entering the SG is released without partitioning

g. Activity released with water from ruptured SG = 50%. Activity contained in water released from the ruptured SG after overfill is not subject to partitioning. However, only 50% of the activity contained in the water is assumed to become airborne. No additional activity release due to evaporation is modeled. These assumptions were made in the analysis approved in Reference 3.

h. Break flow rate for iodine doses:

• The Section 15.6.3.1.3 dose analysis conservatively modeled a constant break flow rate. For the analysis of doses for the overfill case the transient break flow rate from the RETRAN analysis presented in Figure 15.6-33i is used, up until the time when water relief stops. This is consistent with the analysis approved in Reference 3.

• The calculation of iodine doses until RHR conditions are reached conservatively assumes a break flow of 4 lbm/sec until 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after break flow termination. This is consistent with the analysis approved in Reference 3. This portion of the analysis assumes that RHR conditions are achieved within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of break flow termination, even thought the intact SG releases and the noble gas releases assume 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

60 4,000 seconds h.i. The noble gas doses are calculated in Section 15.6.3.1.3 assuming a constant break flow rate of 65 lbm/sec for the first hour of the transient and 10 lbm/sec thereafter for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. For the analysis of the SGTR with overfill the duration of the 65 lbm/sec break flow is extended until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

i. j. Break flow flashing fraction

• The Section 15.6.3.1.3 dose analysis modeled conservative bounding values for the flashing fraction. For the analysis of doses for the overfill case the transient flashing fraction from the RETRAN analysis presented in Figure 15.6-33e is used. This is consistent with the analysis approved in Reference 3. This analysis models the release of all the activity contained in the flashed break flow.

This conservative assumption is consistent with Section 15.6.3.1.3 which modeled the direct release of activity in flashed break flow even after the ruptured SGs failed open atmospheric steam dump (ASD) valve was isolated.

and secondary side water level in Figure 15.6-33j are used. Figure 15.6-33j shows that the tube bundle in the ruptured SG effectively remains covered for the duration of the accident. On this basis, credit is taken for scrubbing of the flashed fraction 15.6-29 Rev. OL-25 6/21

ULNRC-06636 Primary Enclosureto secondary 6 leakage flashing is conservatively modeled as 4%

6,000375 Page 132 of 374 between and 9,000 seconds to account for the intact CALLAWAY - SP steam generators being partially uncovered during this time period. There is no flashing modeled outside of this time period consistent with Reference 3.

• Primary to secondary leakage flashing is not mod eled. This is consistent with the analysis approved in Reference 3. The leak is small and it is assumed that any steam bubbles formed by flashing leakage would collapse before reaching the top of the water level.

j. k. The ruptured SG releases are modeled using the RETRAN analysis results presented in Figure 15.6-33g. In the calculation of doses for the cooldown to RHR conditions it is assumed that a steam flow rate of 8 lbm/sec is maintained due to the failed open safety valve. Thus, 144,000 lbm of steam is released from the ruptured SG in the 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> from break flow termination until RHR c onditions The totalare reached.

steam released from the ruptured SG k.l. is 292,353 lbm over 9.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

The intact SG releases a re modeled using the RETRAN analysis results presented in Figure 15.6-33h. In the calculation of doses for the cooldown to RHR conditions it is assumed that 1.25E6 lbm of steam is released from the intact SGs from the time of break flow termination. This value was conservatively calculated for the case with the failed open ASD presented in Section 15.6.3.1.3 and remains conservative when applied to the analysis of the SGTR with overfill.

l.m. The reactor coolant system, ruptured steam generator and intact steam generators masses are modeled using the RETRAN analysis results from Figure 15.6-33d. This is consistent with the analysis approved in Reference 3. The analysis presented in the Section 15.6.3.1.3 modeled conservative bounding values for the RCS and secondary masses.

Table 15.6-5a lists the offsite doses for the SGTR with overfill and water release.

The total steam released from the intact steam SGs is 977,502 lbm over 15.6.3.3 nclusions C o6.4 hours, with 396,435 LBM released in the first two hours and 581,117 lbm from the two hours until the RHR cut-in conditions are met.

15.6.3.3.1 Filter Loadings The only ESF filtration system considered in the analysis which limits the consequences of the steam generator tube rupture is the control room filtration system. Activity loadings on the control room charcoal filter are based on flow rate through the filter, concentration of activity at the filter inlet, and filter efficiency.

Activity in the control room filter as a function of time has been evaluated for the LOCA, Section 15.6.5. Since the control room filters are capable of accommodating the potential design-basis LOCA fission product iodine loadings, more than adequate design margin is available with respect to postulated SGTR accident releases.

ULNRC-06636 Enclosure 6 Page 133 of 374 CALLAWAY - SP ,

15.6.3.3.2 Doses to Receptor at the Exclusion Area Boundary and Low-Population Zone Outer Boundary and in the Control Room The potential worst 2 radiological consequences resulting from the occurrence of a postulated SGTR have been cons ervatively analyzed, using assumptions and models described in TEDE doses previous sections.

The total-body dose due to immersion and the thyroid dose due to inhalation have been analyzed for the 0-2 hour period at the exclusion area boundary and for a time period effectively greater than the duration of the accident (0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) at the low-population zone outer boundary. Two potentially limiting failure scenarios have been analyzed.

, and for 30 days in Table 15.6-5 presents the offsite dose results for the case of an SGTR with a stuck-open the control room ASD for the ruptured steam generator. Table 15.6-5a presents the offsite dose results for the case of an SGTR with the postulated failure of the ruptured steam generator AFW flow control valve. For both scenarios, the doses considering a pre-accident iodine spike are within the guideline values of 10 CFR 100. For both scenarios, the doses considering an accident-initiated iodine spike are within the 10% of the guideline values of 10 CFR 100.

15.6.3.4 Conclusions 50.67 and Regulatory Guide 1.183 A steam generator tube rupture will cause no subsequent damage to the RCS or the reactor core. An orderly recovery from the accident can be completed, even assuming simultaneous loss of offsite power.

50.67 and Regulatory Guide 1.183 15.6.4 SPECTRUM OF BWR STEAM SYSTEM PIPING FAILURES OUTSIDE OF CONTAINMENT This section is not applicable to the Callaway Plant.

15.6.5 LOSS-OF-COOLANT ACCIDENTS RESULTING FROM A SPECTRUM OF POSTULATED PIPING BREAKS WITHIN THE REACTOR COOLANT PRESSURE BOUNDARY 15.6.5.1 Identification of Causes and Frequency Classification A loss-of-coolant accident (LOCA) is the result of a pipe rupture of the RCS pressure boundary. For the analyses reported here, a major pipe break (large break) is defined as a rupture with a total cross-sectional area equal to or greater than 1.0 ft2. This event is considered an ANS Condition IV event, a limiting fault, in that it is not expected to occur during the lifetime of the plant, but is postulated as a conservative design basis.

For large-break LOCAs, the most limiting single failure is the loss of one train of ECCS injection. The large-break LOCA analyses assume both maximum containment safeguards (to analyze lowest containment pressure conditions) and minimum ECCS safeguards (to analyze the loss of one complete train of emergency core cooling system

ULNRC-06636 Enclosure 6 Page 134 of 374 15.6-32 4 Inch broken loop (BL) & intact loop (IL) Pumped SI Flow Rate Vs.

Time The peak cladding temperature calculated for the limiting small break LOCA is 1043°F.

The maximum local metal-water reaction is 0.02 percent which is below the acceptance criteria limit of 17 percent. The total core metal-water reaction is less than 0.01 percent which is much less than the 1 percent acceptance criteria. These results are below all acceptance criteria limits of 10 CFR 50.46.

15.6.5.4 Radiological Consequences Insert 15.6.5.4.1 15.6.5.4.1 Method of Analysis in phases 15.6.5.4.1.1 Containment Leakage Contribution PHYSICAL MODEL - Following a postulated double-ended rupture of a reactor coolant pipe with subsequent blowdown, the ECCS limits the clad temperature to well below the melting point and ensures that the reactor core remains intact and in a coolable geometry, minimizing the release of fission products to the containment. However, to demonstrate that the operation of the Callaway Plant does not represent any undue radiological hazard to the general public, a hypothetical accident involving a significant release of fission products to the containment is evaluated.

It is assumed that 100 percent of the noble gases and 50 percent of the iodine equilibrium core saturation fission product inventory is immediately released to the containment atmosphere. Of the iodine released to the containment, 50 percent is assumed to plateout onto the internal surfaces of the containment or adhere to internal components. The remaining iodine and the noble gas activity are assumed to be immediately available for leakage from the containment.

released from the core during Once the gaseous fission product activity is released to the containment each of the release atmosphere, it phases is is subject to various mechanisms of removal which operate simultaneously to reduce the amount of activity in the containment. The removal mechanisms include radioactive decay, containment sprays, and containment leakage. For the noble gas fission products, the only removal processes considered in the containment are radioactive decay and containment leakage.

a. Radioactive Decay - Credit for radioactive decay for fission product concentrations located within the containment is assumed throughout the course of the accident. Once the activity is released to the environment, no credit for radioactive decay or deposition is taken.

iodine-removal and retention

b. Containment Sprays - The containment spray system is designed to absorb airborne iodine fission products within the containment atmosphere. To enhance the iodine-retention capability of the containment sprays, trisodium phosphate is added to the spray solution via baskets adjacent to

ULNRC-06636 Enclosure 6 Page 135 of 374 the sumps. The spray effectiveness for the retention of iodine is dependent on maintaining a long-term sump pH greater than 7.0.

c. Containment Leakage - The containment leaks at a rate of 0.2 volume percent/day as incorporated as a Technical Specification requirement at peak calculated internal containment 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 percent of this leak rate for the remaining duration of the accident. The containment leakage is assumed to be directly to the environment.

1.183 ASSUMPTIONS AND CONDITIONS - The major assumptions and parameters assumed in the analysis are itemized in Tables 15A-1 and 15.6-6 and discussed in Section 6.5A.3.

In the evaluation of a LOCA, all the fission product release assumptions of Regulatory Guide 1.4 have been followed. The following specific assumptions were used in the analysis. Table 15.6-7 provides a comparison of the analysis to the requirements of Regulatory Guide 1.4. The nuclide groups and their release fractions are presented in Table 15.6-6

a. The reactor core equilibrium noble gas and iodine inventories are based on long-term operation at a core power level of 3,636 MWt. (includes 2%

uncertainty)

b. One hundred percent of the core equilibrium radioactive noble gas inventory is immediately available for leakage from the containment.

c. Twenty-five percent of the core equilibrium radioactive iodine inventory is immediately available for leakage from the containment. The other 25%

released to the containment atmosphere instantaneously plates out.

c.d. Of the iodine fission product inventory released to the containment, 91 4.85 percent is in the form of elemental iodine, 5 percent is in the form of 95 of iodine and particulate iodine, and 4 percent is in the form of organic iodine.

particulates 0.15 d.e. Credit for iodine removal by the containment spray system is taken, starting at time zero and continuing until a decontamination factor of 28.7 for the elemental species and 50 for the particulate species has been achieved.

e.

f.The following iodine removal constants for the containment spray system are assumed in the analysis: 20.0 2 minutes after Elemental iodine - 10.0 per hr event initiation Organic iodine - 0.0 per hr Particulates Particulate iodine - 0.45 per hr 6.46 per hr prior to a DF of 50 0.646 per hr after a DF of 50 for four hours, which occurs before a decontamination factor of 200 for the elemental species. Credit for the particulate species removal is continued for the duration of spray but is reduced by a factor of 10 after a decontamination factor of 50 is achieved.

ULNRC-06636 Enclosure 6 Page 136 of 374 f.g. The following parameters were used in the two-region spray model:

Fraction of containment sprayed - 0.85 13,500 Fraction of containment unsprayed - 0.15 Mixing rate (cfm) between sprayed and unsprayed regions - 85,000 Table 6.5-2 Section 6.5 contains a detailed analysis of the sprayed and unsprayed volumes and includes an explanation of the mixing rate between the sprayed and unsprayed regions.

g.h. The containment is assumed to leak at 0.2 volume percent/day during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> immediately following the accident and 0.1 volume percent/

day thereafter.

h.i. The containment leakage is assumed to be direct unfiltered to the environment.

i. j. The control building and control room filters will be 95 percent efficient in the removal of all species of iodine. The emergency exhaust ESF filter efficiency is 90% in the assumptions listed in Table15.6-6.

MATHEMATICAL MODELS USED IN THE ANALYSIS - Mathematical models used in the analysis are described in the following sections:

a. The mathematical models used to analyze the activity released during the course of the accident are described in Section 15A.2.

b. The atmospheric dispersion factors used in the analysis were calculated, based on the onsite meteorological measurements program described in ,

Section 2.3 of the Site Addendum, and are provided in Table15A-2.

TEDE

c. The thyroid inhalation and total-body immersion doses to a receptor exposed at the exclusion area boundary and the outer boundary of the low population zone were analyzed, using the models described in Sections 15A.2.4 and 15A.2.5, respectively. and in the control room

d. Buildup of activity in the control room and the integrated doses to the control room personnel are analyzed, based on models described in Section 15A.3.

15A.2, 15A.3 and 15A.4 IDENTIFICATION OF LEAKAGE PATHWAYS AND RESULTANT LEAKAGE ACTIVITY -

For evaluating the radiological consequences of a postulated LOCA, the resultant activity released to the containment atmosphere is assumed to leak directly to the environment.

No credit is taken for ground deposition or radioactive decay during transit to the exclusion area boundary or LPZ outer boundary.

, or control room 15.6-46 Rev. OL-25

ULNRC-06636 Enclosure 6 Page 137 of 374 The offsite doses from all those pathways at the exclusion area boundary (EAB) and low population zone (LPZ) boundary and the doses to the control room personnel are included within the composite results reported in Table 15.6-8.

15.6.5.4.1.2 Radioactive Releases Due to Leakage from ECCS and Containment Spray Recirculation Lines Subsequent to the injection phase of ESF system operation, the water in the containment recirculation sumps is recirculated by the residual heat removal, ECCS centrifugal charging and safety injection pumps, and the containment spray pumps. Due to the operation of the ECCS and the containment spray system, most of the radioiodine released from the core would be contained in the containment sump. It is conservatively assumed that a leakage rate of 2 gpm from the ECCS and containment spray recirculation lines exists for the duration of the LOCA. This leakage would occur inside the containment as well as inside the auxiliary building. For this analysis, all the leakage is assumed to occur inside the auxiliary building. Only trace quantities of radioiodine are expected to be airborne within the auxiliary building due to the temperature and pH level of the recirculated water. However, 10 percent of the radioiodine in the leaked water is assumed to become airborne and exhausted from the unit vent to the environment through the auxiliary building emergency exhaust filters (90% efficient). No credit is taken for holdup (i.e. decay) or mixing in the auxiliary building; however, mixing and holdup in the sumps are factored into the release and decay removal constants for this pathway.

Radiological Consequences of ECCS/CS Recirculation Line Leakage - The assumptions used to calculate the amount of radioiodine released to the environment are given in Table 15.6-6. The dose models are presented in Section 15.A. The offsite doses from all dose pathways at the exclusion area boundary (EAB) and low population zone (LPZ) boundary and the doses to the control room personnel are included within the composite results reported in Table 15.6-8. 4 gpm (3 gpm below the water line, 1 gpm above the water line) 15.6.5.4.1.3 Releases Due to Leakage of Radioactive Iodine from the RWST An assessment was performed to calculate the thyroid doses at the exclusion area boundary (EAB), low population zone (LPZ) outer boundary, and to the control room personnel associated with an assumed 3 gpm leakage pathway from the containment recirculation sumps through ECCS isolation valves back to the RWST, which is vented to the atmosphere. This calculation was performed to address the scenario presented in Reference 25. for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 8% for the remainder above the water line of the event The calculation assumed that 10% of the radioiodine leaked to the RWST becomes airborne, mixes with the RWST volume, and is released to the environment. Credit is taken for decay in the RWST. The assumptions used to calculate the amount of air radioiodine released to the environment are given in Table 15.6-6. The dose models are presented in Section 15.A. The doses at the EAB, LPZ outer boundary, and to control room personnel are less than the values reported in Table 15.6-8.

The vast majority of the radioiodine in the 3 gpm delivered below the water line is retained in the liquid remaining in the RWST. This retention in the liquid is supported by a calculation performed in accordance with NUREG/CR-5950, accounting for gradual changes in 15.6-47 pH and iodine concentration in the RWST liquid.

Rev. OL-25

ULNRC-06636 Enclosure 6 Page 138 of 374 The offsite doses from all dose pathways at the exclusion area boundary (EAB) and low population zone (LPZ) boundary and the doses to the control room personnel are included within the composite results reported in Table 15.6-8.

15.6.5.4.1.4 Releases Prior to Containment Purge Isolation Operation of the containment mini-purge system is allowed during power operation.

Therefore, during the initial stage of the LOCA sequence, it is possible that the containment mini-purge system would not be isolated. Table 9.4-13 discusses NRC guidance regarding modeling of the potential contribution that this pathway would make to post-LOCA radiological consequences.

An assessment was performed to calculate the doses at the exclusion area boundary (EAB), low population zone (LPZ) outer boundary, and to the control room personnel associated with this pathway. Insert 15.6.5.4.1.4 The calculation assumed that the initial radioiodine concentration in the reactor coolant system fluids is the same as used for the pre-accident spike cases analyzed for the SGTR and MSLB accidents. The blowdown rate of RCS fluids into the reactor building is taken from FSAR Table 6.2.1-32. The assumptions used to calculate the amount of radioactivity released to the environment are given in Table 15.6-6. The offsite doses from all dose pathways at the exclusion area boundary (EAB) and low population zone (LPZ) boundary and the doses to the control room personnel are included within the composite results reported in Table 15.6-8.

15.6.5.4.2 Identification of Uncertainties and Conservatisms in the Analysis The uncertainties and conservatisms in the assumptions used to evaluate the radiological consequences of a LOCA result principally from assumptions made involving the amount of the gaseous fission products available for release to the environment and the meteorology present at the site during the course of the accident. The most significant of these assumptions are:

a. The ECCS is designed to prevent fuel cladding damage that would allow the release of the fission products contained in the fuel to the reactor coolant. Severe degradation of the ECCS (i.e., to the unlikely extent of for the sprayed simultaneous failure of redundant components) would be necessary in region, it order for the release of fission products to occur of the magnitude assumed in the analysis.

b. The release of fission products to the containment is assumed tooccur instantaneously. none

c. It is assumed that 50 percent of the iodines released to the containment b.

atmosphere is adsorbed onto the internal surfaces of the containment or adheres to internal components; however, it is estimated that theremoval recognized

ULNRC-06636 Enclosure 6 Page 139 of 374 of airborne iodines by various physical phenomena such as adsorption, adherence, and settling could reduce the resultant doses by a factor of 3 to 10 (Ref. 20).

c.d. The activity released to the containment atmosphere is assumed to leak to the environment at the containment leakage rate of 0.2-volume percent/

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 0.1-volume percent/day thereafter. The initial containment leakage rate is based on the peak calculated internal significantly containment pressure anticipated after a LOCA. The pressure within the containment actually decreases with time. Taking into account that the containment leak rate is a function of pressure, the resultant doses could Insert 15.6.5.4.2 be reduced by a factor of 5 to 10 (Ref. 20).

e. The meteorological conditions assumed to be present at the site during the course of the accident are based on /Q values, which are expected to be exceeded 5 percent of the time. This condition results in the poorest values of atmospheric dispersion calculated for the exclusion area boundary and the LPZ outer boundary. Furthermore, no credit has been taken for the transit time required for activity to travel from the point of release to the exclusion area boundary and LPZ outer boundary. Hence, the radiological consequences evaluated under these conditions are conservative.

Limited credit has been taken for the transit time required for activity to travel from the point of release to the control room air intake ductwork. Since the safety injection signal will generate a Phase A containment isolation signal, which in turn will generate a control room ventilation isolation signal prior to activity reaching the control room air intake ductwork, there is no requirement to perform response time testing on the control room ventilation isolation functions for LOCAmitigation.

15.6.5.4.3 Conclusions 15.6.5.4.3.1 Filter Loadings No recirculating or single-pass filters are used for fission product cleanup and control within the containment following a postulated LOCA. The only ESF filtration systems expected to be operating under post-LOCA conditions are the control room HVAC system and the auxiliary building emergency exhaust filtration system.

Activity loadings on the control room charcoal adsorbers are based on the flowrate through the adsorber, the concentration of activity at the adsorber inlet, and the adsorber efficiency. Based on the radioactive iodine release assumptions previously described, the assumption that 25 percent of the core inventory of isotopes I-127 and I-129 is available for release from the containment atmosphere and the assumption that the charcoal adsorber is 100 percent efficient, the calculated filter loadings are in accordance with Regulatory Guide 1.52, which limits the maximum loading to 2.5 mg of

ULNRC-06636 Enclosure 6 Page 140 of 374 iodine per gram of activated charcoal. The 100 percent efficiency assumption is conservative for the purpose of checking filter loading and is not to be confused with the 95% efficiency assumption used for radiological consequences as listed in Table 15.A-1.

15.6.5.4.3.2 Doses to a Receptor at the Exclusion Area Boundary and Low Population Zone Outer Boundary The potential radiological consequences resulting from the occurrence of the postulated LOCA have been conservatively analyzed, using assumptions and models described in previous sections. TEDE has The total-body dose due to immersion and the thyroid dose due to inhalation have been analyzed for the 0-2 hour dose at the exclusion area boundary and for the duration of the accident at the LPZ outer boundary. The results, with margin, are listed in Table 15.6-8.

The resultant doses are within the guideline values of 10 CFR 100.

worst 2 15.6.5.4.3.3 Doses to Control Room Personnel 50.67 and Regulatory Guide 1.183 Radiation doses to control room personnel following a postulated LOCA are based on the ventilation, cavity dilution, and dose model discussed in Section 15A.3. TEDE. This dose has Control room personnel are subject to a total-body dose due to immersion and a thyroid dose due to inhalation. These doses have been analyzed, and are provided in Table 15.6-8. The listed doses, with margin, are within the limits established by GDC-19.

is and 10 CFR 50.67 15.6.6 A NUMBER OF BWR TRANSIENTS is This section is not applicable to the Callaway Plant.

15.

6.7 REFERENCES

1. Huegel, D. S., et. al., RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses, WCAP-14882-P-A, April 1999.

2. Friedland, A. J., and Ray, S., "Revised Thermal Design Procedure",

WCAP-11397-P-A, April 1989.

3. SGTR Analysis letters SLNRC 86-01 (1-8-86), SLNRC 86-03 (2-11-86) SLNRC 86-05 (4-1-86), SLNRC 86-08 (9-4-86), ULNRC-1442 (2-3-87), ULNRC-1518 (5-27-87), ULNRC-1849 (10-21-88), ULNRC-2145 (1-29-90), and the NRC SER dated 8-6-90.

4. WCAP-16140, Callaway Replacement Steam Generator Program NSSS Engineering Report, June 2004.

ULNRC-06636 Enclosure 6 Page 141 of 374

30. Huegel, D. S., et. al., RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Reactor Non-LOCA Safety Analyses, WCAP-14882-P-A (Proprietary), April 1999.

31. International Commission on Radiological Protection, Limits for Intakes of Radionuclides by Worker, ICRP Publication 30, 1979.

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

33. K. F. Eckerman and J. C. Ryman, External Exposure to Radionuclides in Air, Water, and Soil, Federal Guidance Report 12, EPA-402-R-93-081, Environmental Protection Agency, 1993.

34. EPA Federal Guidance Report No. 11, EPA-520/1-88-020, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1988.

35. EPA Federal Guidance Report No. 12, EPA-402-R-93-081, "External Exposure to Radionuclides in Air, Water, and Soil," 1993.

36. NUREG/CR-5950, "Iodine Evolution and pH Control", ORNL/TM-12242, Revision 3, December 1992.

ULNRC-06636 Page 142 of 374 TABLE 15.6-1 TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH RESULT IN A DECREASE IN REACTOR COOLANT INVENTORY Accident Event Insert T15.6-1 Time (sec)

Inadvertent opening of a pressurizer safety valve Safety valve opens fully 0.0 Low pressurizer pressure reactor trip setpoint reached 32.8 Rods begin to drop 34.8 Minimum DNBR occurs 35.5 Steam generator tube rupture with stuck-open atmospheric steam dump (ASD) valve Tube rupture occurs 0.0 Reactor trip signal 597 Safety injection signal 597 Rod motion 599 Feedwater terminated 603 Ruptured steam generator atmospheric/

steam dump valve opens 604 Safety injection begins 612 Auxiliary feedwater injection 659 Operator isolates ruptured steam generator by closing manual block valve 1804 Operator initiates RCS cooldown via intact steam generator atmospheric steam dump valves 2404 Operator completes RCS cooldown 3383 Operator initiates RCS depressurization via pressurizer PORVs 3563 Operator completes RCS depressurization 3622 Operator terminates safety injection 3922 Operator equalizes primary-secondary pressure 4816 RHR cut-in conditions reached 18000

ULNRC-06636 Page 143 of 374 TABLE 15.6-1 (Sheet 2)

Accident Event Time (sec)

Steam generator tube rupture with overfill Tube rupture occurs 0.

Reactor trip signal and loss of offsite power 0.

Safety injection signal 0.

Auxiliary feedwater injection starts 5.

Safety injection delivered 15.

Feedwater terminated 17.

Operator terminates auxiliary feedwater from TDAFW pump to ruptured steam generator 600.

Ruptured steam generator water relief begins 957.

970.

Operator terminates auxiliary feedwater from MDAFW pump to ruptured steam generator 1200.

Operator initiates RCS cooldown via intact steam generator atmosperhic steam dump valves 1800.

Operator completes RCS cooldown 2495.0 2700.

Operator initiates RCS depressurization via pressurizer PORVs 2675.

2880.

Operator completes RCS depressurization 2947.

2741.0 Operator terminates safety injection 3247.

3041.

Cooldown to RHR cut-in begins 4161.

3840.

Operator equalizes primary-secondary pressure 3941.

4262.

Ruptured SG safety valve begins to relieve steam 5020.

5326.

RHR cut-in conditions reached 23029.

21800.

Ruptured SG reaches 212oF 33804

ULNRC-06636 Enclosure 6 Page 144 of 374 TABLE 15.6-2 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCE OF THE CVCS LETDOWN LINE RUPTURE OUTSIDE OF CONTAINMENT with an assumed iodine spike that increases the rate of iodineI. release Source Data into the reactor coolant by a factor of 500. This increased

a. rate is assumed Core power for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

level, MWT 3636 (includes 2% uncertainty)

b. Reactor coolant initial iodine activity Dose equivalent of 1.0 µCi/gm of I-131 (adjusted consistent with Table 15.6-4 item I.C.1)

c. Reactor coolant initial noble gas Based on 1-percent fuel activity defects. See Table 15A-5.

d. Iodine spiking factor 500 II. Atmospheric Dispersion Factors See Table 15A-2.

III. Activity Release and alkali metal Dose equivalent of 225 Ci/

a. Break flow rate, gpm 158.9 gm Xe-133

b. Duration, secs 1810

c. Fraction of iodine activity in the spill that is airborne 0.10 0.20

ULNRC-06636 Page 145 of 374 TABLE 15.6-3 RADIOLOGICAL CONSEQUENCES OF A CVCS LETDOWN LINE BREAK OUTSIDE OF CONTAINMENT TEDE Doses (rem)

Exclusion Area Boundary (0-2 hr) 3.88E-01 Thyroid 5.5E+00 Whole body 1.9E-01 Low Population Zone Outer Boundary (duration) 1.33E-01 Thyroid 5.5E-01 Whole body 1.9E-02 Control Room (30 days) 1.93

ULNRC-06636 Page 146 of 374 TABLE 15.6-4 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A STEAM GENERATOR TUBE RUPTURE (SGTR)

I. Source Data

a. Core power level, MWt 3,636 (includes 2% uncertainty)

b. Steam generator tube leakage, gpm 1

c. Reactor coolant iodine activity:

1. Case 1 The initial reactor coolant iodine activity corresponds to an isotope mixture that bounds Technical Specification allowable conditions for both tight and open fuel defects. The isotopic mix is based on the initial RCS concentrations from Table 15A-5. This table provides conservative values for the iodine isotopic spectrum that bound the RCS concentrations which could be expected with either tight or open fuel defects.

Case 1 then includes an accident initiated, spiked release rate that increases by a factor of 335 during the accident sequence.

2. Case 2 The initial reactor coolant iodine activity corresponds to an assumed pre-accident iodine spike which results in concentrations that are a factor of 60 higher than those used in Case 1.

d. Reactor coolant noble gas activity, Based on 1-percent failed fuel as both cases provided in Table 15A-5 (225 µCi/gm DOSE EQUIVALENT XE-133).

e. Secondary system initial activity 10% of Case 1 primary side activity

f. Reactor coolant mass in total primary 5.8E+5 system, lbs

g. Steam generator water mass (each), 9.3E+4 lbs 5.86 hours9.953704e-4 days <br />0.0239 hours <br />1.421958e-4 weeks <br />3.2723e-5 months <br />

h. Offsite power Lost

i. Primary-to-secondary leakage 80 minutes duration II. Atmospheric Dispersion Factors See Table 15A-2

ULNRC-06636 Page 147 of 374 TABLE 15.6-4 (Sheet 2)

III. Activity Release Data Insert T15.6-4sheet2

a. Affected steam generator

1. Reactor coolant discharged to 486,000(1) steam generator, lbs

2. Flashed reactor coolant, percent 16(2)

3. Iodine partition factor for flashed 1.0 fraction of reactor coolant

4. Steam release to atmosphere, lbs 0-2 hrs 123,200 2-8 hrs 0

5. Iodine carryover factor for the 0.01 nonflashed fraction of reactor coolant that mixes with the initial iodine activity in the steam generator

b. Unaffected steam generators

1. Primary-to-secondary leakage, 4,032(3) lbs

2. Flashed reactor coolant, percent Variable

3. Total steam release, lbs 0-2 hours 1.53E+6(4) 2-8 hours 0

4. Iodine carryover factor 0.01(5)

5. RHR Cut-in time, hrs 5 Notes:

ULNRC-06636 Page 148 of 374 CALLAWAY - SP TABLE 15.6-4 (Sheet 3)

1. The noble gas release calculation assumed a conservatively high, constant 65 lbm/

sec break flow rate for the first hour and 10 lbm/sec thereafter through 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, even though RHR operation is assumed to begin at 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The iodine release calculation is based on the conservative break flow rate of 65 lbm/sec until cooldown is completed.

2. The assumed flashed fraction is 16% until closure of the SG atmospheric steam dump block valve. Following closure of the block valve, a variable flashed fraction is assumed which conservatively bounds the values calculated by the RETRAN code.

The intact steam generator flashed fraction is conservatively assumed to be the same as in the ruptured steam generator.

3. Based on 1 gpm leakage and conservative density of 62.4 lbm/cu.ft., giving a mass flow rate of 0.14 lbm/sec for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, even though RHR operation is assumed to begin at 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

4. Assumes that 1.25E06 lbm of steam is relieved for decay heat removal during 5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> cooldown to RHR operating conditions. To maximize dose effects, this release is included in the first two hours following tube rupture.

5. A partition factor of 1.0 is assumed for the flashed fraction.

Insert T15.6-4sheet3

ULNRC-06636 Page 149 of 374 375 Control Room (30 days) CALLAWAY - SP 5.9E-01 TABLE 15.6-5 RADIOLOGICAL CONSEQUENCES OF A STEAM GENERATOR TUBE RUPTURE WITH STUCK-OPEN ATMOSPERHIC STEAM DUMP VALVETEDE Doses (rem)

1. Case 1, accident initiated iodine spike 1.38 Exclusion Area Boundary (0-2 hr)

Thyroid 2.3E01 Whole body 0.1-2.1 8.0E-01 Low Population Zone Outer Boundary (duration) 5.0E-01 Thyroid 2.3E00 Whole body 8.5E-02

2. Case 2, pre-accident iodine spike 2.00 Exclusion Area Boundary (0-2 hr)

Thyroid 5.9E01 Whole body 5.4E-01 Low Population Zone Outer Boundary (duration) 7.2E-01 Thyroid 5.9E00 Whole body 5.8E-02 Control Room (30 days) 2.44

ULNRC-06636 Page 150 of 374 375 Control Room (30 days) CALLAWAY - SP 6.9E-01 TABLE 15.6-5A RADIOLOGICAL CONSEQUENCES OF A STEAM GENERATOR TUBE RUPTURE WITH OVERFILL Doses (rem) 0.2 - 2.2 hrs

1. Case 1, accident initiated iodine spike 2.13 Exclusion Area Boundary (0-2 hr)

Thyroid 2.3E01 Whole body 7.5E-01 Low Population Zone Outer Boundary (duration) 7.3E-01 Thyroid 2.4E00 Whole body 7.5E-02

2. Case 2, pre-accident iodine spike 3.50 Exclusion Area Boundary (0-2 hr)

Thyroid 7.1E01 Whole body 6.3E-01 1.21 Low Population Zone Outer Boundary (duration)

Thyroid 8.0E00 Whole body 6.3E-02 Control Room (30 days) 1.60

ULNRC-06636 Enclosure 6 Page 151 of 374 TABLE 15.6-6 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A LOSS-OF-COOLANT-ACCIDENT Insert T15.6-6A I. Source Data

a. Core power level, MWt 3636 (includes 2% uncertainty)

b. Burnup, full power days 1,000

c. Percent of core activity initially airborne in the containment 547

1. Noble gas 100

2. Iodine 50*

d. Percent of core activity immediately deposited in containment sump

3. Other Particulates 100 (remains in liquid)

1. Noble gas 0

2. Iodine 50 see Item c.2 above Containment

e. Core inventories Table 15A-3 Leakage ECCS and

f. Iodine distribution, percent Release RWST Insert T15.6-6B 1. Elemental 4.85 91 97 Release

2. Organic 0.15 4 3

3. Particulate 95 5 0 II. Atmospheric Dispersion Factors See Table 15A-2 III. Activity Release Data

a. Containment leak rate, volume percent/day

1. 0-24 hours 0.20

2. 1-30 days 0.10

b. Percent of containment leakage that is unfiltered 100

c. Credit for containment sprays

1. Spray iodine removal constants (per 20.0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br />)

a. Elemental iodine 10.0

b. Organic iodine 0.0

c. Particulate 0.45 6.46 prior to DF limit s 0.646 thereafterRev. OL-13

ULNRC-06636 Enclosure 6 Page 152 of 374 Key TABLE 15.6-6 (Sheet 2)

2. Maximum iodine decontamination 200 factors for the containment atmosphere

a. Elemental 28.7

b. Organic 0

c. Particulate 50 13,500

3. Sprayed volume, percent 85

4. Unsprayed volume, percent 15 2.7E+6

5. Sprayed-unsprayed mixing rate, CFM 85,000

6. Containment volume, ft3 2.5E+6 e.d. ECCS recirculation leakage

1. Leak rate (0 - 30 days), gpm 2.0 initial air volume at 2. beginning Sump volume, ft3 57,224 (of 1 gpm above water line) of recirculation for first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

3. Fraction iodine airborne 0.1 0.05 (of 1 gpm above water

4. Emergency exhaust ESF filter 90.0 line) for remainder of event efficiency, %

f.e. RWST leakage 256,626

1. Leak rate (0 - 30 days), gpm 3.0

2. RWST volume, gal. 400,000

3. Fraction iodine airborne 0.1 IV. Control room parameters Tables 15A-1 and 15A-2

• Half instantaneously plates out leaving 25% immediately available for leakage from the containment. 4 gpm total Insert T15.6-6C 3 gpm below water line 1 gpm above water line Note: The release rate from the RWST to the environment is based on the volume displacement from the incoming leakage.

ULNRC-06636 Page 153 of 374 CALLAWAY - SP TABLE 15.6-7 DESIGN COMPARISON TO THE REGULATORY POSITIONS OF Delete REGULATORY GUIDE 1.4 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A LOSS OF COOLANT ACCIDENT FOR PRESSURIZED WATER REACTORS, REVISION 2, JUNE 1974 Regulatory Guide 1.4 Position Design

1. The assumptions related to the release of radioactive material from the fuel and containment are as follows:

a. Twenty-five percent of the equilibrium 1a. Complies.

radioactive iodine inventory developed from maximum full power operation of the core should be assumed to be immediately available for leakage from the primary reactor containment. Ninety-one percent of this 25 percent is to be assumed to be in the form of elemental iodine, 5 percent of this 25 percent in the form of particulate iodine, and 4 percent of this 25 percent in the form of organic iodides.

b. One hundred percent of equilibrium 1b. Complies.

radioactive noble gas inventory developed from maximum full power operation of the core should be assumed to be immediately available for leakage from the reactor containment.

c. The effects of radiological decay 1c. Complies. Credit for during holdup in the containment or other buildings radioactive decay is should be taken into account. taken until the activity is assumed to be released.

d. The reduction in the amount of 1d. Complies. See Table radioactive material available for leakage to the 15.6-6 for reduction environment by containment sprays, recirculating taken.

filter systems, or other engineered safety features may be taken into account, but the amount of reduction in concentration of radioactive materials should be evaluated on an individual case basis.

ULNRC-06636 Page 154 of 374 CALLAWAY - SP TABLE 15.6-7 (Sheet 2)

Regulatory Guide 1.4 Position Design

e. The primary reactor containment 1e. Complies.

should be assumed to leak at the le akDelete rate incorporated or to be incorporated as a technical specification requirement at peak 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 percent of this leak rate for the remaining duration of the accident.

Peak accident pressure is the maximum pressure defined in the technical specifications for containment leak testing.

2. Acceptable assumptions for atmospheric 2a. Complies. Atmospheric diffusion and dose conversion are: dispersion factors were calculated based on the

a. The 0-8 hour ground level release onsite meteorological concentrations may be reduced by a factor ranging measurement programs from one to a maximum of three (see Figure 1) for described in Section 2.3 additional dispersion produced by the turbulent of the Site Addendum.

wake of the reactor building in calculating potential exposures. The volumetric building wake correction, as defined in section 3-3.5.2 of Meteorology and Atomic Energy 1968, should be used only in the 0-8 hour period; it is used with a shape factor of 1/2 and the minimum cross-sectional area of the reactor building only.

b. No correction should be made for 2b. Same as 2a above.

depletion of the effluent plume of radioactive iodine due to deposition on the ground, or for the radiological decay of iodine in transit.

ULNRC-06636 Page 155 of 374 TABLE 15.6-7 (Sheet 3)

Regulatory Guide 1.4 Position Design

c. 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 2c. Complies. See Table of persons offsite should be assumed to bDelete e 3.47 x 15A-1.

-4 10 cubic meters per second. From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the breathing rate should be assumed to be 1.75 x 10-4 cubic meters per second.

After that until the end of the accident, the breathing rate should be assumed to be 1.75 x 10-4 cubic meters per second. After that until the end of the accident, the rate should be assumed to be 2.32 x 10-4 cubic meters per second. (These values were developed from the average daily breathing rate [2 x 107 cm3/day] assumed in the report of ICRP, Committee II-1959.)

d. The iodine dose conversion factors 2d. The dose conversion are given in ICRP Publication 2, Report of factors provided in Committee II, "Permissible Dose for Internal Regulatory Guide 1.109 Radiation," 1959. are used. See Table 15A-4.

e. External whole body doses should be 2e. The dose factors given calculated using "Infinite Cloud" assumptions, i.e., in Regulatory Guide the dimensions of the cloud are assumed to be large 1.109, for noble gases; compared to the distance that the gamma rays and for iodine whole body beta particles travel. "Such a cloud would be dose factors with 5 cm considered an infinite cloud for a receptor at the body tissue attenuation; center because any additional [gamma and] beta and for beta-skin dose emitting material beyond the cloud dimensions factors with credit for would not alter the flux of [gamma rays and] beta attenuation in the dead particles to the receptor" (Meteorology and Atomic skin layer, are used.

Energy, Section 7.4.1.1-editorial additions made so See Table 15A-4.

that gamma and beta emitting material could be considered). Under these conditions the rate of energy absorption per unit volume is equal to the rate of energy released per unit volume. For an infinite uniform cloud containing curies of beta radioactivity per cubic meter the beta dose in air at the cloud center is:

D = 0.457 E

ULNRC-06636 Page 156 of 374 TABLE 15.6-7 (Sheet 4)

Regulatory Guide 1.4 Position Design The surface body dose rate from beta emitters in the infinite cloud can be approximated a sDelete being one-half this amount (i.e., D = 0.23 E ).

For gamma emitting material the dose rate in air at the cloud center is:

D

' = 0.507 E From a semi-infinite cloud, the gamma dose rate in air is:

D

' = 0.25 E Where:

' = beta dose rate from an infinite cloud D

(rad/sec)

D

' = gamma dose rate from an infinite cloud (rad/sec)

E = average beta energy per disintegration (Mev/dis)

E = average gamma energy per disintegration (Mev/dis)

= concentration of beta or gamma emitting isotope in the cloud (curie/

m 3)

f. The following specific assumptions are 2f. See response to 2e.

acceptable with respect to the radioactive cloud dose calculations:

ULNRC-06636 Page 157 of 374 TABLE 15.6-7 (Sheet 5)

Regulatory Guide 1.4 Position Design Delete (1) The dose at any distance from the reactor should be calculated based on the maximum concentration in the plume at that distance taking into account specific meteorological, topographical, and other characteristics which may affect the maximum plume concentration. These site related characteristics must be evaluated on an individual case basis. In the case of beta radiation, the receptor is assumed to be exposed to an infinite cloud at the maximum ground level concentration at that distance from the reactor. In the case of gamma radiation, the receptor is assumed to be exposed to only one-half the cloud owing to the presence of the ground. The maximum cloud concentration always should be assumed to be at ground level.

(2) The appropriate average beta 2f.2 See response to 2e.

and gamma energies emitted per disintegration, as given in the Table of Isotopes, Sixth Edition, by C. M.

Lederer, J. M. Hollander, I. Perlman; University of California, Berkeley; Lawrence Radiation Laboratory; should be used.

g. The atmospheric diffusion model should be as follows:

ULNRC-06636 Page 158 of 374 TABLE 15.6-7 (Sheet 6)

Regulatory Guide 1.4 Position Design (1) The basic equation for 2g.1 Short-term accident atmospheric diffusion from a ground level point atmospheric dispersion source is: Delete factors were calculated based on onsite 1 meteorological

/Q = ------------------

uyz measurement programs described in Section 2.3 Where: of the Site Addendum.

These factors are for

= the short term average centerline ground level releases value of the ground level and are based on concentration (curie/meter3) Regulatory Guide 1.XXX methodology and Q = amount of material released (curie/ represent the worst of sec) the 5-percent site meteorology and the u = windspeed (meter/sec) 0.5-percent worst sector y = the horizontal standard deviation of meteorology.

the plume (meters) [See Figure V-1, Page 48, Nuclear Safety, June 1961, Volume 2, Number 4, "Use of Routine Meteorological Observations for Estimating Atmospheric Dispersion," F.A.

Gifford, Jr.]

z = the vertical standard deviation of the plume (meters) [See Figure V-2, Page 48, Nuclear Safety, June 1961, Volume 2, Number 4, "Use of Routine Meteorological Observations for Estimating Atmospheric Dispersion," F. A.

Gifford, Jr.]

(2) For time periods of greater than 2g.2 See response to 2g.1 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> the plume should be assumed to meander above.

and spread uniformly over a 22.5° sector. The resultant equation is:

2. 32

/Q = --------------

zux

ULNRC-06636 Page 159 of 374 TABLE 15.6-7 (Sheet 7)

Regulatory Guide 1.4 Position Design Delete Where:

x = distance from point of release to the receptor; other variables are given in g(1).

(3) The atmospheric diffusion 2g.3 See response to 2g.1 2

model for ground level releases is based on the above.

information in the following table.

Time Following Accident Atmospheric Conditions 0-8 hours Pasquill Type F, windspeed 1 meter/

sec, uniform direction 8-24 hours Pasquill Type F, windspeed 1 meter/

sec, variable direction within a 22.5° sector 1-4 days (a) 40% Pasquill Type D, windspeed 3 meter/sec (b) 60% Pasquill Type F, windspeed 2 meter/sec (c) Wind direction variable within a 22.5° sector 4-30 days (a) 33.3% Pasquill Type C, windspeed 3 meter/sec (b) 33.3% Pasquill Type D, windspeed 3 meter/sec (c) 33.3% Pasquill Type F, windspeed 2 meter/sec (d) Wind direction 33.3% frequency in a 22.5° sector (4) Figures 2A and 2B give the 2g.4 See response to 2g.1 ground level release atmospheric diffusion factors above.

based on the parameters given in g(3).

ULNRC-06636 Enclosure 6 Page 160 of 374 TABLE 15.6-8 RADIOLOGICAL CONSEQUENCES OF A LOSS-OF-COOLANT-ACCIDENT TEDE TEDE Total 0.5 - 2.5 hr Reported Regulatory Doses (rem) Limits (rem)

I. Exclusion Area Boundary (0-2 hr) 5.63 25 Thyroid 128.4 300 Whole body 4.75 25 II. Low Population Zone Outer Boundary 25 6.24 (0-30 day)

Thyroid 132.8 300 Whole body 1.28 25 III. Control Room (0-30 day) 4.18* 5 Thyroid 25.55 30 Whole body 0.453 5 Beta-skin 7.49 30

• Control Room Dose includes a 0.81 rem TEDE contribution for operator transit dose to/from the Control Room over the duration of the event.

ULNRC-06636 Page 161 of 374

b. All noble gas activity has been removed from the reactor coolant system and transferred to the gas decay tank that is assumed to fail.

c. The maximum content of the waste gas decay tank was conservatively assumed to calculate the isotopic activities for the accumulated radioactivity in the gaseous waste processing system after 40 years' operation and immediately following plant shutdown and degasification of the reactor coolant system.

d. The failure is assumed to occur immediately upon completion of the waste gas transfer, releasing the entire contents of the tank to the radwaste building.

e. The dose is calculated as if the release were from the radwaste building at ground level during the 2-hour period immediately following the accident.

No credit for radioactive decay is taken.

15.7.1.5.1.3 Mathematical Models Used in the Analysis The mathematical models used in the analysis are described in the following sections:

a. The mathematical models used to analyze the activity released during the course of the accident are described in Appendix 15A.

b. The atmospheric dispersion factors used in the analysis were calculated based on the onsite meteorological measurement programs described in Section 2.3 of the Site Addendum.

15A.5.2.3 and 15A.5.2.4,

c. The thyroid inhalation and total-body immersion doses to a receptor at the exclusion area boundary or outer boundary of the low-population zone were analyzed, using the models described in Appendix 15A, Sections 15A.2.4 and 15A.2.5, respectively.

15.7.1.5.1.4 Identification of Leakage Pathways and Resultant Leakage Activity For the purposes of evaluating the radiological consequences due to the postulated waste gas decay tank rupture, the resultant activity is conservatively assumed to be released directly to the environment during the 2-hour period immediately following the occurrence of the accident. This is a considerably higher release rate than that based on the actual building exhaust ventilation rate. Therefore, the results of the analysis are based on the most conservative pathway available.

15.7.1.5.2 Identification of Uncertainties and Conservatisms in the Analysis The uncertainties and conservatisms in the assumptions used to evaluate the radiological consequences of a waste gas decay tank rupture result from assumptions

ULNRC-06636 Page 162 of 374 were analyzed, using the models described in Appendix 15A, Sections 15A.2.4 and 15A.2.5, respectively.

15.7.2.5.1.4 Identification of Leakage Pathways and Resultant Leakage Activity For the purposes of evaluating the radiological consequences due to the postulated liquid radwaste tank rupture, the resultant activity is conservatively assumed to be released directly 15A.5.2.3 andto15A.5.2.4, the environment during the 2-hour period immediately following the occurrence of the accident. This is a considerably higher release rate than that based on the actual building exhaust ventilation rate. Therefore, the results of the analysis are based on the most conservative pathway available.

15.7.2.5.2 Identification of Uncertainties and Conservatisms in the Analysis The uncertainties and conservatisms in the assumptions used to evaluate the radiological consequences of the liquid radwaste tank rupture result from assumptions made involving the release of the radioactivity from the tanks and the meteorology assumed for the site.

a. It was assumed that the liquid radwaste tank fails when the inventory in the tank is a maximum. This assumption results in the greatest amount of activity available for release to the environment.

b. The contents of the ruptured tank are assumed to be released over a 2-hour period immediately following the accident. If the contents of the tank were assumed to mix uniformly with the volume of air within the radwaste building where the tanks are located, then, using the actual building exhaust ventilation rate, a considerable amount of holdup time would be gained. This reduces the amount of activity released to the environment due to the natural decay. Also, no credit for iodine removal by the radwaste building HVAC charcoal adsorbers is taken.

c. The meteorological conditions which may be present at the site during the course of the accident are uncertain. However, it is highly unlikely that meteorological conditions assumed will be present during the course of the accident for any extended period of time.

d. A tank is assumed to have collected liquid waste based on operation at 100-percent power with 1 percent failed fuel for an extended period of time, which is eight times higher than under normal operating conditions.

ULNRC-06636 Page 163 of 374 loaded HI-TRAC VW transfer casks to the ISFSI pad is performed within the bounds of the 10 CFR 72.212 Evaluation Report and the HI-STORM UMAX FSAR.

15.7.4.2 Sequence of Events and Systems Operations The first step in fuel handling is the safe shutdown of the reactor. After a radiation survey of the containment, the disassembly of the reactor vessel is started. After disassembly is complete, the first fuel handling is started. It is estimated that the earliest time to first fuel transfer after shutdown is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The fuel handling accident is assumed to occur after a fuel assembly has been transferred through the fuel storage pool transfer gate but before it has been placed in its designated location in the fuel storage racks.

15.7.4.3 Core and System Performance The fuel handling accident in the fuel building does not impact the integrity of the core or its system performance.

15.7.4.4 Barrier Performance A barrier between the released activity and the environment is the reactor building and the fuel building. Since these buildings are designed seismic Category I, it is safe to assume that during the course of a fuel handling accident their integrity is maintained.

This means that the pathway for release of radioactivity for a postulated accident in the fuel building is initially via the auxiliary/fuel building normal exhaust system. After it is isolated on a high radiation signal, the release pathway is via the ESF emergency filtration system. For a postulated accident in the reactor building, the release consists of the total amount of radioactivity which could potentially be released. The fuel storage pool and the refueling pool provide minimum decontamination factors of 100 for iodine.

15.7.4.5 Radiological Consequences as discussed in Section 15.7.4.5.1.2 15.7.4.5.1 Method of Analysis 15.7.4.5.1.1 Physical Model The possibility of a fuel-handling accident is remote because of the many administrative controls and physical limitations imposed on the fuel-handling operations (refer to Section 9.1.4). All refueling operations are conducted in accordance with prescribed procedures.

When transferring irradiated fuel from the core to the fuel storage pool for storage, the reactor cavity and refueling pool are filled with borated water at a boron concentration equal to that in the fuel storage pool, which ensures subcritical conditions in the core even if all rod cluster control (RCC) assemblies were withdrawn. After the reactor head

ULNRC-06636 Enclosure 6 Page 164 of 374 15.7.4.5.1.2 Assumptions and Conditions The major assumptions and parameters assumed in the analysis are itemized in Tables 15.7-7 and 15A-1. 1.183 1.183 In the evaluation of the fuel-handling accident, all the fission product release assumptions of Regulatory Guide 1.25 have been followed. Table 15.7-2 provides a comparison of the design to the requirements of Regulatory Guide 1.25. The following assumptions, related to the release of fission product gases from the damaged fuel assembly, were used in the analyses:

a. The dropped fuel assembly is assumed to be the assembly containing the peak fission product inventory. All the fuel rods contained in the dropped assembly are assumed to be damaged. In addition, for the analyses for the accident in the reactor building the dropped assembly is assumed to damage 20 percent of the rods of an additional assembly.

b. The assembly fission product inventories are based on a radial peaking factor of 1.65.

c. The accident occurs 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown, which is the earliest time fuel-handling operations can begin. Radioactive decay of the fission product inventories was taken into account during this time period.

d. Only that fraction of the fission products which migrates from the fuel matrix to the gap and plenum regions during normal operation was assumed to be available for immediate release to the water following clad damage.

Insert 15.7.4.5.1.2.e

e. The gap activity released to the fuel pool from the damaged fuel rods consists of 10 percent of the total noble gases other than Kr-85, 30 percent of the Kr-85, and 10 percent of the total radioactive iodine contained in the fuel rods at the time of the accident.

f. The pool decontamination factor is 1.0 for noble gases.

g. The effective pool decontamination factor is 100 for iodine.

Insert 15.7.4.5.1.2.g

h. The iodine above the fuel pool is assumed to be composed of 75 percent inorganic and 25 percent organic species.

Insert 15.7.4.5.1.2.h

i. The activity which escapes from the pool is assumed to be available for release to the environment in a time period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

j. No credit for decay or depletion during transit to the site boundary and outer boundary of the low-population zone is assumed.

ULNRC-06636 Page 165 of 374 CALLAWAY - SP from the FHB

k. No credit is taken for mixing or holdup in the fuel building atmosphere. The filter efficiency for the ESF emergency filtration system is assumed to be 90 percent for all species of iodine. 120 0.0

l. The fuel building is switched from the auxiliary/fuel building normal exhaust system to the ESF emergency exhaust system within 90 seconds from the time the activity reaches the exhaust duct. The activity released before completion of the switchover is assumed to be discharged directly to the environment with no credit for filtration or dilution. Even if fuel building ventilation isolation does not occur automatically, the calculated doses will be less than those reported in Table 15.7-8 for the bounding case, inside the reactor building. Response time testing is required per Technical Specification 3.3.8 for the fuel building ventilation isolation function.

m. For the inside the reactor building case, no credit has been taken for the mixing or holdup of the radioactivity in the reactor building atmosphere. It is assumed that no containment coolers or hydrogen mixing fans are operating.

n. All gap activity assumed available for release is assumed to be released over two hours.

15.7.4.5.1.3 Mathematical Models Used in the Analysis Mathematical models used in the analysis are described in the following sections:

a. The mathematical models used to analyze the activity released during the course of the accident are described in Appendix 15A, Section 15A.2.

and control room

b. The atmospheric dispersion factors are calculated, based on the onsite meteorological measurements programs described in Section 2.3 of the Site Addendum, and are provided in Table 15A-2.

c. The thyroid inhalation and total-body immersion doses to a receptor located at the exclusion area boundary and outer boundary of the low population zone are described in Appendix 15A, Sections 15A.2.4 and 15A.2.5, respectively.

15.7.4.5.1.4 Identification of Leakage Pathways and Resultant Leakage Activity For evaluating the radiological consequences due to the postulated fuel-handling accident, the resultant activity is conservatively assumed to be released to the environment during the 0-2-hour period immediately following the occurrence of the accident. This is a considerably higher release rate than that based on the actual ventilation rate. Therefore, the results of the analysis are based on the most conservative pathway available.

Sections 15A.2, 15A.3 and 15A.4

ULNRC-06636 Page 166 of 374 15.7.4.5.2 Identification of Uncertainties and Conservatisms in Analysis The uncertainties and conservatisms in the assumptions used to evaluate the radiological consequences of a fuel-handling accident result from assumptions made involving the amount of fission product gases available for release to the environment and the meteorology present at the site during the course of the accident. The most significant of these assumptions are:

a. It is assumed in the analysis that all the fuel rods in the dropped assembly are damaged. This is a highly conservative assumption since, transferring fuel under strict fuel handling procedures, only under the worst possible circumstances could the dropping of a spent fuel assembly result in damage to all the fuel rods contained in the assembly.

b. The fission product gap inventory in a fuel assembly is dependent on the power rating of the assembly and the temperature of the fuel. It has been conservatively assumed that the core has been operating at 100 percent for the entire burnup period. The gap activities are listed in Table 15A-3.

defined by RG 1.183 guidance.

c. Iodine removal from the released fission product gas takes place as the gas rises to the pool surface through the body of liquid in the spent fuel pool. The extent of iodine removal is determined by mass transfer from the gas phase to the surrounding liquid and is controlled by the bubble diameter and contact time of the bubble in the solution. The values used in the analysis result in a release of activity approximately a factor of 5 greater than anticipated. The release of activity from the pool to the containment atmosphere is time-dependent and consequently there would be sufficient time for this activity to mix homogeneously in a significant percent of the containment volume.

d. The ESF emergency filtration system charcoal filters are known to operate with at least a 99-percent efficiency. This means a further reduction in the iodine concentrations and thus a reduction in the thyroid doses at the exclusion area boundary and the outer boundary of the low-population zone.

e. The meteorological conditions which may be present at the site during the course of the accident are uncertain. However, it is highly unlikely that meteorological conditions assumed will be present during the course of the accident for any extended period of time. Therefore, the radiological consequences evaluated, based on the meteorological conditions assumed, are conservative.

ULNRC-06636 Enclosure 6 Page 167 of 374 15.7.4.5.2.1 Filter Loadings The ESF filtration systems which function to limit the consequences of a fuel-handling accident in the fuel building are the ESF emergency filtration system and the control room filtration system.

The activity loadings on the control room charcoal adsorbers as a function of time have been evaluated for the loss-of-coolant accident, Section 15.6.5. Since these filters are capable of accommodating the design basis LOCA fission product iodine loadings, more than adequate design margin is available with respect to postulated fuel-handling accident releases.

The activity loadings on the ESF filtration system charcoal adsorbers have been evaluated in accordance with Regulatory Guide 1.52, which limits the maximum loading to 2.5 mg of iodine per gram of activated charcoal. ,

and Control Room 15.7.4.5.2.2 Doses to Receptor at the Exclusion Area Boundary and Low-Population TEDE doses Zone Outer Boundary worst 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ,

The potential radiological consequences resulting from the occurrence of a postulated fuel-handling accident occurring in the fuel building and in the reactor building have been conservatively analyzed, using assumptions and models described in previous sections.

The total-body dose due to immersion from direct radiation and the thyroid dose due to inhalation have been analyzed for the 0-2-hour dose at the exclusion area boundary and for the duration of the accident (0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) at the low-population zone outer boundary.

The results are listed in Table 15.7-8. The resultant doses are well within the guideline values of 10 CFR 100.

50.67 and Regulatory Guide 1.183 f

Insert 15.7.5 and in the control room for 30 days

ULNRC-06636 Page 168 of 374 Delete Table TABLE 15.7-2 DESIGN COMPARISON TO THE REGULATORY POSITIONS OF REGULATORY GUIDE 1.25 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT IN THE FUEL HANDLING AND STORAGE FACILITY FOR BOILING AND PRESSURIZED WATER REACTORS REVISION 0, DATED MARCH 23, 1972 Regulatory Guide 1.25 Position Case 1 (in Fuel Building) Case 2 (in Reactor Building)

1. The assumptions1 related to the release of radioactive material from the fuel and fuel storage facility as a result of a fuel handling accident are:

a. The accident occurs at a time after shutdown identified in the Complies, except the time after shutdown is Complies, except the time after shutdown is technical specifications as the earliest time fuel handling identified in Section 16.9.5. Accident occurs 72 identified in Section 16.9.5. Accident occurs 72 operations may begin. Radioactive decay of the fission hours after shutdown. hours after shutdown product inventory during the interval between shutdown and commencement of fuel handling operations is taken into consideration.

b. The maximum fuel rod pressurization2 is 1200 psig. Calculations performed as directed by footnote 2 Calculations performed as directed by footnote 2 indicate that the assumed pool water indicate that the assumed pool water decontamination factor is valid for internal decontamination factor is valid for internal pressures up to 1500 psig. pressures up to 1500 psig.

c. The minimum water depth2 between the top of the damaged Complies. Water depth is greater than 23 feet. Complies. Water depth is greater than 23 feet.

fuel rods and the fuel pool surface is 23 feet. The release point is assumed to be at the top of the The release point is assumed to be at the top of the fuel pool storage racks. reactor vessel flange.

d. All of the gap activity in the damaged rods is released and Complies. Complies.

consists of 10% of the total noble gases other than Kr-85, 30% of the Kr-85, and 10% of the total radioactive iodine in the rods at the time of the accident. For the purpose of sizing filters for the fuel handling accident addressed in this guide, 30% of the I-127 and I-129 inventory is assumed to be released from the damaged rods.

e. The values assumed for individual fission product inventories Complies. A peaking factor of 1.65 is used. Complies. A peaking factor is 1.65 is used.

are calculated assuming full power operation at the end of core life immediately preceding shutdown and such calculation should include an appropriate radial peaking factor. The minimum acceptable radial peaking factors are 1.5 for BWR's and 1.65 for PWR's.

f. The iodine gap inventory is composed of inorganic species Complies. Complies.

(99.75%) and organic species (.25%).

ULNRC-06636 Page 169 of 374 TABLE 15.7-2 (Sheet 2)

Delete table Regulatory Guide 1.25 Position Case 1 (in Fuel Building) Case 2 (in Reactor Building)

g. The pool decontamination factors for the inorganic and Complies. Complies.

organic species are 133 and 1, respectively, giving an overall effective decontamination factor of 100 (i.e., 99% of the total iodine released from the damaged rods is retained by the pool water). This difference in decontamination factors for inorganic and organic iodine species results in the iodine above the fuel pool being composed of 75% inorganic and 25% organic species.

h. The retention of noble gases in the pool is negligible (i.e., Complies. A decontamination factor of 1 is used. Complies. A decontamination factor of 1 is used.

decontamination factor of 1).

i. The radioactive material that escapes from the pool to the Complies. A 0-2 hour release from the pool to the Complies. A 0-2 hour release from the pool to the building is released from the building3 over a 2-hour time building to the environment is assumed. building to the environment is assumed.

period.

j. If it can be shown that the building atmosphere is exhausted Not applicable; complies with Regulatory Guide No credit is taken for the normal purge filters.

through adsorbers designed to remove iodine, the removal 1.52 as described in Table 9.4-2.

efficiency is 90% for inorganic species and 70% for organic species.4

k. The effluent from the filter system passes directly to the Complies. Complies.

emergency exhaust system without mixing5 in the surrounding building atmosphere and is then released (as an elevated plume for those facilities with stacks6).

2. The assumptions for atmospheric diffusion are: Short-term atmospheric dispersion factors corresponding to ground level release and accident conditions were based on meteorological measurement programs described in Section 2.3 of the Site Addendum.

a. Ground Level Releases The dispersion factors are in compliance with the methodology described in Regulatory Guide 1.XXX (see Site Addendum Section 2.3.4.2.1) and represent the worst of the 5 percent overall site meteorology and (1) The basic equation for atmospheric diffusion from a the 0.5 percent worst sector meteorology.

ground level point source is:

---

1

/Q = ---------

uz Where:

= the short term average centerline value of the ground level concentration (curies/m3)

ULNRC-06636 Page 170 of 374 TABLE 15.7-2 (Sheet 3)

Regulatory Guide 1.25 Position Case 1 (in Fuel Building) Case 2 (in Reactor Building)

Q = amount of material released (curies/sec)

Delete table u = windspeed (meters/sec) y = the horizontal standard deviation of the plume (meters) [See Figure V-1, Page 48, Nuclear Safety, June 1961, Volume 2, Number 4, "Use of Routine Meteorological Observations for Estimating Atmospheric Dispersion," F. A.

Gifford, Jr.]

z = the vertical standard deviation of the plume (meters) [See Figure V-2, Page 48, Nuclear Safety, June 1961, Volume 2, Number 4, "Use of Routine Meteorological Observations for Estimating Atmospheric Dispersion," F. A.

Gifford, Jr.]

(2) For ground level releases, atmospheric diffusion factors7 used in evaluating the radiological consequences of the accident addressed in this guide are based on the following assumptions:

(a) windspeed of 1 meter/sec; (b) uniform wind direction; (c) Pasquill diffusion category F.

(3) Figure 1 is a plot of atmospheric diffusion factors (/Q) versus distance derived by use of the equation for a ground level release given in regulatory position 2.a.(1) and under the meteorological conditions given in regulatory position 2.a.(2).

ULNRC-06636 Page 171 of 374 TABLE 15.7-2 (Sheet 4)

Regulatory Guide 1.25 Position Case 1 (in Fuel Building) Case 2 (in Reactor Building)

(4) Atmospheric diffusion factors for ground level releases may be reduced by a factor ranging from one to a maximum of three (see Figure 2) for additional dispersion produced by the turbulent wake ofDelete the table reactor building. The volumetric building wake correction as defined in Subdivision 3-3.5.2 of Meteorology and Atomic Energy-1968, is used with a shape factor of 1/2 and the minimum cross-sectional area of the reactor building only.

b. Elevated Releases (1) The basic equation for atmospheric diffusion from an Not applicable. Ground level releases were Not applicable. Ground level releases were elevated release is: assumed. assumed.

z -h /2 2 2

e /Q = -----------------------

uyz Where:

= the short term average centerline value of the ground level concentration (curies/m3)

Q = amount of material released (curies/sec) u = windspeed (meters/sec) y = the horizontal standard deviation of the plume (meters) [See Figure V-1, Page 48, Nuclear Safety, June 1961, Volume 2, Number 4, "Use of Routine Meteorological Observations for Estimating Atmospheric Dispersion," F. A.

Gifford, Jr.]

z = the vertical standard deviation of the plume (meters) [See Figure V-2, Page 48, Nuclear Safety, June 1961, Volume 2, Number 4, Use of Routine Meteorological Observations for Estimating Atmospheric Dispersion," F. A.

Gifford, Jr.]

ULNRC-06636 Page 172 of 374 TABLE 15.7-2 (Sheet 5)

Regulatory Guide 1.25 Position Case 1 (in Fuel Building) Case 2 (in Reactor Building)

Delete table h = effective height of release (meters)

(2) For elevated release; atmospheric diffusion factors7 used in evaluating the radiological consequences of the accident addressed in this guide are based on the following assumptions:

(a) windspeed of 1 meter/sec; (b) uniform wind direction; (c) envelope of Pasquill diffusion categories for various release heights; (d) a fumigation condition exists at the time of the accident.8 (3) Figure 3 is a plot of atmospheric diffusion factors versus distance for an elevated release assuming no fumigation, and Figure 4 is for an elevated release with fumigation.

(4) Elevated releases are considered to be at a height equal to no more than the actual stack height. Certain site conditions may exist, such as surrounding elevated topography or nearby structures, which will have the effect of reducing the effective stack height. The degree of stack height reduction will be evaluated on an individual case basis.

3. The following assumptions and equations may be used to obtain conservative approximations of thyroid dose from the inhalation of radioiodine and external whole body dose from radioactive clouds:

a. The assumptions relative to inhalation thyroid dose Complies. See Appendix 15A, Section 15A.2.4. Complies. See Appendix 15A, Section 15A.2.4.

approximations are:

(1) The receptor is located at a point on or beyond the site boundary where the maximum ground level concentration is expected to occur.

ULNRC-06636 Page 173 of 374 TABLE 15.7-2 (Sheet 6)

Regulatory Guide 1.25 Position Case 1 (in Fuel Building) Case 2 (in Reactor Building)

(2) No correction is made for depletion of the effluent plume of radioiodine due to deposition on the ground, or for the radiological decay of radioiodine in transit.

Delete table (3) Inhalation thyroid doses may be approximated by use of the following equation:

g F IFPBR(/Q)

D =

(DFp)(DFf)

Where:

D = thyroid dose (rads)

Fg = fraction of fuel rod iodine inventory in fuel rod void space (0.1)

I = core iodine inventory at time of accident (curies)

F = fraction of core damaged so as to release void space iodine P = fuel peaking factor B = Breathing rate = 3.47 x 10-4 cubic meters per second (i.e., 10 cubic meters per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> work day as recommended by the ICRP)

DFp = effective iodine decontamination factor for pool water DFf = effective iodine decontamination factor for filters (if present)

/Q = atmospheric diffusion factor at receptor location (sec/mµ)

ULNRC-06636 Page 174 of 374 TABLE 15.7-2 (Sheet 7)

Regulatory Guide 1.25 Position Case 1 (in Fuel Building) Case 2 (in Reactor Building)

R = adult thyroid dose conversion factor for the iodine isotope of interest (rads per curie). Dose conversion factors for Iodine 131-135 are listed in Table I.9 These values were derived from "standard man" parameters recommended in ICRP Publication 2.10 TABLE 1 Adult Inhalation Thyroid Dose Conversion Factors Table 1; the thyroid dose conversion factors given Table 1; the thyroid dose conversion factors given in ICRP-30 are used. See Table 15A-4. in ICRP-30 are used. See Table 15A-4.

Iodine Conversion Factor (R)

Isotope (Rads/curie inhaled)

Delete table 131 1.48 x 106 132 5.35 x 104 133 4.0 x 105 134 2.5 x 104 135 1.24 x 105

b. The assumptions relative to external whole body dose Complies. See Appendix 15A, Section 15A.2.5. Complies. See Appendix 15A, Section 15A.2.5.

approximations are:

(1) The receptor is located at a point on or beyond the site boundary where the maximum ground level concentration is expected to occur.

(2) External whole body doses are calculated using "Infinite (2) See Table 15A-4 for whole body dose conversion factors from Federal Guidance Report 12.

Cloud" assumptions, i.e., the dimensions of the cloud are assumed to be large compared to the distance that the gamma rays and beta particles travel. The dose at any distance from the reactor is calculated based on the maximum ground level concentration at that distance.

For an infinite uniform cloud containing curies of beta radioactivity per cubic meter, the beta dose rate in air at the cloud center is:11

ULNRC-06636 Page 175 of 374 TABLE 15.7-2 (Sheet 8)

Regulatory Guide 1.25 Position Case 1 (in Fuel Building) Case 2 (in Reactor Building)

D = 0.457 E Delete table Where:

D = beta dose rate from an infinite cloud (rad/sec)

E = average beta energy per disintegration (Mev/dis)

= concentration of beta or gamma emitting isotope in the cloud (curie/m3)

Because of the limited range of beta particles in tissue, the surface body dose rate from beta emitters in the infinite cloud can be approximated as being one-half this amount or:

D = 0.23 E For gamma emitting material the dose rate in tissue at the cloud center is:

D = 0.507 E Where:

D = gamma dose rate from an infinite cloud (rad/sec)

ULNRC-06636 Page 176 of 374 TABLE 15.7-2 (Sheet 9)

Regulatory Guide 1.25 Position Case 1 (in Fuel Building) Case 2 (in Reactor Building)

E = average gamma energy per disintegration (MEV/dis)

Delete table However, because of the presence of the ground, the receptor is assumed to be exposed to only one-half of the cloud (semi-infinite) and the equation becomes:

D' = 0.25 E Thus, the total beta or gamma dose to an individual located at the center of the cloud path may be approximated as:

D = 0.23 E or D = 0.25 E Where is the concentration time integral for the cloud (curie sec/m3)

(3) The beta and gamma energies emitted per disintegration, as given in Table of Isotopes,12 are averaged and used according to the methods described in ICRP Publication 2.

Notes:

1. The assumptions given are valid only for oxide fuels of the types currently in use and in cases where the following conditions are not exceeded:

a. Peak linear power density of 20.5 kW/ft for the highest power assembly discharged.

b. Maximum center-line operating fuel temperature less than 4500°F for this assembly.

ULNRC-06636 Page 177 of 374 TABLE 15.7-2 (Sheet 10)

Regulatory Guide 1.25 Position Case 1 (in Fuel Building) Case 2 (in Reactor Building)

c. Average burnup for the peak assembly of 25,000 MWD/ton Gap fractions of 10% remain valid for fuel Gap fractions of 10% remain valid for fuel or less (this corresponds to a peak local burnup of about assemblies up to 33,000 MWD/MTU. Beyond this assemblies up to 33,000 MWD/MTU. Beyond this 45,000 MWD/ton). burnup, a 12% gap fraction will be used. burnup, a 12% gap fraction will be used.

2. For release pressures greater than 1200 psig and water depths less than 23 feet, the iodine decontamination factors will be less Delete table than those assumed in this guide and must be calculated on an individual case basis using assumptions comparable to conservatism to those of this guide.

3. The effectiveness of features provided to reduce the amount of radioactive material available for release to the environment will be evaluated on an individual case basis.

4. These efficiencies are based upon a 2-inch charcoal bed depth with 1/4 second residence time. Efficiencies may be different for other systems and must be calculated on an individual case basis.

5. Credit for mixing will be allowed in some cases; the amount of credit will be evaluated on an individual case basis.

6. Credit for an elevated release will be given only if the point of release is (a) more than two and one-half times the height of any structure close enough to affect the dispersion of the plume or (b) located far enough from any structure which could affect the dispersion of the plume. For those plants without stacks the atmospheric diffusion factors assuming ground level release given in regulatory position 2.b. should be used.

7. These diffusion factors should be used until adequate site meteorological data are obtained. In some cases, available information on such site conditions as meteorology, topography and geographical location may dictate the use of more restrictive parameters to ensure a conservative estimate of potential offsite exposures.

8. For sites located more than 2 miles from large bodies of water such as oceans or one of the Great Lakes, a fumigation condition is assumed to exist at the time of the accident and continue for 1/2 hour. For sites located less than 2 miles from large bodies of water a fumigation condition is assumed to exist at the time of accident and continue for the duration of the release (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />).

Rev. OL-19

ULNRC-06636 Page 178 of 374 TABLE 15.7-2 (Sheet 11)

Delete table Regulatory Guide 1.25 Position Case 1 (in Fuel Building) Case 2 (in Reactor Building)

9. Dose conversion factors taken from "Calculation of Distance Factors for Power and Test Reactor Sites," TID-14844, J. J.

DiNunno, R. E. Baker, F. D. Anderson, and R. L. Waterfield (1962).

10. Recommendations of the International Commission on Radiological Protection, "Report of Committee II on Permissible Dose for Internal Radiation (1959)," ICRP Publication 2, (New York: Permagon Press, 1960).

11. Meteorology and Atomic Energy-1968, Chapter 7.

12. C. M. Lederer, J. M. Hollander, and I. Perlman, Table of Isotopes, Sixth Edition (New York: John Wiley and Sons, Inc. 1967).

Rev. OL-19

ULNRC-06636 Page 179 of 374 TABLE 15.7-7 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A FUEL-HANDLING ACCIDENT In Fuel Building In Reactor Building I. Source Data

a. Core power level, MWt 3,636 3,636

b. Radial peaking factor 1.65 1.65

c. Decay time, hours 72 72

d. Number of fuel assemblies 1.0 1.2 affected

e. Fraction of fission product 1.183 1.183 gases contained in the gap region of the fuel assembly Per R.G. 1.25 Per R.G. 1.25 II. Atmospheric Dispersion Factors See Table 15A-2 See Table 15A-2 III. Activity Release Data

a. Percent of affected fuel assemblies gap activity released to pool 100 100

b. Pool decontamination 200 200 factors

1. Iodine (effective) 100 100

2. Noble gas 1 1

c. Filter efficiency, 0 until isolation 0 assumed after Isolation percent 90 thereafter 0

d. Building mixing volumes assumed, percent of total volume 0 0

e. HVAC exhaust rate, 20,000 until isolation Activity cfm Activity completely 9,000 thereafter completely released over 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> released 120 over 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

f. Building isolation time, 90 sec 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

g. Activity release period, hrs 2 2 Control room isolation 120 sec

ULNRC-06636 Page 180 of 374 CALLAWAY - SP TABLE 15.7-8 RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT (rem TEDE)

Doses (rem)

In Fuel Building 7.8E-01 Exclusion Area Boundary (0-2 hr)

Thyroid 6.40 Whole-body 0.235 2.7E-01 Low Population Zone Outer Boundary (duration)

Control Room Thyroid 0.640 6.5E-01 Whole-body 0.0235 In Reactor Building Exclusion Area Boundary (0-2 hr) 9.7E-01 Thyroid 61.7 Whole-body 0.359 3.4E-01 Low Population Zone Outer Boundary (duration)

Thyroid 6.17 Whole-body 0.0359 Control Room 1.58

ULNRC-06636 Enclosure 6 375 Page 181 of 374

. The higher value between the 95th percentile overall -site CALLAWAY SP concentration factors for a 90° direction window and the maximum sector concentration factors are used for the off-site doses (EAB and LPZ),

and the 95th percentile concentration factors for a 90° direction window are used for the on-site doses APPENDIX (CR and TSC). These 15A - ACCIDENT dispersion factors ANALYSIS are given inRADIOLOGICAL Table 15A-2 (seeCONSEQUENCES Section 2.3.4 and the Site EVALUATION MODELS AND PARAMETERS 15A.1 GENERAL ACCIDENT PARAMETERS 1.145 (1983)

This section contains the parameters used in analyzing the radiological consequences of postulated accidents. Table 15A-1 contains the general parameters used in all the and ventspecific accident analyses. For parameters stack toonly particular accidents, refer to that accident parameter section. The site specific, ground-level release, short-term dispersion factors (for accidents, ground-level releases are assumed) are based on Regulatory Guide 1.XXX (1978) (Ref. 1) methodology and the 0.5 percent worst-sector meteorology and these are given in Table 15A-2 (see Section 2.3.4 and the Site Addendum for additional details on meteorology). The core and gap inventories are given in Table 15A-3. The thyroid (via inhalation pathway), beta skin, and total-body (via TEDE submersion pathway) dose factors based on References 2 and 3 are given in Table 15A-4.

8 and 9 15A.2 OFFSITE RADIOLOGICAL CONSEQUENCES CALCULATIONAL MODELS This section presents the models and equations used for calculating the integrated activity released to the environment, the accident flow paths, and the equations for dose calculations. Two major release models are considered: (1) a single holdup system with no internal cleanup and (2) a holdup system wherein a two-region spray model is used for internal cleanup. via the A-E pathway equipment (pathway C-D-E) 15A.2.1 ACCIDENT RELEASE PATHWAYS The release pathways for the major accidents are given in Figure 15A-1. The accidents and their pathways are as follows:

LOCA: Immediately following a postulated loss-of-coolant accident (LOCA), the release of radioactivity from the containment is to the environment with the containment spray and ESF systems in full operation. The release in this case is calculated using equation (12

1) which takes into account a two-region spray model within the containment. The release of radioactivity to the environment due to assumed ESF system leakages in the auxiliary building will be via ESF filters and is calculated using equation (5), using a factor of 0.01 to account for the combined effect of the airborne fraction of radioiodine and the ESF filter efficiency. The total removal constant, 1, for this release pathway includes decay (1d) and release (1l) removal constants associated with holdup and mixing in the sumps (no holdup or mixing assumed in the auxiliary building); however, no internal removal constant (r) is assumed. The release of radioactivity to the environment due to assumed leakage from the RWST is calculated using in equation (5a 1) via the R-E pathway 15.A-1 Rev. OL-16

ULNRC-06636 Enclosure 6 Section Page 15A.2.2 182 of 374 CALLAWAY - SP from Section 15A.2.2.b. The release of radioactivity to the environment due to the assumed operation of the containment mini-purge system for the first few seconds after a LOCA is calculated using equation (5) from Section 15A.2.2.a, with no credit for filtration, plateout, or valve stroke time. E WGDTR: The activity release to the environment due to waste gas decay tank rupture (WGDTR) and includes will be direct consideration andand of pH iodine with no holdup. The release pathway is A'-D. The unfiltered, total evolution in activity releasewith conformance in this case is ther efore assumed to be the initial source activity NUREG/

itself.

CR-5950. B-E FHA: The release to the environment due to a fuel handling accident (FHA) in the fuel building is via filters.

vents. The release pathway is B-C-D. Since the release is calculated without any credit for holdup in the fuel building, the total release will be the product of the initial activity and the filter nonremoval efficiency fraction (for noble gases, the nonremoval efficiency fraction is 1). The release of radioactivity to the environment due to FHA inside the containment is direct and unfiltered, via the A-D pathway without any credit for holdup (see Figure 15A-1). The release is calculated assuming the total gap inventoryinclude a partition factor of 1.2 assemblies of 0.01over is released for non-a two hour period, reduced only by the pool RE decontamination noble gases, and 1 for noble gases.

factor. A-E CAE: Radioactivity release to the environment due to the control assembly rod ejection (CAE) accident is direct and unfiltered. The releases from the primary system are calculated using equation 1 5 which considers holdup in the single-region primary system RE (the spray removal is not assumed); the secondary (steam) releases via the relief valves are calculated without any holdup. The pathways for these releases are A-D and A'-D. A'-E.

MSLB, SGTR: Radioactivity releases to the environment due to main steam line break A-E (MSLB) or steam generator tube rupture (SGTR) accidents are direct and unfiltered with no holdup via the A'-D A'-E pathway. The activity release calculations for these accidents are complex, involving spiking effects, time-dependent flashing fractions, and scrubbing of flashed activities; the release calculations are described in those sections that address these accidents.

Insert 15A.2.2 As used in the radiological consequence evaluations, partition factor refers to the fraction of the total release that is airborne.

15A.2.2.a SINGLE - REGION RELEASE MODEL It is assumed that any activity released to the holdup system instantaneously diffuses to uniformly occupy the system volume.

The following equations are used to calculate the integrated activity released from postulated accidents.

A1(0) = initial source activity at time to, Ci

ULNRC-06636 Page 183 of 374 A1(t) = source activity at time t seconds, Ci A1(t) = -1t (1)

A1 (0)e Delete where 1 = total removal constant from primary holdup system, sec-1 1 = d + 1 + r (2) where d = decay removal constant, sec-1 1 = primary holdup leak or release rate, sec-1 r = internal removal constant (i.e., sprays, plateout, etc.), sec-1 Thus, the direct release rate to the atmosphere from the primary holdup system Ru(t) = 1A1(t)

Ru(t) = unfiltered release rate (Ci/sec) (3)

The integrated activity release is the integral of the above equation.

t t

-1t IAR(t) = Ru(t) = 1A1(o)e (4) o o This yields:

-1t IAR(t) = ( 1 A 1 (o) 1 )(1 - e ) (5) 15A.2.2.b TWO - REGION RELEASE MODEL FOR DOSES DUE TO LEAKAGE FROM THE ECCS SUMPS TO THE RWST It is assumed that the activity released to the holdup system (in this case, the containment recirculation sumps) instantaneously diffuses to uniformly occupy the sump volume. Removal mechanisms from the sumps include decay and release (i.e., leakage) to the RWST. Expanding upon the equations developed in Section 15A.2.2.a above, the release rate from the RWST to the environment is given by R2(t) = 0.1 2A2(t) (1a)

ULNRC-06636 Page 184 of 374 where R2(t) = the unfiltered release rate from the RWST vent, Ci/sec 0.1 = assumed percent of radioiodine released to the RWST that becomes airborne Delete 2 = release rate constant for leakage from the RWST to the environment, based on an assumed 3 gpm leak rate from the sumps that is uniformly mixed in the RWST volume, hr-1 A2(t) = RWST activity, Ci The activity time rate of change for the RWST is given by dA2 (t) (2a)

---

= 1A1(t) - 2A2(t) dt where 1 = release rate constant for leakage from the uniformly mixed sumps to the RWST, based on an assumed 3 gpm leak rate from the sumps to the RWST, hr-1 2 = d + 2 A1(t) = containment sump activity, Ci Using equation (1) from Section 15A.2.2.a above, dA2 (t) -1t (3a)

---

+ 2A2 (t) = 1A1(0)e dt where A1(0) = initial sump activity, assumed to be 50% of the initial iodine core inventory 1 = d + 1 d = decay removal constant for the sumps, hr-1 The solution of this equation is given by A 1 (0) 1 -1t - t -2t (4a)

A 2 (t) = -----------------------[e - e 2 ] + A2(0)e (2 - 1)

Therefore, the integrated release from the RWST is given by

ULNRC-06636 Page 185 of 374 t

IAR 2 (t) = R2(t)dt o

t IAR 2 (t) = 0.1 2 A 2 (t)dt Delete o

(5a) 0.1A 1 (0) 1 2 0.1A 2 (0) 2 IAR 2 (t) = ----------------------------------------[ (1 - e-1t ) + 1(e -2t - 1)] + -------------------------------- -2t (1 - e )

2 21(2 - 1) 2 15A.2.3 TWO - REGION SPRAY MODEL IN CONTAINMENT (LOCA)

A two-region spray model is used to calculate the integrated activity released to the environment. The model consists of a sprayed and unsprayed region in containment and a constant mixing rate between them.

As it is assumed that there are no sources after initial release of the fission products, the remaining processes are removal and transfer so that the multivolume containment is described by a system of coupled first-order differential equations of the form Ki -n -1 n-1 dAi Ai Ai (6)

---

Vi dt = - - ijAi - Q i ----- +

Vi Qi----

j=1 =1 =1 where Ai = fission product activity in volume i, Ci n = number of volumes considered in the model Qi = transfer rate from volume i to volume , cc/sec Vi = volume of the ith compartment, cc ij = removal rate of the jth removal process in volume i, sec-1 Ki = total number of removal processes in volume i This system of equations is readily solved if the coefficients are known.

ULNRC-06636 Page 186 of 374 For a two-region model, the above system reduces to K1 A1 A2 (6a) dA1

---

= - - 1jA1 - Q12 ----- + Q21------

dt V1 V2 j=1 Delete dA2 K2 A2 A1 (6b)

---

= - - 2jA1 - Q21 ----- + Q12------

dt V2 V1 j=1 To solve the two preceding equations, use the method of Laplace transforms.

df (t )

use L ----------- = SF(s) - f(0) dt where f(t) is any regular function of t, L is the symbol of L-transform S is the Laplace variable and F(s) is the Laplace transform of f(t).

The solutions to the equations are:

K1 (7)

Q21 Q1 2 A 2(0)--------- - A 1 (0) 1j + --------

S2t S1 t V2 j = 1 V1 S1 t S2t A (t) = -A 1 (0) -(S e - S e ) + --------------------------------------------- ------------------------------- (e - e )

1 1 2 S 1- S2 S1 - S2 and K2 (8)

Q1-- Q A 1(0)----- - A 2 (0) 2i + ------2--1 S2t S1 t V1 j = 1 V2 S1 t S2t A (t) = -A 2 (0) -(S e - S e ) + --------------------------------------------- ------------------------------- (e - e )

2 1 2 S 1- S2 S1 - S2

ULNRC-06636 Page 187 of 374 K1 K2 Q12 Q21 where, S1, S2 = -1 2 1j + 2j + --------- + --------

V1 V2 j=1 j=1 Delete K 2 K1 2

Q12 Q K1 K2 1Q K2 K 12 (8)

+/-1 2 1j + 2j + --------- + --------- - 4 1j + 2j + ---------Q 21 2 12 1 V V 2j + --------- 1j 2 V1 V1 j = 1 j=1 1 j = 1 j=1 j=1 j=1 At time t1 > t0, according to equations (7) and (8):

K Q2-- 1 Q A 2(t 0 )----- - A 1 (t 0 ) 1j + -----1--

V2 V1 A (t) = --------------------------------------------------------------------------------

j =1 (eS1(t1 - t0) - eS2(t1 - t0))

1 1 S -S --- ----

1 2 (10a)

A1(t0) S2(t1 - t0)

- S 2e S1(t1 - t0)

)

+ -------------------

S1 - S(S e1 2

Q K2 Q2-- 12 A 1(t 0 )--------- - A 2 (t 0 ) 2j + -----

V1 V2 A (t) = --------------------------------------------------------------------------------

j =1 (eS1(t1 - t0) - eS2(t1 - t0))

2 1 S -S --- ----

1 2 A2(t0) S2(t1 - t0) S1(t1 - t0) (10b)

+ ------------------- e1 - S 2e )

S1 - S(S 2 Where S1, S2 are given in equation (9).

A(t1) = Activity in the primary containment at any time, t1

= A1(t1) + A2(t1)

= K1 K2 A 1 (t 0 ) 1j + A 2 (t 0 ) 2j S (t - t ) S (t - t )

---

j---=----1---------------------------------j---=----1--------(e 2 1 0 - e 1 1 0 )

S1 - S2

ULNRC-06636 Page 188 of 374 1

A0 (t ) 2+ A0 (t ) S2(t1 - t0) S1(t1 - t0) (11)

+ (S e

---

- S 2e )

S1- S 2 1 To calculate the integrated activity released to the atmosphere we must first calculate the release rate of activity. This is found from R(t) = 1A(t) Delete (11a)

The integrated activity released from time t0 - t1 is then t

IAR = R(t)dt t0 This solves as S2(t1 - t0) S (t - t ) S2 S1(t1 - t0) (12)

IAR = 1 m 2e-------------------------------- e 1 1 0 - 1 + m 3 ------

---

(e - 1)

S2 S1 S1 where K1 K2 (13a)

A 1 (t 0 ) 1j + A2(t0 ) 2i j=1 j = --1 m2 = ------------------------------------------------------------ ---------

S1 - S2 and A 1 (t 0 ) + A 2 (t 0) (13b) m3 = ---------------------------------------

S1 - S2 15A.2.4 OFFSITE THYROID DOSE CALCULATION MODEL Offsite thyroid doses are calculated using the equation:

(14)

DTH = DCFThi (IAR)ij j (BR)j ( Q)j i

where (IAR)ij = integrated activity of isotope i released* during the time interval j in Ci

ULNRC-06636 Page 189 of 374 and (BR)j = breathing rate during the time interval j in meter3/second

(/Q)j = offsite atmospheric dispersion factor during time interval j in second/meter3 (DCF)Thi = thyroid dose conversion factor via inhalation for isotope i in rem/

Ci Delete DTh = thyroid dose via inhalation in rems

• No credit is taken for cloud depletion by ground deposition and radioactive decay dur-ing transport to the exclusion area boundary or the outer boundary of the low-popula-tion zone.

15A.2.5 OFFSITE TOTAL-BODY DOSE CALCULATIONAL MODEL Assuming a semi-infinite cloud of gamma emitters, offsite total-body doses are calculated using the equation:

DTB = DCFi (IAR)ij j ( Q)j i

where (IAR)ij = integrated activity of isotope i released* during the jth time interval in Ci and (/Q)j = offsite atmospheric dispersion factor during time interval j in second/meter3 (DCF)i = total-body gamma dose conversion factor for the ith isotope in rem-meter3/Ci-sec DTB = total-body dose in rems 15A.3 CONTROL ROOM RADIOLOGICAL CONSEQUENCES CALCULATIONAL MODELS Only radiation doses to a control room operator due to postulated LOCA are presented in this chapter since a study of the radiological consequences in the control room due to various postulated accidents indicate that the LOCA is the limiting case.

ULNRC-06636 Page 190 191 of 374 376 CALLAWAY - SP Unfiltered air may also leak into the equipment room and from the equipment room into the control building 15A.3.1 and control INTEGRATED room filtration ACTIVITY inlet plenum.¶ IN CONTROL ROOM Experience gained from the development and performance of inleakage measurement testing using air Make-up theisatmospheric brought intotracer depletion the control room test viamethodology the control roomled tofiltration the identification of three system which drawsinteracting in air fromzones for modeling the control CREVS building. Outside operation and theinto air is brought analysis of control the control room dose. The building through model safety grade applies thefilters via the alternate control source term room pressurization established fan. Some___

per Amendment unfiltered air of the Callaway alsoOperating may leak License.

into the control buildingthe Accordingly, anddosecontrol roommodel analysis via anincludes assumed a inleakage three-zonerate.

model for Allowable inleakage values for the Control Room and Control Building are which the atmospheric tracer depletion test method explicitly determines inleakage values fordependent upon theeach other.

control Seeenvelope room Figure 15A-2 (CRE), forcontrol allowable inleakage building combinations envelope (CBE) andfor the Control equipment room Room and Control Building. The activity concentrations at the control envelope (ERE) in accordance with the requirements of the Technical Specifications.¶ building intake for each time interval are found by multiplying the activity release to the environment by the appropriate /Q for that time interval. The flow path model is shown below.

in Figure 15A-3 Once activity is brought into the control building, mixing within the control building is afforded by the control room pressurization fan. Only one-half of the control building volume is considered as the mixing volume. The control room filtration system fan takes air from the control building and the control room (recirculation) and discharges to the control room through the control room filtration safety grade filters.

The control room ventilation isolation signal (CRVIS) starts both trains of the control room pressurization system and the control room filtration system. For the determination of doses to control room personnel, the worst single failure has been ascertained to be the failure of the filtration fan in one of the two filtration system trains. Operator action is required to isolate the train with the failed filtration fan. At the same time, one train of the control room pressurization system will also be isolated. Prior to isolation, a potential pathway exists allowing air from the control building to enter the control room, bypassing the control room pressurization fan and one control room filtration fan operate for the duration of the accident. No bypass pathways then exist for unfiltered air to enter the control room. This failure allows control building air to bypass the control room filtration filter and enter the control room.

Owning to this single failure of the control room filtration fan, the assumed failure of one of the two containment spray (CS) trains, and two of the four hydrogen mixing subsystem fans, inherent in the LOCA analysis parameters given in Table 15.6-6 should not be Owing 15.A-10 Rev. OL-16 10/07

ULNRC-06636 Enclosure 6 Page 191 of 374 applied in this analysis. With both trains of CS and four hydrogen mixing fans operating, Inmore accordance volumetric withcoverage of theofcontainment Section 5.1.3 spray and Regulatory Guide 1.183,more amixing between the new sprayed and unsprayed conservative combination of cases may regions would be expected, thereby used instead of iving g much greater iodine removal within comparing the containment the results of multiple atmosphere. However, individual cases with sesthe do to control room personnel have alternative single failures. been based on the LOCA analysis parameters given in Table 15.6-6.

The activity in the control building and control room is calculated by solving the following coupled set of first order differential equations.

dACB(t)

---

= [(1 - )F + F ] Q[R(t)] + A (t) - A (t) 1 2 4 CR 3 CB dt dA CR (t)

---

= [(1 - )3f + 3u]ACB(t) - 4ACR(t) + F6[R(t)] Q dt where ACB(t) = activity in control building at time t, curies ACR(t) = activity in control room at time t, curies

= filter efficiency, RADTRAD-NAI by the codefraction which uses Equation 1 In Section 15A.2.2 for all compartments for the duration of the event. The

= fraction totaleffective of control (TEDE) dose equivalent room outleakage and to the which returns is then determined Control Buildingfor reported by RADTRAD-NAI mixing volume.

the full 30 day event.

F1 = filtered intake rate, meter3/sec F2 = unfiltered Control Building intake (inleakage), meter3, /sec F6 = unfiltered Control Room in leakage, meter3/sec

/Q = atmospheric dispersion factor, sec/meter3 R(t) = activity release rate in Ci/sec as given in Equation 3 of Section 15A.2.2.a, Equation 1a of Section 15A.2.2.b, or Equations 11 and 11a of Section 15A.2.3 3 = d + 3 + 3f + 3u, total removal rate from the control building, sec-1 d = isotopic decay constant, sec-1 3 = outleakage to atmosphere from the control building (=((F1

+ F2) - ( - 1)(F3 + F4)+F6)/VCB with VCB being control building mixing volume in meter3), sec-1

ULNRC-06636 Page 192 of 374 3f = filtered flow from control building into control room (=F3/

VCB, F3 in meter3/sec), sec-1 3u = unfiltered flow from control building into control room (=F4/

Delete VCB, F4 in me ter3/sec), sec-1 4 = d + r + 4, total removal rate from the control room, sec-1 r = recirculation removal rate (=F5/VCR with F5 being recirculation flow rate in meter3/sec through filter with efficiency and VCR being control room volume in meter3), sec-1 4 = leakage to all destinations from the control room (= [F3 +

F4+F6 ]/VCR), sec-1 Upon solving this coupled set of differential equations, the integrated activity in the control room (IACR) is determined by the expression t

IACR(t) = o ACR(t)dt This IACR(t) is used to calculate the doses to the operator in the control room. This activity is multiplied by an occupancy factor which accounts for the time fraction the operator is in the control room.

15A.3.2 CONTROL ROOM THYROID DOSE CALCULATIONAL MODEL Control room thyroid doses via inhalation pathway are calculated using the following equation:

BR DTH - CR = --------- DCF Thi ( IA CRij) x Oj VCR i j where DTh-CR = control room thyroid dose in rem and

ULNRC-06636 Page 193 of 374 BR = breathing rate assumed to be always 3.47 x 10-4 meter3/

second VCR = volume of the control room in cubic meters Delete DCFThi = thyroid dose conversion factor for adult via inhalation in rem/Ci for isotope i IACRij = integrated activity in control room in Ci-sec for isotope i during time interval j oj = control room occupancy fraction during time interval j 15A.3.3 CONTROL ROOM BETA-SKIN DOSE CALCULATIONAL MODEL The beta-skin doses to a control room operator are calculated using the following equation:

1 D - CR = ---------- DCF i ( IACRij) x 0j V CR i j where D-CR and DCFi are the beta-skin doses in the control room in rem and the beta-skin dose conversion factor for isotope i in rem-meter3/Ci-sec, respectively. The other symbols are explained in Section 15A.3.4.

15A.3.4 CONTROL ROOM TOTAL-BODY DOSE CALCULATION Due to the finite structure of the control room, the total-body gamma doses to a control room operator will be substantially less than what they would be due to immersion in an infinite cloud of gamma emitters. The finite cloud gamma doses are calculated using Murphy's method (Ref. 4) which models the control room as a hemisphere. The following equation is used:

DTB - CR = ---------1----------

DCF i ( IACRij ) x oj V (GF)

CR i j where GF = dose reduction due to control room geometry factor GF = 1173/(V1)0.338

ULNRC-06636 Page 194 of 374 V1 = volume of the control room in cubic feet DTB-CR = total-body dose in the control room in rem, and other quantities have been defined in subsections 15A.2.5 and 15A.3.4.

15A.3.4.1 Model for Radiological Consequences Due to Radioactive Cloud External to the Control Room This dose is calculated based on the semi-infinite cloud model which is modified using the protection factors described in Section 7.5.4 of Reference 5 to account for the control room walls. Insert 15A.4,5 15A.46 REFERENCES 1.145

1. USNRC Regulatory Guide 1.XXX, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," See Appendix 3A and Section 2.3.4.2.1 of the Callaway Site Addendum. Revision 1, Reissued February 1983

2. USNRC Regulatory Guide 1.109, Rev. 1, "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," October 1977.

3a. andData Kocher, D.C., "Nuclear De cay Section 2.3.4.4 ides Occurring in Routine for Radionucl Releases from Nuclear Fuel Cycle Facilities," ORNL/NUREG/TM-102, August 1977.

3b. Berger, M.J., "Beta-Ray Dose in Tissue-Equivalent Material Immersed in a Radioactive Cloud," Health Physics, Vol. 26, pp. 1-12, January 1974.

4. Murphy, K.G. and Campe, K.M., "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19," Paper presented at the 13th AEC Air Cleaning Conference.

5. "Meteorology and Atomic Energy 1968," D. H. Slade (ed.), USAEC Report, TID 24190, 1968.

Insert 15A.6 15.A-14 Rev. OL-16

ULNRC-06636 Page 195 of 374 TABLE 15A-1 PARAMETERS USED IN ACCIDENT ANALYSIS I. General

1. Core power level, Mwt 3636 (102% power)

2. Number of fuel assemblies in the core 193

3. Maximum radial peaking factor 1.65

4. Percentage of failed fuel 1.0

5. Steam generator tube leak, lb/hr 500 II. Sources

1. Core inventories, Ci Table 15A-3

2. Gap inventories, Ci Table 15A-3 3.

2. Primary coolant specific activities, µCi/gm Table 15A-5

4. Primary coolant activity, technical specification limit for iodines - I-131 dose equivalent, µCi/gm 1.0

5. Secondary coolant activity technical specification limit for iodines - I-131 dose equivalent, µCi/gm 0.1 III. Activity Release Parameters 2.70

1. Free volume of containment, ft3 2.5 x 106

2. Containment leak rate

i. 0-24 hours, % per day 0.2 ii. after 24 hrs, % per day 0.1 IV. Control Room Dose Analysis (for LOCA)

1. Control building

i. Mixing volume, cf 148,000 150,000 ii. Filtered intake, cfm Prior to operator action (0-30 minutes) 900 After operator action (30 minutes - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />) 450 iii. Unfiltered inleakage, cfm **

iv. Filter efficiency (all forms of iodine), % 95

2. Control room

i. Volume, cf particulates 48,500 100,000 ii. Filtered flow from control building, cfm 440

ULNRC-06636 Page 196 of 374 TABLE 15A-1 (Sheet 2) iii. Unfiltered flow from control building, cfm Prior to operator action (0-30 minutes) 440 After operator action (30 minutes - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />) 0 iv. Filtered recirculation, cfm 1030 1360

v. Filter efficiency (all forms of iodine), % 95 vi. Unfiltered in leakage, cfm **

V. Miscellaneous TEDE

1. Atmospheric dispersion factors, /Q sec/m3 Table 15A-2

2. Dose conversion factors

i. total body and beta skin, rem-meter3/Ci-sec (Sv-meter3/Bq-sec) Table 15A-4 ii. thyroid, rem/Ci (Sv/Bq) Table 15A-4

3. Breathing rates, meter3/sec

i. control room at all times 55 x 10-4 3.47 ii. offsite 0-8 hrs 5 x 10-4 3.47

-4 8-24 hrs 1.75 8 x 10 24-720 hrs 3 x 10-4 2.32

4. Control room occupancy fractions 0-24 hrs 1.0 24-96 hrs 0.6 96-720 hrs 0.4

• • See Figure 15A-2 for inleakage values used in the accident analysis.

This Figure is replaced with Insert F15A-2

ULNRC-06636 Page 197 of 374 TABLE 15A-2 LIMITING SHORT-TERM ATMOSPHERIC DISPERSION FACTORS

(/QS) FOR ACCIDENT ANALYSIS Location Type/ /Q Replace with Insert T15A-2 Time Interval (hrs) (sec/meter3)

Site boundary 0-2 1.5E-4 Low-population zone 0-8 1.5E-5 8-24 1.0E-5 24-96 4.6E-6 96-720 1.5E-6 Control room (via containment leakage) 0-8 7.2E-4 8-24 5.3E-4 24-96 1.7E-4 96-720 0 Control room (via unit vent exhaust) 0-8 1.3E-4 8-24 9.0E-5 24-96 4.1E-5 96-720 0

ULNRC-06636 Page 198 of 374 TABLE 15A-3 FUEL AND ROD GAP INVENTORIES - CORE (CI)

Core Isotope Gap Replace entire table with insert T15A-3 I-131 9.95E+6 I-132 1.44E+8 1.44E+7 I-133 2.04E+8 2.04E+7 I-134 2.25E+8 2.25E+7 I-135 1.91E+8 1.91E+7 Kr-83m 1.27E+7 1.27E+6 Kr-85m 2.72E+7 2.72E+6 Kr-85 8.61E+5 2.58E+5 Kr-87 5.24E+7 5.24E+6 Kr-88 7.38E+7 7.38E+6 Kr-89 9.03E+7 9.03E+6 Xe-131m 1.12E+6 1.12E+5 Xe-133m 6.35E+6 6.35E+5 Xe-133 1.99E+8 1.99E+7 Xe-135m 3.96E+7 3.96E+6 Xe-135 4.38E+7 4.38E+6 Xe-137 1.78E+8 1.78E+7 Xe-138 1.70E+8 1.70E+7

• Gap activity is assumed to be 10 percent of fuel activity for all isotopes except for Kr-85; for Kr-85 it is assumed to be 30 percent of the fuel activity.

ULNRC-06636 Page 199 of 374 TABLE 15A-4 DOSE CONVERSION FACTORS USED IN ACCIDENT ANALYSIS Total Body Beta Skin Rem-meter3 Rem-meter3 Thyroid

• Nuclide Ci-sec Ci-sec Rem/Ci I-131 8.72E-2 3.17E-2 1.49E+6 I-132 5.13E-1 1.32E-1 1.43E+4 I-133 1.55E-1 7.35E-2 2.69E+5 I-134 5.32E-1 9.23E-2 3.73E+3 I-135 4.21E-1 1.29E-1 5.60E+4 Kr-83m 2.40E-6 0 NA Kr-85m 3.71E-2 4.63E-2 NA Kr-85 5.11E-4 4.25E-2 NA Kr-87 1.88E-1 3.09E-1 NA Kr-88 4.67E-1 7.52E-2 NA Kr-89 5.27E-1 3.20E-1 NA Xe-131m 2.91E-3 1.51E-2 NA Xe-133m 7.97E-3 3.15E-2 NA Xe-133 9.33E-3 9.70E-3 NA Xe-135m 9.91E-2 2.25E-2 NA Xe-135 5.75E-2 5.90E-2 NA Xe-137 4.51E-2 3.87E-1 NA Xe-138 2.80E-1 1.31E-1 NA

• The radiological consequences for the replacement SG program (1) have been re-analyzed using the following thyroid dose conversion factors from ICRP-30 and whole body dose conversion factors from Federal Guidance Report 12 (except that RG 1.109 Table B-1 is used for Kr-89 and Xe-137). These factors may be applied to other accident sequences as they are re-analyzed (e.g., the fuel handling accident cases addressed in Section 15.7.4): Replace with Insert T15A-4 Total Body

• • REM-meter3 Nuclide Ci-sec Thyroid Rem/ci I-131 6.73E-02 1.07E+06 I-132 4.14E-01 6.29E+03 I-133 1.09E-01 1.81E+05 I-134 4.81E-01 1.07E+03 I-135 2.95E-01 3.14E+04 Kr-83m 5.55E-06 NA Kr-85m 2.77E-02 NA

ULNRC-06636 Page 200 of 374 TABLE 15A-4 (Sheet 2)

Total Body

• • REM-meter3 Nuclide Ci-sec Delete Thyroid Rem/ci Kr-85 4.40E-04 NA Kr-87 1.52E-01 NA Kr-88 3.77E-01 NA Kr-89 5.27E-01 NA Xe-131m 1.44E-03 NA Xe-133m 5.07E-03 NA Xe-133 5.77E-03 NA Xe-135m 7.55E-02 NA Xe-135 4.40E-02 NA Xe-137 4.51E-02 NA Xe-138 2.14E-01 NA 3

Sv - meter

• • Federal Guidance Report 12 uses units of ------------------------------- .

Bq - sec Conversion factors are: 1 Sv = 100 Rem and 1 Bq = 2.7E-11 Ci. The above WB dose conversion factors are equal to those in Federal Guidance Report 12.

(1) FSAR sections re-analyzed for radiological consequences as part of the replacement steam generator program include:

15.1.5 STEAM SYSTEM PIPING FAILURE 15.2.6 LOSS OF NONEMERGENCY AC POWER TO THE STATION AUXILIARIES 15.3.3 REACTOR COOLANT PUMP SHAFT SEIZURE (LOCKED ROTOR) 15.4.8 SPECTRUM OF ROD CLUSTER CONTROL ASSEMBLY EJECTION ACCIDENTS 15.6.2 BREAK IN INSTUMENT LINE OR OTHER LINES FROM REACTOR COOLANT PRESSURE BOUNDARY THAT PENETRATE CONTAINMENT 15.6.3 STEAM GENERATOR TUBE FAILURE

ULNRC-06636 Page 201 of 374 TABLE 15A-5 INITIAL RADIOACTIVITY FOR ACCIDENTS THAT USE THE PRIMARY-TO-SECONDARY LEAKAGE RELEASE PATHWAY I. Reactor Coolant System Inventories Concentration Isotope (µCi/gm)

a. Iodines 1 I-131 0.793 I-132 I-130 1.75E-02 2.2 I-133 1.12 I-134 4.0 I-135 2.2

b. Noble Gases 2 Kr-83m 2.02E-01 Kr-85m 1.00E+00 Kr-85 7.45E-02 Kr-87 5.86E-01 Kr-88 1.88E+00 Kr-89 5.04E-02 Xe-131m 1.77E-01 Xe-133m 9.64E-01 Xe-133 4.81E+01 Xe-135m 1.31E-01 Xe-135 2.87E+00 Xe-137 9.06E-02 Xe-138 4.40E-01 Insert T15A-5 (additional pages)

ULNRC-06636 Page 202 of 374 TABLE 15A-5 (Sheet 2)

II. Boron Recycle Holdup Tank Inventories 3 4 Concentration Isotope (Ci)

a. Iodines I-131 4.07 I-132 0.044 I-133 0.740 I-134 3.81E-3 I-135 0.119

b. Noble Gases Kr-83m 0.169 Kr-85m 1.92 Kr-85 11.53 Kr-87 0.330 Kr-88 2.34 Kr-89 1.18E-03 Xe-131m 15.91 Xe-133m 23.26 Xe-133 2560 Xe-135m 0.0353 Xe-135 11.83 Xe-138 0.0463

ULNRC-06636 Page 203 of 374 5

TABLE 15A-5 (Sheet 3)

III. Boron Recycle Holdup Tank Inventories 3 Concentration Isotope (Ci)

a. Iodines I-131 26.4 I-132 0.533 I-133 3.40 I-134 4.60E-4 I-135 0.377 Notes:

1. The RCS iodine values were obtained by starting with the original Licensing Bases 1% failed fuel projections. Then the shorter-lived iodine isotopic concentrations were increased based on steady-state conditions observed during fuel cycles in which Callaway operated with failed fuel.

This isotopic spectrum is intended to bound concentrations that would be encountered with either tight or open fuel defects.

2. The RCS noble gas values were obtained based on the original Licensing Bases 1% failed fuel projections, and then adjusted upwards to account for calorimetric error and capacity factor variations.

3. Radwaste Tank inventories are based on the original Licensing Bases projections and adjusted for capacity factor and plant power uprating.

, Class 3 and Class 6,

ULNRC-06636 Page 204 of 374 Attachment B - SGTR Figures

ULNRC-06636 Page 205 of 374

ULNRC-06636 Enclosure 6 Page 206 of 374 SGTR_FSAR_Plots.xlsx Tab A Pressurizer Ruptured Steam Generator Intact Steam Generator (Loop 2) 3000 PRESSURE (PSIA) 2000 1000 0

0 1000 2000 3000 4000 5000 Time (s)

ULNRC-06636 Page 207 of 374

ULNRC-06636 Enclosure 6 Page 208 of 374 SGTR_FSAR_Plots.xlsx Tab B Ruptured Loop Hot Leg Ruptured Loop Cold Leg 700 600 TEMPERATURE (DEG F) 500 400 300 200 100 0 1000 2000 3000 4000 5000 Time (s)

ULNRC-06636 Page 209 of 374

ULNRC-06636 Enclosure 6 Page 210 of 374 SGTR_FSAR_Plots.xlsx Tab C Intact Loop Hot Leg (Loop 2)

Intact Loop Cold Leg (Loop 2) 700 600 TEMPERATURE (DEG F) 500 400 300 200 100 0 1000 2000 3000 4000 5000 Time (s)

ULNRC-06636 Page 211 of 374

ULNRC-06636 Enclosure 6 Page 212 of 374 SGTR_FSAR_Plots.xlsx Tab D Intact Steam Generators (Total) 4000 3000 PRESSURE (PSIA) 2000 1000 0

0 1000 2000 3000 4000 5000 Time (s)

ULNRC-06636 Page 213 of 374

ULNRC-06636 Enclosure 6 Page 214 of 374 SGTR_FSAR_Plots.xlsx Tab E Ruptured Steam Generator 1200 1000 STEAM FLOW RATE (LBM/SEC) 800 600 400 200 0

0 1000 2000 3000 4000 5000 Time (s)

ULNRC-06636 Page 215 of 374

ULNRC-06636 Enclosure 6 Page 216 of 374 SGTR_FSAR_Plots.xlsx Tab F Ruptured Steam Generator Intact Steam Generator (Loop 2) 700 600 TEMPERATURE (DEG F) 500 400 300 0 1000 2000 3000 4000 5000 Time (s)

ULNRC-06636 Page 217 of 374

ULNRC-06636 Enclosure 6 Page 218 of 374 SGTR_FSAR_Plots.xlsx Tab G Ruptured Steam Generator 500 400 ARV/SV FLOW RATE (LBM/SEC) 300 200 100 0

0 1000 2000 3000 4000 5000 Time (s)

ULNRC-06636 Page 219 of 374

ULNRC-06636 Enclosure 6 Page 220 of 374 SGTR_FSAR_Plots.xlsx Tab H Intact Steam Generators (Total) 1200 1000 ARV/SV FLOW RATE (LBM/SEC) 800 600 400 200 0

0 1000 2000 3000 4000 5000 Time (s)

ULNRC-06636 Page 221 of 374

ULNRC-06636 Enclosure 6 Page 222 of 374 SGTR_FSAR_Plots.xlsx Tab I Ruptured Steam Generator 60 50 BREAK FLOW RATE (lbm/s) 40 30 20 10 0

0 1000 2000 3000 4000 5000 Time (s)

ULNRC-06636 Page 223 of 374

ULNRC-06636 Enclosure 6 Page 224 of 374 SGTR_FSAR_Plots.xlsx Tab J Ruptured Steam Generator AFW Flow Ruptured Steam Generator NR Level 100 90 80 AFW FLOW (LBM/SEC) OR NR LEVEL (%)

70 60 50 40 30 20 10 0

0 1000 2000 3000 4000 5000 Time (s)

ULNRC-06636 Page 225 of 374

ULNRC-06636 Enclosure 6 Page 226 of 374 SGTR_FSAR_Plots.xlsx Tab K Intact Steam Generators AFW Flow (Total)

Intact Steam Generator NR Level (Loop 2) 110 100 90 AFW FLOW (LBM/SEC) OR NR LEVEL (%)

80 70 60 50 40 30 20 10 0

0 1000 2000 3000 4000 5000 Time (s)

ULNRC-06636 Page 227 of 374

ULNRC-06636 Enclosure 6 Page 228 of 374 SGTR_FSAR_Plots.xlsx Tab L Ruptured Steam Generator 5000 4000 LIQUID VOLUME (FT3) 3000 2000 1000 0 1000 2000 3000 4000 5000 Time (s)

ULNRC-06636 Page 229 of 374

ULNRC-06636 Enclosure 6 Page 230 of 374 SGTR_FSAR_Plots.xlsx Tab M PORV Flow Rate 90 80 70 PORV FLOW RATE (LBM/SEC) 60 50 40 30 20 10 0

0 1000 2000 3000 4000 5000 Time (s)

ULNRC-06636 Page 231 of 374

ULNRC-06636 Enclosure 6 Page 232 of 374 SGTR_FSAR_Plots.xlsx Tab N Pressurizer Liquid Volume 1200 1100 1000 900 LIQUID VOLUME (FT3) 800 700 600 500 400 300 200 100 0

0 1000 2000 3000 4000 5000 Time (s)

ULNRC-06636 Page 233 of 374

ULNRC-06636 Enclosure 6 Page 234 of 374 SGTR_FSAR_Plots.xlsx Tab O Ruptured Steam Generator 1200 1100 1000 900 FW FLOW RATE (LBM/SEC) 800 700 600 500 400 300 200 100 0

0 1000 2000 3000 4000 5000 Time (s)

ULNRC-06636 Page 235 of 374

ULNRC-06636 Enclosure 6 Page 236 of 374 SGTR_FSAR_Plots.xlsx Tab P Intact Steam Generators (Total) 4000 3000 FW FLOW RATE (LBM/SEC) 2000 1000 0

0 1000 2000 3000 4000 5000 Time (s)

ULNRC-06636 Page 237 of 374

ULNRC-06636 Enclosure 6 Page 238 of 374 Overfill_FSAR_Plots.xlsx Tab A Pressurizer Intact SG Ruptured SG 2400 2200 2000 1800 PRESSURE (PSIA) 1600 1400 1200 1000 800 600 400 200 0

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 Time (s)

ULNRC-06636 Page 239 of 374

ULNRC-06636 Enclosure 6 Page 240 of 374 Overfill_FSAR_Plots.xlsx Tab B Hot Leg Cold Leg 700 600 500 Temp (DEG F) 400 300 200 100 0

Time (s)

ULNRC-06636 Page 241 of 374

ULNRC-06636 Enclosure 6 Page 242 of 374 Overfill_FSAR_Plots.xlsx Tab C Hot Leg (Loop 2)

Cold Leg (Loop 2) 700 600 500 Temp (DEG F) 400 300 200 100 0

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 Time (s)

ULNRC-06636 Page 243 of 374

ULNRC-06636 Enclosure 6 Page 244 of 374 Overfill_FSAR_Plots.xlsx Tab D RCS Ruptured SG Intact SGs 700000 600000 500000 MASS (LBM) 400000 300000 200000 100000 0

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 Time (s)

ULNRC-06636 Page 245 of 374

ULNRC-06636 Enclosure 6 Page 246 of 374 Overfill_FSAR_Plots.xlsx Tab E Break Flow Flashing Fraction 0.15 FLASHING FRACTION 0.10 0.05 0.00 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 Time (s)

ULNRC-06636 Page 247 of 374

ULNRC-06636 Enclosure 6 Page 248 of 374 Overfill_FSAR_Plots.xlsx Tab F Ruptured Steam Generator Intact Steam Generator (Loop 2) 700 600 500 TEMPERATURE (DEG F) 400 300 200 100 0

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 Time (s)

ULNRC-06636 Page 249 of 374

ULNRC-06636 Enclosure 6 Page 250 of 374 Ruptured SG SV 260 240 220 200 FLOW RATE (LBM/S) 180 160 140 120 100 80 60 40 20 0

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 Time (s)

ULNRC-06636 Page 251 of 374

ULNRC-06636 Enclosure 6 Page 252 of 374 Intact ARV Flow 500 450 400 FLOW RATE (LBM/S) 350 300 250 200 150 100 50 0

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 Time (s)

ULNRC-06636 Page 253 of 374

ULNRC-06636 Enclosure 6 Page 254 of 374 RCS to Ruptured SG 70 65 60 55 BREAK FLOW (LBM/S) 50 45 40 35 30 25 20 15 10 5

0 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 Time (s)

ULNRC-06636 Page 255 of 374

ULNRC-06636 Enclosure 6 Page 256 of 374 Overfill_FSAR_Plots.xlsx Tab J Ruptured SG AFW Ruptured SG Narrow Range 300 275 250 225 200 FLOW RATE (LBM/S) / NR (%)

175 150 125 100 75 50 25 0

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 Time (s)

ULNRC-06636 Page 257 of 374

ULNRC-06636 Enclosure 6 Page 258 of 374 Overfill_FSAR_Plots.xlsx Tab K Intact SG AFW Intact SG Narrow Range (Loop 2) 300 275 250 225 200 FLOW RATE (LBM/S) / NR (%)

175 150 125 100 75 50 25 0

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 Time (s)

ULNRC-06636 Page 259 of 374

ULNRC-06636 Enclosure 6 Page 260 of 374 Overfill_FSAR_Plots.xlsx Tab L Ruptured SG + Steam Line SG + Steam Line SG + 100 7000 6000 5000 TOTAL LIQUID VOLUME (FT3) 4000 3000 2000 1000 0

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 Time (s)

ULNRC-06636 Page 261 of 374

ULNRC-06636 Enclosure 6 Page 262 of 374 Overfill_FSAR_Plots.xlsx Tab M Pressurizer PORV 100 90 80 FLOW RATE (LBM/S) 70 60 50 40 30 20 10 0

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 Time (s)

ULNRC-06636 Page 263 of 374

ULNRC-06636 Enclosure 6 Page 264 of 374 Indicated Level 100 90 80 LEVEL (% RANGE) 70 60 50 40 30 20 10 0

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 Time (s)

ULNRC-06636 Page 265 of 374 Attachment C - FSAR Addendum Markups

ULNRC-06636 Page 266 of 374 TABLE OF CONTENTS (Continued)

Section Page 2.

2.2 DESCRIPTION

S.................................................................................................. 2.2-8 2.2.2.1 Hydrogen System .......................................................................................... 2.2-8 2.2.3 EVALUATION OF POTENTIAL ACCIDENTS .............................................. 2.2-10 2.2.3.1 Design Basis Events ................................................................................... 2.2-10 2.3.4.4 Alternative Source Term Short-Term 2.3 METEOROLOGY..........................................................

Diffusion Estimates 2.3-1 2.3.1 REGIONAL CLIMATOLOGY ............................................................................. 2.3-1 2.3.1.1 General Climate............................................................................................. 2.3-1 2.3.1.2 Regional Meterological Conditions for Design and Operating Bases... 2.3-2 2.3.2 LOCAL METEOROLOGY ................................................................................ 2.3-19 2.3.2.1 Normal and Extreme Values of Meteorological ...................................... 2.3-20 2.3.2.2 Potential Influence of the Plant and Its Facilities on Local Meteorology. ................................................................................................ 2.3-28 2.3.3 ON-SITE METEOROLOGICAL MEASUREMENT PROGRAMS .............. 2.3-48 2.3.3.1 Preoperational and Operational Monitoring Programs .......................... 2.3-48 2.3.3.2 Representativeness of the Data Base ..................................................... 2.3-58 2.3.4 SHORT-TERM DIFFUSION ESTIMATES .................................................... 2.3-59 2.3.4.1 Objective ....................................................................................................... 2.3-59 2.3.4.2 Calculations .................................................................................................. 2.3-60 2.3.4.3 Data Representativeness ........................................................................... 2.3-63 2.3.5 LONG-TERM DIFFUSION ESTIMATES ....................................................... 2.3-64 2.3.5.1 Objective ....................................................................................................... 2.3-64 2.3.5.2 Calculations .................................................................................................. 2.3-64 2.4 HYDROLOGIC ENGINEERING.............................................................................. 2.4-1 2.4.1 HYDROLOGIC DESCRIPTION ........................................................................ 2.4-1 2.4.1.1 Site and Facilities .......................................................................................... 2.4-1 2.0-ii

ULNRC-06636 Page 267 of 374 LIST OF TABLES (Continued)

Insert AddendumChapter2LOT Number Tit 2.3-81 Average Meteorological Relative Concentration Analysis Standard Distances, Radwaste Building Vent Release 2.3-82 Average Meteorological Relative Concentration Analysis Special Distances, Unit Vent Release 2.3-83 Average Meteorological Relative Concentration Analysis Standard Distances, Unit Vent Release Data Period: May 4, 1973 to May 4, 1975 and March 16, 1978 to March 16, 1979 Combined 2.3-84 Average Meteorological Relative Concentration analysis Special Distances, Radwaste Building Vent Release 2.3-85 Atmospheric Relative Concentrations Unit Vent Release Grazing Season 2.3-86 Atmospheric Relative Concentrations Radwaste Building Release Grazing Season 2.4-1 Four Tributary Stream Systems of the Missouri River Near the Callaway Plant Site 2.4-2 Missouri River Drainage Basin 2.4-3 Missouri River and Alluvium Water Supplies 2.4-4 Missouri River Discharges 2.4-5 Major Recorded Floods at Hermann, Missouri 2.4-6 Estimated Magnitude and Frequency of Floods in Missouri River for Existing Conditions 2.4-7 Probable Maximum Precipitation (PMP) at the Site 2.4-8 Hourly Distribution of Maximum 6-Hour Increment Within 48-Hour PMP Storm 2.4-9 Rainfall Intensities at Callaway Plant Site for 100-Year Storm and Probable Maximum Precipiation Storm 2.4-10 Estimated Magnitude and Frequency of Consecutive Annual Low Flows for Various Durations in the Missouri River

ULNRC-06636 Page 268 of 374 Columbia annual mean and the 3-year annual mean site dew-point measurements are identical.

Monthly variation in wind direction amounted to no more than three 22.5-degree sectors, and the annual means of the two data sources (Columbia and on site) were within one 22.5degree sector. Mean monthly wind speed was as much as 1.7 m/sec lower on site than at Columbia (during the month of February) and was an average of 1.2 m/sec lower on site on an annual basis. Since the tendency toward significantly lower wind speed measurements on meteorological towers using stateof-the-art instrumentation compared with airport measurements has been noted in several cases, the disparity between the measurements may be attributed to difference in instrument accuracy rather than actual wind speed differences. On-site data were measured at 10 meters, while the anemometer height at Columbia was 6 meters. Whatever reason for the disparity, the lower speeds measured on site are conservative with respect to dispersion calculations.

The parameter of paramount importance other than wind speed and direction to dispersion calculations, atmospheric stability, is not routinely measured by the NWS. The NWS STAR computer program approximates stability measurements by computing Pasquill stability classes on the basis of cloudiness, sun angle, and time of day. This approximation of long-term regional stability, based on Columbia, Missouri, data, 1960 through 1969, is compared with stability measured on site in Table 2.3-55. It is apparent that the on-site data provide a somewhat greater frequency of stable conditions than does the STAR approximation. The difference is probably due to the crudeness of the STAR method Sections 2.3.4.1, of calculation. 2.3.4.2 Again, the and 2.3.4.3 on-site datadescribe historicalcompared are conservative calculations to of the Columbia datathe withshort-term respect todiffusion dispersion calculation.

estimates. Chapter 15 dose consequences for accidents, as described in Sections 15A.1 through 15A.4, were re-Annual joint f requency analyzeddistributions using the Alternative (JFDs) of Source Term (AST) wind direction, wind analysis.

speed, and atmospheric stability for the 10- and 60-meter wind levels and 60-10 meter T (or 90-10 meter T when 60-10 meter are missing) for the data periods, May 4, 1973 to May 4, 1974 and May 4, 1974 to May 4, 1975 are provided in Tables 2.3-56 and 2.3-57, respectively. Annual JFDs at 10, 60, and 90 meters for the period March 16, 1978 to March 16, 1979 are provided in Table 2.3-58. Table 2.3-59 provides annual JFDs at 10 and 60 meters for the three data periods combined. Monthly JFDs, at 10 and 60 meters, for the three data periods combined are provided in Table 2.3-60.

2.3.4 SHORT-TERM DIFFUSION ESTIMATES 2.3.4.1 Objective Conservative and realistic estimates of atmospheric diffusion /Q at the site boundary (exclusion area) and the outer boundary of the LPZ were performed for time periods up to 30 days after an accident. Diffusion evaluations for short-term accidents are based on the assumption of release points or areas which are effectively lower than 2-1/2 times the height of adjacent solid structures. Description of models used and assumptions made are discussed in section 2.3.4.2.2.

ULNRC-06636 Page 269 of 374 on-site wind direction persistence were discussed in Sections 2.3.2.2.2.4 and 2.3.2.2.2.5, respectively. It is concluded that on-site and regional persistence are similar.

The topography in the vicinity of the site is similar to that in the vicinity of Columbia. Low rolling hills without significant relief occur in both ares, as shown in Figure 2.3-12.

A direct comparison of diffusion estimates based on the on-site data and the long-term (Columbia, Missouri) data would be quite meaningless, because the long-term data do not contain measurements of vertical temperature difference or wind direction variability.

In addition, long-term wind speed data are based on anemometer starting thresholds of approximately 2 to 2.5 mph versus starting thresholds of 0.75 mph for the on-site anemometers. The Pasquill-Turner approximation, used to obtain stability classification for long-term meteorology data based on sun angle, cloudiness, and time of day (described in Section 2.3.2 and in Table 2.3-31), is too crude to yield stability values comparable to those based on vertical temperature difference and low-threshold wind speed measurements for determination of stability classification for on-site meteorology data.

2.3.5 LONG-TERM DIFFUSION ESTIMATES Insert Addendum2.3.4.4 2.3.5.1 Objective The objective of Section 2.3.5 is to provide realistic long-term diffusion estimates at distances up to 80 km (50 miles) from the plant for annual average release limit calculations and man-rem estimates. The terrain within 80 km (50 miles) of the site is gently rolling; no important ranges of hills or mountains are within the region. There are several small lakes and reservoirs in the region; however, no substantial water bodies are present, which are large enough to affect ambient dispersion parameters.

The analyses were based on on-site meteorological data over the periods, May 4, 1973 to May 4, 1975 and March 16, 1978 to March 16, 1979.

2.3.5.2 Calculations Both the variable trajectory plume segment atmospheric transport model, MESODIF-II (NUREG/CR-0523), and the straight-line Gaussian dispersion model, XOQDOQ (NUREG/CR-2919), were used to determine for the long-term (annual average) diffusion estimates.

2.3.5.2.1 Plume Segment Atmospheric Transport Model (MESODIF-II)

MESODIF-II is a variable trajectory plume segment atmospheric transport model. It is designed to predict relative atmospheric dispersion factors, /Q and deposition factors, D/Q, of radioactive, but otherwise non-reactive material. In such a model, calculated

ULNRC-06636 Enclosure 6 Page 270 of 374 REFERENCES - SECTION 2.3 (Continued)

Dames & Moore, 1980, Final safety analysis report, Wolf Creek, for Kansas Gas &

Electric Co.

Dunn, W., Boughton, B., and Policastro, A., 1978, Evaluation of droplet evaporation formulations employed in drift deposition models. Proceedings of Symposium on Environmental Effects Hanna, S.R., of Cooling 1978, A simple Tower Emissions,model drift deposition University of Maryland applied to the (May 2-4). chalk point dye tracer experiment. Proceedings of Symposium on Environmental Egan, B.A., 1975, Turbulent Effects diffusionofinCooling complex Tower Emissions, terrain, University in Lectures of on air pollution and Maryland (May 2-4).

environmental impact analyses - AMS workshop on meteorology and environmental assessment, Hagen, D., Amer. Meteor. Soc., Boston, Massachusetts, p. 123-124.

Englemann, R.J., 1968, The calculation of precipitation scavenging, in Meteorology and atomic energy. U.S. Dept. of Commerce, Springfield, Virginia, p. 208-220.

Environmental Data Service, 1968a, Climatological data -national summary. U.S. Dept.

of Commerce, Asheville, North Carolina, ESSA.

Environmental Systems Corporation, 1977, Cooling tower drift dye tracer experiment.

PPSP-CPCTP, Prepared for Maryland Power Plant Siting Program by ESC, Knoxville, Tennessee, 103 p.

Frost, R., 1948, Quart. J. Roy. Met. Soc., vol. 74.

Gifford, F.A., Jr., 1969, Use of routine meteorological observations for estimating atmospheric dispersion: Nuclear Safety, vol. 2, no. 4 (June). Hanna, S.R., 1978, A simple drift deposition model applied to the chalk point dye tracer experiment.

Proceedings of Symposium on Environmental Effects of Cooling Tower Emissions, University of Maryland (May 2-4).

Hanna, S.R., and Gifford, F.A., 1975, Meteorological effects of energy dissipation at large power parks, National Oceanographic and Atmospheric Administration, Oak Ri, Tennessee (March).

Hershfield, D.M., 1961, Rainfall frequency atlas of the United States for durations from 30 minutes to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and return periods from 1 to 100 years. U.S. Government Printing Office.

Holzworth, 1972, Mixing heights, wind speeds and potential for urban air pollution throughout the contiguous United States. United States EPA, no. AP101.

Hosler, C.R., 1961, Low-level inversion frequency in the contiguous United States.

Monthly Weather Review, 89, 379-339.

Harvey, R.B, LaVie, S., and Brown, L, 2003, Using ARCON96 for Control Room Radiological Habitability Assessments. Proceedings of the Ninth Nuclear Utility Meteorological Data Users Group (NUMUG) Meeting, Chattanooga, Tennesse (October 8-10)

ULNRC-06636 Page 271 of 374 REFERENCES - SECTION 2.3 (Continued)

Jumod, A., et al., 1975, Meteorological differences on atmospheric cooling systems as projected in Switzerland. ERDA Symposium series, CONF-740302, Cooling Tower Environment (1974).

Leason, D.B., 1974, Planning aspects of cooling towers. Atmospheric Environment, vol.

8, p. 307-312.

Markee and Beckerly, 1974, Technical basis for interim regional tornado criteria. U.S.

Atomic Energy Commission, Washington, D.C.

Marshall, J.L., 1973, Lightning protection: John Wiley and Sons.

Meyer, J.H., and Jenkins, W.R., 1977, Cooling tower drift dye tracer experiment, June 16 and 17, surface weather and ambient atmospheric profile data. Prepared for Maryland Power Plant Siting Program by the Johns Hopkins University Applied Physics Laboratory, Laurel, Maryland, PPSP-CPTP-16, vol. 3, 102 p.

Meyer, J.H., and Stanbro, W.D., 1977a, Cooling tower drift dye tracer experiment, June 16 and 17. Prepared for Maryland Power Plant Siting Program by the Johns Hopkins University Applied Physics, Laurel, Maryland, PPSP-CPTP-16, vol. 2, 102 p.

, 1977b, Fluorescent dye, a novel technique to trace cooling tower drift.

Presented at 4th Joint Conference on Sensing Environmental Pollutants, November 6-11, 1977, New Orleans, Johns Hopkins University, Applied Physics Laboratory, Laurel, Maryland, 6 p.

National Ramsdell, Climatic J.V. Jr, Simonen, Center, C.A., and of no date, Summary Smyth, hourlyS.B, 1995, Atmospheric observations, Columbia, Relative Missouri,Concentrations (1960-69). in Building Wakes. NUREG/CR-6331. Prepared for the U.S.N.R.C by Pacific Northwest Laboratory.

National Weather Service, 1945-1969, Three hourly data, Columbia, Missouri. National Oceanographic and Atmospheric Administration.

Pautz, M.E., 1969, Severe local storm occurrences 1955-1967. Office of Meteorological Operations, Silver Spring, Maryland, ESSA Technical Memo, WBTM FCST 12.

Poultney, N.E., 1973, The tornado season of 1972. Weatherwise, 26, 22-27.

Ryan, P.J., and Harleman, D.R.F., 1973, An analytical and experimental study of transient cooling pond behavior. Ralph H. Parsons Laboratory, Department of Civil Engineering, MIT, Report no. 161 (January).

Sagendorf, J.F., 1974, A program for evaluating atmospheric dispersion from a nuclear power station. National Oceanic and Atmospheric Administration Technical Memorandum ERL-ARL-42.

Slade, D.H., 1968, Meteorology and atomic energy, National Technical Information Service, U.S. Department of Commerce, Springfield, Virginia.

ULNRC-06636 Page 272 of 374 APPENDIX 3.A - CONFORMANCE TO NRC REGULATORY GUIDES This appendix briefly discusses the extent to which Union Electric conforms to NRC published regulatory guides for the site related portions of Callaway Plant. The Standard Plant FSAR Appendix 3A may refer to the Addendum Appendix 3A or the Union Electric Operational Quality Assurance Manual (OQAM) for the specific regulatory commitment for certain regulatory guides. However in cases where a reference is not made to the Addendum Appendix 3A or the OQAM, the commitment is as stated in the Standard Plant Appendix 3A and the same regulatory position is not repeated in the Addendum Appendix 3A or in the OQAM. The statement of specific regulatory commitment for the following regulatory guides is located as indicated:

Callaway FSAR, Standard Plant - Regulatory Guides 1.1, 1.2, 1.3, 1.4, 1.5, 1.6, 1.7, 1.9, 1.10, 1.11, 1.12, 1.13, 1.14, 1.15, 1.18, 1.20, 1.22, 1.24, 1.25, 1.26, 1.29, 1.31, 1.32, 1.34, 1.35, 1.36, 1.40, 1.41, 1.42, 1.43, 1.44, 1.45, 1.46, 1.47, 1.48, 1.49, 1.50, 1.51, 1.52, 1.53, 1.54, 1.55, 1.56, 1.57, 1.59, 1.60, 1.61, 1.62, 1.63, 1.65, 1.66, 1.67, 1.68, 1.68.1, 1.68.2, 1.69, 1.70, 1.71, 1.72, 1.73, 1.75, 1.76, 1.77, 1.78, 1.79, 1.80, 1.81, 1.82, 1.83, 1.84, 1.85, 1.87, 1.89, 1.90, 1.92, 1.93, 1.95, 1.96, 1.97, 1.98, 1.99, 1.100, 1.101, 1.102*, 1.103, 1.104, 1.105, 1.106, 1.107, 1.108, 1.110, 1.112, 1.115, 1.117, 1.118, 1.119, 1.120, 1.121, 1.122, 1.124, 1.126, 1.128, 1.129, 1.130, 1.131, 1.133, 1.136, 1.137, 1.139, 1.140, 1.141, 1.142, 1.143, 1.147, 1.150, 1.152, 1.155, 1.158, 1.160, 1.163, 1.181, 1.182, 1.187, and 1.195.

1.183, and 1.194 Callaway FSAR, Site Addendum - Regulatory Guides 1.17, 1.21, 1.23, 1.27, 1.59, 1.86, 1.91, 1.102*, 1.109, 1.111, 1.113, 1.114, 1.125, 1.127, 1.132, 1.134, 1.138, and 1.145.

Union Electric Operational Quality Assurance Manual - Regulatory Guides 1.8, 1.28, 1.30, 1.33, 1.37, 1.38, 1.39, 1.58, 1.64, 1.74, 1.88, 1.94, 1.116, 1.123, 1.144, and 1.146.

Clarifications, alternatives, and exceptions to these guides are identified and justification is presented or referenced. In the discussion of each guide, the sections or tables of the FSAR where more detailed information is presented are referenced. The referenced tables provide a comparison of Union Electric's position to each regulatory position of section C of the regulatory guides. All statements within the Regulatory Position Section (C) of the Regulatory Guides are considered requirements unless a specific exception or clarification has been committed to by Union Electric. This is true regardless of the qualifier (i.e., "shall" or "should") which prefaces the statement. As regards to standards endorsed by the Regulatory Guide, unless further qualified within the Regulatory Guide, "shall" statements denote requirements while "should" statements denote recommendations. A glossary of definitions is provided in the Quality Assurance Procedures Manual.

• Refer to both the Standard Plant and the Site Addendum for the Complete statement of regulatory commitment.

ULNRC-06636 Enclosure 6 Page 273 of 374 REGULATORY GUIDE 1.144 REGULATORY GUIDE 1.145, Revision 1, DATED 11/82 Auditing of Quality Atmospheric Assurance Dispersion ModelsPrograms for Nuclear for Potential AccidentPower Plants Assessments at Consequence Nuclear Power Plant DISCUSSION:

DISCUSSION The recommendations of this regulatory guide are met as described in Site Addendum Refer to2.3.4.4 Section the Union Electric Company for Alternative Source Operational QualityDiffusion Term Short-Term Assurance Manual.

Estimates.

REGULATORY GUIDE 1.145 Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants DISCUSSION:

UE complies with the recommendations described in the Draft Regulatory Guide 1.XXX (1978). Refer to Site Addendum Section 2.3.4.2.1 for a discussion of short-term diffusion estimates.

REGULATORY GUIDE 1.146 Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants DISCUSSION:

Refer to the Union Electric Company Operational Quality Assurance Manual.

REGULATORY GUIDE 1.160 REVISION 2 DATED 3/97 Monitoring the Effectiveness of Maintenance at Nuclear Power Plants DISCUSSION:

Refer to Appendix 3A of the Standard Plant FSAR.

REGULATORY GUIDE 1.194 DATED 6/03 Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants DISCUSSION:

The recommendations of this regulatory guide are met as described in Site Addendum Section 2.3.4.4 for Alternative Source Term Short-Term Diffusion Estimates.

ULNRC-06636 Page 274 of 374 Attachment D - Inserts for FSAR and Addendum

ULNRC-06636 Page 275 of 374 Insert 2.3.4.2.2 2.3.4.2.2.2 Alternative Source Term (AST Analysis)

The short-term atmospheric dispersion factors (/Qs) are based on onsite meteorological data for the Callaway Plant site. The diffusion equations and assumptions used in the calculations are those outlined in NRC Regulatory Guide 1.194, "USNRC, Regulatory Guide 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants, Revision 1, June 2003." Table 2.3-1 lists the limiting /Qs for the Callaway site. The detailed procedures used in the calculations are given in Section 2.3.4.4 of the Site Addendum.

ULNRC-06636 Page 276 of 374 Insert 6.5.5

4. NUREG-0800, Standard Review Plan Section 6.5.2, Revision 4, Containment Spray as a Fission product Cleanup System, March 2007.

Insert T6.5-1 Intake Pathways - Typical Accident Alignment Filter Removal Efficiency Aerosol Organic Elemental Nominal HEPA Iodine Iodine Flow (scfm)

Control Building (Compartment 12)

Path 16 Filtered Intake from Env 95% 0% 0% 900.0 (30 Min) 450.0 (to end)

Path 17 Unfiltered Inleakage from Env 0% 0% 0% 0 (up to 6000 maximum)(1)

Path 30 Unfiltered from CBER via HVAC 0% 0% 0% 1018.8 (flow balance) 688.8 (to end)

Path 40 Unfiltered from Normal HVAC 0% 0% 0% 0 (N/A for accident)

Control Room Filter Unit (CRFU) Intake Plenum (Compartment 13)

Path 18 Unfiltered Inleakage from CB 0% 0% 0% 440 Path 26 Unfiltered Inleakage from CR 0% 0% 0% 1030 (flow balance)

Path 28 Unfiltered from CBER 0% 0% 0% 330 Control Room Air Conditioning (CRAC) Mixing Plenum (Compartment 14)

Path 21 Filtered Intake from CRFU 95% 95% 95% 1800 Path 31 Unfiltered Inleakage from CB 0% 0% 0% 16200 (flow balance)

Control Building Equipment Room (Compartment 15)

Path 22 Unfiltered from CRAC 0% 0% 0% 385.0 Path 25 Unfiltered from Env 0% 0% 0% 663.8 (30 min) 0% 0% 0% 333.8 (to end)

Path 29 Unfiltered from Env 0% 0% 0% 300. 0 (Maximized)

Path 41 Unfiltered from Normal HVAC 0% 0% 0% 0 (N/A for accident)

Main Control Room (Compartment 16)

Path 19 Unfiltered from CB 0% 0% 0% 440 (30 Min) 0 (to end)

Path 23 Unfiltered from CRAC 0% 0% 0% 17615 Path 24 Unfiltered Inleakage from Env 0% 0% 0% 60 maximum(1)

Path 39 Unfiltered from Normal HVAC 0% 0% 0% 0 (N/A for accident)

Normal CB/CBER/CR HVAC (Compartment 19)

Path 36 Unfiltered Intake from Env 0% 0% 0% 18,000 until control room isolation, then 0

ULNRC-06636 Page 277 of 374 Environment (Compartment 11)

The environment compartment receives all exhaust/outleakage that does not enter other compartments. There are no dummy exhaust compartments in the dose model. All exhaust to the environment is unfiltered.

LOCA Path 13 Aux Bldg Emer Exhaust to Env 90% 90% 90% Flowrate Not Used (no mixing/no holdup in Aux Bldg)

FHA Path 13 Aux Bldg Emer Exhaust to Env 0% 0% 0%

Path 20 Exhaust from CB 0% 0% 0% 6993.8 (30 min) 6653.8 (to end)

Path 27 Exhaust from CR (accident) 0% 0% 0% 206.3 (30 min) 96.3 (to end)

Path 37 Exhaust from CR (normal) 0% 0% 0% 0 (N/A for accident)

Path 38 Exhaust from normal HVAC 0% 0% 0% 0 (N/A for accident)

Note: Compartment and pathway names and numbering are arbitrarily assigned, and are generally consistent with the dose analysis computer code model. Individual dose analysis models used in calculations may differ slightly in name or numbering scheme. See FSAR Figure 15A-3 for pathway number context.

(1) The limiting Control Building and Control Room inleakage values for radiological consequences are obtained from Figure 15A-2.

Insert 6.5A.3 Section III.4.c (1) of Reference 14 specifies the following formula for the spray removal of elemental iodine:

6K g TF S =

VD Where: S = spray removal rate coefficient for elemental iodine Kg = gas-phase mass transfer coefficient T = Fall time for spray droplets F = volumetric flow rate of the spray V = net free (air) volume of sprayed region D = mass-mean diameter of the spray drops From Table 6.5-2:

Containment volume: 2.70E6 ft3 Fraction containment sprayed: 85%

Average fall height: 131.4 ft Spray flow rate: 3086 gpm

ULNRC-06636 Page 278 of 374 Gas phase mass transfer coefficient, Kg: 9.5 ft/min Terminal Velocity: 790 ft/min:

Mean drop diameter: 831 microns The spray flow, F = (3086 gal/min)*(0.13368 ft3/gal) = 412.5 ft3/min In this application, a smaller flow produces a slower removal of iodine; therefore, the lowest expected flow value is used. As a conservative simplification, the increase in flow associated with the recirculation phase is neglected.

Volume sprayed, V = (2.7 million ft3 total) (0.85 as fraction sprayed) = 2.3 million ft3 In this application, a larger volume produces a slower removal of iodine.

Drop diameter, D = (831 microns)(1 m/106 microns)(3.281 ft/m) = 0.00273 ft The fall time (T) may be calculated as the ratio of the average fall height to the terminal velocity.

Fall time, T = 131.4 ft / 790 ft/min = 0.166 minutes

(

6(9.5 ft / min )(0.166 min ) 412.5 ft 3 / min )(60 min/ hr) s = = 37.3 hr 1 (2.30E + 06 ft )(0.00273 ft) 3 The Standard Review Plan (SRP) 6.5.2 (Reference 14) limits S to 20; therefore, 20 is used for the elemental iodine spray removal coefficient in the dose calculations of Section 15.6.5. In accordance with the SRP, the effectiveness of the spray in removing elemental iodine is required to end when the amount has been reduced by a factor of 200 (DF=200).

ULNRC-06636 Page 279 of 374 Insert 6.5A.4 6.5A.4 PARTICULATE MODEL FOR OFFSITE AND CONTROL ROOM DOSE CALCULATIONS Section III.4.c (4) of Reference 14 specifies the following formula for the spray removal of aerosols (particulates):

3hFE P = 2VD Where: P = spray removal rate coefficient for aerosols h = spray fall height F = volumetric flow rate of the spray E/D = ratio of dimensionless collection efficiency to spray drop diameter V = volume of sprayed region From Table 6.5-2:

Containment volume: 2.70E6 ft3 Spray flow rate: 3086 gpm Average Fall Height: 131.4 ft Spray Flow, F = (3086 gal/min)(0.13368 ft3/gal) = 412.5 ft3/min Volume sprayed, V = (2.7 million ft3 total) (0.85 as fraction sprayed) = 2.3 million ft3 SRP 6.5.2 (Reference 14) specifies that E/D = 10 m-1 initially and E/D = 1 m-1 after the aerosol mass has been reduced by a factor of 50.

( )

3(131.4 ft) 412.5 ft 3 / min (10/ m)(60 min/ hr) p = = 6.46 hr 1

(

2(3.281 ft / m) 2.30E + 06 ft 3

)

Appendix A of Reference 15 requires that the particulate spray removal coefficient be reduced by a factor of 10 when a Decontamination Factor of 50 is reached for the aerosols (particulates).

ULNRC-06636 Page 280 of 374 Insert 6.5A.5

14. NUREG-0800, Standard Review Plan Section 6.5.2, Revision 4, Containment Spray as a Fission product Cleanup System, March 2007.

15. Regulatory Guide 1.183, original version, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000

ULNRC-06636 Page 281 of 374 Insert Chapter15TOC-A 15A.2.2 GOVERNING EQUATIONS 15A.3 CONTROL ROOM RADIOLOGICAL CONSEQUENCES CALCULATION MODELS 15A.3.1 Integrated Activity In Control Room 15A.4 MODEL FOR RADIOLOGICAL CONSEQUENCES TO RADIOACTIVE SHINE AND TRANSIT DOSE OF CONTROL ROOM OPERATORS 15A.5 WASTE GAS DECAY TANK RUPTURE RADIOLOGICAL CONSEQUENCES EVALUATION MODELS AND PARAMETERS 15A.5.1 General Accident Parameters 15A.5.2 Offsite Radiological Consequences Calculational Models 15A.5.2.1 Accident Release Pathways 15A.5.2.2 Single - Region Release Model 15A.5.2.3 Offsite Thyroid Dose Calculation Model 15A.5.2.4 Offsite Total-Body Dose Calculational Model 15A.5.3 Control Room Radiological Consequences Calculational Models in WGDTR Insert Chapter15List_of_Tables 15B-1 Conformance with Regulatory Guide 1.183, Revision 0 Main Sections 15B-2 Conformance with Regulatory Guide 1.183, Revision 0, Appendix A (Loss of Coolant Accident) 15B-3 Conformance with Regulatory Guide 1.183, Revision 0, Appendix B (Fuel Handling Accident) 15B-4 Conformance with Regulatory Guide 1.183, Revision 0, Appendix E (PWR Main Steam Line Break Accident) 15B-5 Conformance with Regulatory Guide 1.183, Revision 0, Appendix F (PWR Steam Generator Tube Rupture Accident) 15B-6 Conformance with Regulatory Guide 1.183, Revision 0, Appendix G (PWR Locked Rotor Accident) 15B-7 Conformance with Regulatory Guide 1.183, Revision 0, Appendix H (PWR Rod Ejection Accident)

ULNRC-06636 Page 282 of 374 Insert 15.0.9 The calculation of the core fission product inventory employs ORIGEN-S of the Oak Ridge National Laboratory (ORNL) SCALE 6.1.3 code package (Reference 22). ORIGEN-S is an isotopic depletion and decay code which allows the user to specify fuel type, enrichment and periods of irradiation/decay and uses the latest cross-section data from ORNL to determine the existing nuclide inventory at specified intervals.

The core is modeled as an eight batch enveloping cycle core with a core power level of 3636 MWt (3565 MWt plus 2% postulated calorimetric error).

The Batch 1 assembly operated at an average power of 58.58 MW/MTU for 501.5 EFPD and 39.35 MW/MTU for 546.6 EFPD.

The Batch 2 assemblies operated at an average power of 55.23 MW/MTU for 502.5 EFPD, 21.17 MW/MTU for 498.5 EFPD, and 13.10 MW/MTU for 546.6 EFPD.

The Batch 3 assemblies operated at an average power of 53.25 for 498.5 EFPD and 34.39 MW/MTU for 546.6 EFPD.

The Batch 4 assemblies operated at an average power of 54.24 MW/MTU for 498.5 EFPD and 44.17 MW/MTU for 546.6 EFPD.

The Batch 5 assemblies operated at an average power of 51.42 MW/MTU for 498.5 EFPD and 46.39 MW/MTU for 546.6 EFPD. Batch 6 assemblies operated at an average power of 52.07 MW/MTU for 546.6 EFPD.

The Batch 7 assemblies operated at an average power of 47.15 MW/MTU for 546.6 EFPD.

The Batch 8 assemblies operated at an average power of 109.78 MW/MTU for 546.6 EFPD.

The total burnups in Batches 1 through 8 at the end of the analyzed cycle (MWD/MTU) are as follows:

Batch Number of Exposure assemblies (MWd/MTU) 1 1 50,885 2 16 45,462 3 60 45,340 4 12 51,176 5 4 50,985 6 60 28,460 7 32 25,770 8 8 60,000

ULNRC-06636 Page 283 of 374 The isotopic yields utilize data for fissioning of U-235, U-238, and Pu-239 and account for the depletion of the fuel. Radiological consequences are evaluated with source terms based on the 3636 MWt core rating (Table 15A-3), Callaway-specific meteorology based on four years of combined meteorological data (Table 15A-2), and appropriate dose conversion factors (Table 15A-4)

ULNRC-06636 Page 284 of 374 Insert 15.0.11.8 RETRAN-3D evolved from continued development of RETRAN-02. Both the steady-state and transient numerical solutions methods in RETRAN-3D have been revised to use an implicit solution. This results in much improved steady-state initialization convergence for two-phase systems and a more stable transient solution. RETRAN-3D retains the analysis capabilities of RETRAN-02 and also has improved modeling capability for small break loss-of-coolant accidents and anticipated transients without scram. RETRAN-3D also has model extensions designed to provide analysis capabilities for long-term transients and transients with limited thermodynamic nonequilibrium phenomena. RETRAN-3D is used by a large number of domestic and foreign electric utilities and research organizations. RETRAN-3D has been reviewed by the USNRC and was issued a generic SER in 2001 that removed many of the conditions for RETRAN-02. RETRAN-3D is implemented in the RETRAN-02 mode for the analyses supporting the Alternate Source Term (AST) steam release calculations.

ULNRC-06636 Page 285 of 374 Insert 15.1.5.3.1.2A Case 1 The initial reactor coolant concentrations of radioactive isotopes are assumed to be the dose equivalent of 1.0 Ci/gm of I-131 with an iodine spike that increases the rate of iodine release into the reactor coolant by a factor of 500. This increased rate of transfer to the coolant is assumed to be for the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Case 2 An assumed reactor coolant concentration of radioactive isotopes with a dose equivalent of 60Ci/gm of I-131 as a result of a pre-accident iodine spike.

Insert 15.1.5.3.1.2B

f. The reactor coolant concentrations of alkali metals correspond to 1-percent of the fuel having cladding defects as provided in Table 11.1-5.

Insert 15.1.5.3.2

d. Reactor coolant activities based on iodine spiking effects are conservatively high.

e. The most conservative return to power, decay heat, and metal mass sensible heat are all included in the steam release calculation to provide the most conservative result.

f. It is assumed that AFW is isolated to one of the intact loops as well as the faulted loop as part of the operator actions taken to isolate the faulted steam generator. Thus the ASDs on only two loops are used for the cooldown to RHR conditions which further extends the cooldown and therefore increases the total steam release Insert T15.1-3A Case 1 The initial reactor coolant concentrations of radioactive isotopes are assumed to be the dose equivalent of 1.0 Ci/gm of I-131 with an iodine spike that increases the rate of iodine release into the reactor coolant by a factor of 500. This increased rate of transfer to the coolant is assumed 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Case 2 An assumed reactor coolant concentration of radioactive isotopes with a dose equivalent of 60Ci/gm of I-131 as a result of a pre-accident iodine spike.

Insert T15.1-3B

e. Reactor coolant alkali metal activity:

1) Case 1 Based on 1-percent fuel having cladding defects as provided in Table 11.1-5

2) Case 2 Based on 1-percent fuel having cladding defects as provided in Table 11.1-5

ULNRC-06636 Page 286 of 374 Insert T15.1-3C

h. Alkali metal partition factors

1) Faulted steam generator 1

2) Intact steam generators 0.01

ULNRC-06636 Page 287 of 374 Insert 15.2.6.3.1.2A

a. The reactor coolant initial iodine activity is determined by two methods, and both cases are analyzed. These are:

Case 1 The initial reactor coolant concentrations of radioactive isotopes are assumed to be the dose equivalent of 1.0 Ci/gm of I-131 with an iodine spike that increases the rate of iodine release fuel into the into the reactor coolant by a factor of 500. This increased rate of transfer to the coolant is assumed to be for the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Case 2 An assumed reactor coolant concentration of radioactive isotopes with a dose equivalent of 60Ci/gm of I-131 as a result of a pre-accident iodine spike.

b. The reactor coolant activity assumed for noble gas is the Technical Specification limit of 225 Ci/gm Xe-1 33 dose equivalent

c. The reactor coolant system activity assumed for alkali metals is based on 1% fuel defects, as provided in Table 11.1-5.

Insert 15.2.6.3.1.2B

e. The alkali particulates are conservatively combined with, and treated as, halogens for transport through the steam generators.

Insert 15.2.6.3.1.2C The partition fraction for iodine and alkali metals in the steam generators is taken as:

i.) 0.01 for bulk boiling of the water in the secondary, and ii.) Equal to the fraction of primary-to-secondary leakage that flashes to steam. This flashing fraction is conservatively held at an initial value of 5% for the first 2.667 hours0.00772 days <br />0.185 hours <br />0.0011 weeks <br />2.537935e-4 months <br /> and decreased to zero thereafter.

Insert T15.2-2A

c. The reactor coolant initial iodine activity is determined by two methods, and both cases are analyzed. These are:

Case 1 The initial reactor coolant concentrations of radioactive isotopes are assumed to be the dose equivalent of 1.0 Ci/gm of I-131 with an iodine spike that increases the rate of iodine release fuel into the into the reactor coolant by a factor of 500. This increased rate of transfer to the coolant is assumed to be for the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Case 2 An assumed reactor coolant concentration of radioactive isotopes with a dose equivalent of 60Ci/gm of I-131 as a result of a pre-accident iodine spike.

ULNRC-06636 Page 288 of 374 Insert 15.3.3.3.1.1 The assumptions used to determine the initial concentrations of isotopes in the reactor coolant and secondary coolant prior to the accident are as follows:

a. Based upon inclusion of fuel failure in this event, the dose contribution from the initial RCS activity was neglected.

b. The secondary side coolant initial concentrations are assumed to be the dose equivalent of 10%

of 1.0 Ci/gm dose equivalent of I-131.

Insert 15.3.3.3.1.2

g. The partition factor for iodine released by bulk boiling in the steam generators is taken as 0.01 for secondary side releases.

h. Five percent of the primary-to-secondary leakage flashes to vapor during the first 2.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and has no mitigation when released to the environment. Ninety-five percent of the primary-to-secondary leakage mixes with the secondary water. These assumptions are conservatively based on a leak in the upper tubes which are assumed to be uncovered for the first 2.48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

ULNRC-06636 Page 289 of 374 Insert T15.3-3

c. Core Inventories See Table 15A-3

d. Radial leaking factor 1.65

e. Extent of core damage 5%

f. Percent of core inventory initially present in the fuel gap with a maximum of 35 rods per assembly exceeding Regulatory Guide 1.183 burnup limits:

Burnup >54 Burnup Isotope GWD/MTU* <54 GWD/MTU I-131 0.12 0.08 Kr-85 0.30 0.10 Other Noble Gases 0.10 0.05 Other Halogens 0.10 0.05 Alkali Metals 0.17 0.12

g. Iodine and alkali metal partition factor in the steam generators 0.01 for secondary side releases

h. Primary-to secondary leakage flashing to vapor 5%*

i. Primary-to-secondary leakage mixing with secondary water 95%*

• for the first 2.48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />

ULNRC-06636 Enclosure 6 Page 290 of 374 Insert 15.4.8.3.1.2A This is the percent of core inventory initially present in the fuel gap:

Burnup >54 Burnup Isotope GWD/MTU <54 GWD/MTU I-131 0.12 0.10 Kr-85 0.30 0.10 Other Noble Gases 0.10 0.10 Other Halogens 0.10 0.10 Alkali Metals 0.17 0.12 As a result of fuel failure, 10% of the fuel gap activity is released (in separate cases) to the reactor and to the containment atmosphere with adjustment for the radial power distribution. This release is in addition to that released from the assumed 0.25% fuel melt.

Insert 15.4.8.3.1.2B 100% of the noble gases in the melted fuel is released. 50% of the iodines in the melted fuel is released to the reactor coolant. In a separate case, 25% of the iodines in the melted fuel is released to the containment atmosphere.

Insert 15.4.8.3.1.2C

g. The partition factor for iodine in the steam generators is taken as 0.01 for secondary side releases.

h. Five percent of the primary-to-secondary leakage flashes to vapor for the first 2.622 hours0.0072 days <br />0.173 hours <br />0.00103 weeks <br />2.36671e-4 months <br /> and has no mitigation when released to the environment. Ninety-five percent of the primary-to-secondary leakage mixes with the secondary water for the first 2.622 hours0.0072 days <br />0.173 hours <br />0.00103 weeks <br />2.36671e-4 months <br /> and 100% thereafter. These assumptions are conservatively based on a leak in the upper tubes which are assumed to be uncovered for the accident duration.

Insert T15.4-3

k. Primary-to secondary leakage flashing to vapor 5%*

l. Primary-to-secondary leakage mixing with secondary water 95%*

• for the first 2.622 hours0.0072 days <br />0.173 hours <br />0.00103 weeks <br />2.36671e-4 months <br />

ULNRC-06636 Page 291 of 374 Insert 15.6.5.4.1 The analysis of the radiological consequences of a postulated LOCA uses the recommended dose conversion factors (DCFs) as follows:

1. The analysis uses Committed Effective Dose Equivalent (CEDE) dose conversion factors (DCFs) for inhalation of radionuclides based on the date provided on 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." (Reference 34)

2. The Effective Dose Equivalent (EDE) dose conversion factors, provided in Table III.1 of Federal Guidance Report 12, External Exposure to Radionuclides in Air, Water, and Soil, (Reference 35) are used to calculate the external exposure doses.

Insert 15.6.5.4.1.4 This calculation assumes:

1. The initial activity concentration in the Reactor Coolant System corresponds to the 1 µCi/gm Dose Equivalent Iodine-131 and 225 µCi/gm Dose Equivalent Xe-133 equilibrium limits. Iodine spikes need not be considered.

2. The release of all the fission products in the RCS to the containment is assumed to occur instantaneously following the break in reactor coolant piping.

3. Because the 11 second duration is less than the 30 second delay for onset of the gap release phase in Table 4 of Regulatory Guide 1.183, no direct release of activity from the fuel is applicable.

4. Two initially open 18 mini-purge lines are isolated at 11 seconds.

5. The maximum flow rate through the mini-purge lines is calculated as a function of containment pressure until isolation at 11 seconds. See Part g. of Table 15.6-6.

6. Containment pressure is calculated in response to the mass and energy release from a Double Ended Cold Leg pipe break with two open 18 mini-purge lines.

7. From FSAR Figure 11.3-2, the mini-purge exhaust is routed through the Unit Vent to the environment.

8. While the mini-purge flow leaving containment is normally filtered, this filtration is not included in the Engineered Safety Feature portion of system and so is not modeled.

9. Because the 11 sec duration of the release is faster/shorter than the time delay associated with control room isolation, no filtration of air flow to the control room is modeled.

ULNRC-06636 Page 292 of 374 Insert 15.6.5.4.2

d. Atmospheric Dispersion Exclusion Area Boundary and LPZ. The meteorological conditions assumed to be present at the site during the course of the accident are based on /Q values, which are the larger of the 5 percent overall site values and the 0.5 percent maximum sector values. This condition results in the poorest values of atmospheric dispersion calculated for the exclusion area boundary and the LPZ outer boundary. Furthermore, no credit has been taken for the transit time required for activity to travel from the point of release to the exclusion area boundary and LPZ outer boundary. Hence, the radiological consequences evaluated under these conditions are conservative.

Control Room. The meteorological conditions assumed to be present at the site during the course of the accident are based on /Q values, which are expected to be exceeded 5 percent of the time.

No credit has been taken for the transit time required for activity to travel from the point of release to the exclusion area boundary and LPZ outer boundary. Hence, the radiological consequences evaluated under these conditions are conservative.

ULNRC-06636 Page 293 of 374 Insert T15.6-1 Accident Event Time (sec)

Steam generator tube rupture with stuck-open atmospheric steam dump (ASD) valve Tube rupture occurs 0.0 Reactor trip and Safety Injection Injection Signal 600*

Loss of offsite power 600*

Ruptured steam generator atmospheric steam dump valve opens 600*

Delivery of Safety Injection begins ~620 Auxiliary feedwater injection begins ~660 Operator isolates ruptured steam generator by manually closing block valve 1800 = 600 + 1200 Operator initiates RCS cooldown via intact steam generator atmospheric steam dump valves 2400 = 1800 + 600 Operator completes RCS cooldown 3160 Operator initiates RCS depressurization via pressurizer PORVs 3340 = 3160 + 180 sec Operator completes RCS depressurization 3400 = 3340 + 60 sec Operator terminates safety injection 3700 = 3400 + 300 sec Operator equalizes primary-secondary pressure 4600 = 3700 + 900 sec Operator opens intact SG ASDs 4800 Operator opens block valve on ruptured SG 19,000 RHR cut-in conditions reached 21,100 = 5.86 hours9.953704e-4 days <br />0.0239 hours <br />1.421958e-4 weeks <br />3.2723e-5 months <br />

• This value reflects application of a slight conservative bias.

ULNRC-06636 Page 294 of 374 Insert T15.6-4sheet2 III. Activity Release Data stuck open ASD overfill case

a. Ruptured steam generator

1. Reactor coolant discharged to steam generator, lbs 300,000 (1) 240,000 (2)

2. Flashed reactor coolant, % 16.1 for first zero (4) 3000 sec (3)

3. Iodine partition factor for flashed fraction of reactor coolant 1.0 not applicable

4. Steam release to atmosphere, lbs 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 114,107 1.361 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 13,598 2 - 5.86 hours9.953704e-4 days <br />0.0239 hours <br />1.421958e-4 weeks <br />3.2723e-5 months <br /> 111,380 2 - 9.39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br /> 83,769 (5)

5. Iodine carry over factor for the non-flashed fraction of reactor coolant that mixes with 0.01 0.01 the initial iodine activity in the steam generator (as bulk boiling) for steam release

6. Liquid release to atmosphere, lbs 0.264 - 1.361 hours0.00418 days <br />0.1 hours <br />5.968915e-4 weeks <br />1.373605e-4 months <br /> 194,985

7. Iodine partition factor for liquid release from SG 0.5

8. Iodine partition factor for normal steam flow to condenser prior to reactor trip and loss of offsite power 0.01 not applicable

b. Unaffected steam generators

1. Primary-to-secondary leakage, lbs 2932 (6) 3198 (7)

2. Flashed reactor coolant, % 16.1 for first 4.0 between 6000 3000 sec (3) and 9000 sec

3. Total steam release, lbs 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 321,930 396,435 2 - 5.86 hours9.953704e-4 days <br />0.0239 hours <br />1.421958e-4 weeks <br />3.2723e-5 months <br /> 647,375 2 - 6.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 581,117

4. Iodine carry over factor 0.01 0.01 for bulk boiling

5. RHR Cut-in time, hours 5.86 6.4

ULNRC-06636 Page 295 of 374 Insert T15.6-4sheet3 Notes:

(1) While RETRAN calculates a total integrated break flow of 220,243 lbs, this is conservatively biased to a value of (60 lb/sec) x (5000 sec) = 300,000 lbs for use as the basis for offsite and control room dose.

(2) While RETRAN calculates a total integrated break flow of 193,200 lbs, this is conservatively biased to a value of (60 lb/sec) x (4000 sec) = 240,000 lbs for use as the basis for offsite and control room dose.

(3) The 16.1% flashing fraction is based on initial conditions in the RCS and steam generator secondary side. As a conservative simplification, this fraction is applied for the first 3000 seconds in both the ruptured and intact steam generators. After that, the RETRAN computer code model used to calculate steam releases to the environment shows that the collapsed water level in the secondary side covers the entire tube bundle. The SG tubes remain covered with water for the remainder of the accident.

(4) For the overfill case, the collapsed liquid level in the secondary side of the ruptured SG remains above the top of the tube bundle for the duration of the accident. Since the tube break is covered with water, credit is taken for scrubbing by liquid in the secondary side.

(5) While RHR cut-in conditions are reached at 6.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, additional time is required to reduce the ruptured steam generator secondary side temperature below 212F.

(6) All of the normal allowable steam generator tube leakage is conservatively assumed to occur within the three intact steam generators. The total leakage is 1 gpm = 8.34 lb/min for a duration of 5.86 hours9.953704e-4 days <br />0.0239 hours <br />1.421958e-4 weeks <br />3.2723e-5 months <br />.

(7) The total leakage is 1 gpm = 8.333 lb/min for a duration of 6.397 hours0.00459 days <br />0.11 hours <br />6.564153e-4 weeks <br />1.510585e-4 months <br />.

ULNRC-06636 Page 296 of 374 Insert T15.6-6A

c. Release fraction and timing of core activity in the containment:

Gap Release Phase Early In-vessel Phase Group Fraction Fraction (30 sec-0.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />) (0.5 hour-1.3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) 1 Noble gases 0.05 0.95 2 Halogens 0.05 0.35 3 Alkali metals 0.05 0.25 4 Tellurium metals 0.00 0.05 5 Barium and Strontium 0.00 0.02 6 Noble metals 0.00 0.0025 7 Cerium group 0.00 0.0005 8 Lanthanides 0.00 0.0002 Insert T15.6-6B

g. Equilibrium sump pH > 7.0

h. Reactor Coolant Activity (mini-purge only)

1. Iodine Dose equivalent of 1.0 Ci/gm of I-131

2. Noble gas Dose equivalent of 225 Ci/gm of Xe-133

3. Alkali metal Based on 1% failed fuel as provided in Table 11.1-5 Insert T15.6-6C

g. Mini-purge initially in operation initial air mass in Containment = 172,222 lb.

total RCS fluid mass (released to containment) = 551,068 lb.

ULNRC-06636 Page 297 of 374 Flow from Containment to Environment Containment Mini-purge flow rate Time after break pressure to environment (sec) (psia) (lbm/min) 0 14.7 0.00E+00 0.6 19.7 1.09E+04 0.8 21.1 1.22E+04 1 22.5 1.32E+04 2 28.4 1.65E+04 3 32.8 1.85E+04 4 36.3 2.01E+04 5 39.3 2.15E+04 5.5 40.7 2.21E+04 6 42 2.27E+04 6.5 43.2 2.33E+04 7 44.3 2.38E+04 7.5 45.3 2.43E+04 8 46.2 2.47E+04 8.5 47 2.50E+04 9 47.6 2.53E+04 9.5 48.2 2.56E+04 10 48.7 2.58E+04 10.5 49.2 2.60E+04 10.999 49.5 2.62E+04 11 0.00E+00

ULNRC-06636 Page 298 of 374 Insert 15.7.4.5.1.2.e is defined by Table 3 of Reg Guide 1.183 (Reference 1) that lists the fraction of the fission product inventory that is in the fuel gap subject to the following provisions:

The release fractions listed here have been determined to be acceptable for use with currently approved LWR fuel with a peak burnup up to 62,000 MWD/MTU provided that the maximum linear heat generation rate does not exceed 6.3 kw/ft peak rod average power for burnups exceeding 54 GWD/MTU For the assumed maximum of 32 fuel rods per assembly that does not meet the above criteria (total of 264 rods per assembly), release fractions from NUREG/CR-5009 (Reference 2) have been applied. The table below lists the Reg. Guide 1.183 and NUREG/CR-5009 gap fractions used for Non-LOCA accidents.

Non-LOCA Fraction of Fission Products in Fuel Gap RG1.183 Table 3 NUREG/CR-5009 Group Fraction Fraction I-131 0.08 0.12 Kr-85 0.10 0.30 Other Noble Gases 0.05 0.10 Other Halogens 0.05 0.10 Alkali Metals 0.12 0.17 The gap release from the fuel during the accident has been adjusted to account for the higher release fraction in the portion of high burnup fuel in each assembly. For the FHA in the containment, the 20% of fuel rods damaged in the additional assembly are conservatively assumed to include all of the high burnup rods in that assembly.

Insert 15.7.4.5.1.2.g As noted in Item 8 of Reference 3, the elemental iodine decontamination factor to be used is 285. The organic decontamination factor is 1.0, for an overall iodine decontamination factor of 200.

Insert 15.7.4.5.1.2.h Per Section 1.3 of Appendix B of Regulatory Guide 1.183, the chemical form of radioiodine released from the fuel to the spent fuel pool is assumed to be 95% cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. The CsI released from the fuel is assumed to completely re-evolve as elemental iodine. Therefore, the chemical form of the released iodine is 99.85% elemental and 0.15%

organic.

ULNRC-06636 Page 299 of 374 Following application of the iodine decontamination factors discussed in 15.7.4.5.1.2.g the resulting chemical composition of the iodine release above the pool is 70% elemental and 30% organic.

Insert 15.7.5 15.

7.5 REFERENCES

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

2. NUREG/CR-5009, Assessment of the Use of Extended Burnup Fuel in Light Water Power Reactors February 1988

3. NRC Regulatory Issue Summary (RIS) 2006-04: Experience with Implementation of Alternative Source Terms.

ULNRC-06636 Page 300 of 374 Insert 15A.2.2 15A.2.2 Governing Equations It is assumed that any activity released to the holdup system instantaneously diffuses to uniformly occupy the system volume. The following equation, from Reference 10, applies to the storage, source, transport and removal of radionuclides in compartment i, including the transport from the previous compartment h and to the next compartment j. This calculational model applies to both single and multiple region release models. Potential removal mechanisms considered include containment spray, natural deposition, filters, suppression pools, decay and release. The RADTRAD computer code applies the calculational model to all compartments during the event to determine the integrated offsite TEDE dose.

1

, = ,, + ,

=1

,() ,()

[,() + ,() + + ] + + , + ,

=1

[

+

,() ,

100

]

, ,() ,()

+ [(1 ),() + ,() + + ] ,

100

=1 () ()

(1) where ln(2)

= 1/2 (2) and 1/2

= half life of nuclide [s]

, = number of atoms of nuclide in compartment [dimensionless]

, = fraction of nuclide that decays to nuclide [dimensionless]

= radiological decay constant for nuclid [s1]

, = source rate of nuclide in compartment [atoms/s]

,()

= volume-normalized flow rate from compartment to compartment through a convection pathway [s1]

,() = volume-normalized filtered flow rate from compartment to compartment through a pipe [s1]

,() = volumetric flow rate from compartment to compartment through a suppression pool [ft3/s]

,() = volumetric flow rate from compartment to compartment through a pipe [ft3/s]

() = suppression pool decontamination factor for nuclide [dimensionless]

ULNRC-06636 Page 301 of 374

() = piping decontamination factor for nuclide [dimensionless]

, = spray removal coefficient for nuclide [s1]

, = natural deposition removal coefficient for nuclide [s1]

, = filter efficiency for nuclide for a recirculating filter in compartment [%]

, = filter efficiency for nuclide for a filter in the pathway from compartment to compartment [%]

= number of compartments defined in the model [dimensionless]

= volume of compartment [ft3]

Insert 15A.4,5 15A.4 MODEL FOR RADIOLOGICAL CONSEQUENCES DUE TO RADIOACTIVE SHINE AND TRANSIT DOSE TO CONTROL ROOM OPERATORS The shine dose is calculated based on the semi-infinite radioactive cloud surrounding the Control Room, the shine from the radioactivity inside Containment, and the shine from the Control Room Filtration Unit filter loading. The three source terms (environment, containment, and filter) from the LOCA analysis are used in combination with the MicroShield computer code to determine the total shine contribution to the Control Room dose.

The transit dose is calculated in the LOCA analysis based on additional dose points in the RADTRAD-NAI models representing the operators path to and from the Control Room. The four components of the transit dose (inhalation, immersion, containment shine, and ground deposition) are calculated using the RADTRAD-NAI code and/or its output to determine the total transit dose contribution to the operator.

Transit dose is added to the separately determined Control Room dose and control room shine dose.

15A.5 WASTE GAS DECAY TANK RUPTURE ANALYSIS RADIOLOGICAL CONSEQUENCES EVALUATION MODELS AND PARAMETERS This section is historical as the Alternative Source Term (AST), as described in Regulatory Guide 1.183 (Reference 6), was not applied to the waste gas decay tank rupture (WGDTR) accident.

15A.5.1 GENERAL ACCIDENT PARAMETERS This section contains the parameters used in analyzing the radiological consequences of a waste gas decay tank rupture (WGDTR). Refer to Section 15.7 for more details on WGDTR parameters. The site specific, ground-level release, short-term dispersion factors (for accidents, ground-level releases are assumed) are based on Regulatory Guide 1.XXX (Reference 7) methodology and the 0.5 percent worst-sector meteorology and these are given in Table 15A-2 (see Section 2.3.4 and the Site Addendum for additional details on meteorology). The core and gap inventories for WGDTR are given in Table 15A-3.

The thyroid (via inhalation pathway), beta skin, and total-body (via submersion pathway) dose factors based on References 2 and 3 are given in Table 15A-4.

15A.5.2 OFFSITE RADIOLOGICAL CONSEQUENCES CALCULATIONAL MODELS This section presents the models and equations used for calculating the integrated activity released to the environment, the accident flow paths, and the equations for dose calculations. One major release models are considered: a single holdup system with no internal cleanup.

ULNRC-06636 Page 302 of 374 15A.5.2.1 ACCIDENT RELEASE PATHWAYS The activity release to the environment due to waste gas decay tank rupture (WGDTR) will be direct and unfiltered, with no holdup. The total activity release in this case is therefore assumed to be the initial source activity itself.

15A.5.2.2 SINGLE - REGION RELEASE MODEL It is assumed that any activity released to the holdup system instantaneously diffuses to uniformly occupy the system volume.

The following equations are used to calculate the integrated activity released from postulated accidents.

A1(0) = initial source activity at time to, Ci

ULNRC-06636 Page 303 of 374 15A.5.2.3 OFFSITE THYROID DOSE CALCULATION MODEL Offsite thyroid doses are calculated using the equation:

15A.5.2.4 OFFSITE TOTAL-BODY DOSE CALCULATIONAL MODEL Assuming a semi-infinite cloud of gamma emitters, offsite total-body doses are calculated using the equation:

ULNRC-06636 Page 304 of 374 15A.5.3 CONTROL ROOM RADIOLOGICAL CONSEQUENCES CALCULATIONAL MODELS in WGDTR The WGDTR analysis does not provide control room doses.

Insert 15A.6

6. USNRC Regulatory Guide 1.183, Revision 0, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000

7. DELETED.

8. EPA Federal Guidance Report No. 11, EPA-520/1-88-020, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1988.

9. EPA Federal Guidance Report No. 12, EPA-402-R-93-081, "External Exposure to Radionuclides in Air, Water, and Soil," 1993.

10. NUREG/CR-6604, "RADTRAD: A Simplified Model for RADionuclide Transport and Removal And Dose Estimation," December 1997.

ULNRC-06636 Page 305 of 374 Insert T15A-2 Note: The WGDTF (Section 15.7.1) and Radioactive Liquid Waste System Leak or Failure (Section 15.7.2) use the Atmospheric Dispersion Factors on Sheet 5

/Qs for Alternative Source Term (AST) Radiological Consequences Location Type/

Event(s) (Release Location)/

Time Interval /Q (hours) (Sec/meters3)

Exclusion Area Boundary RWST Vent 0 - 720 2.05E-04 Reactor Building/Other Onsite Release Locations 0 - 720 2.00E-04 Low Population Zone RWST Vent 0-2 6.87E-05 2-8 3.57E-05 8 - 24 2.57E-05 24 - 96 1.26E-05 96 - 720 4.54E-06 Reactor Building/Other Onsite Release Locations 0-2 6.87E-05 2-8 3.42E-05 8 - 24 2.42E-05 24 - 96 1.13E-05 96 - 720 3.83E-06 Control Room LOCA Containment Leakage, Rod Ejection (Diffuse Containment) 0 - Isolation 7.12E-03 Isolation - 2 7.49E-04 2-8 5.32E-04 8 - 24 2.29E-04 24 - 96 1.50E-04 96 - 720 9.56E-05

ULNRC-06636 Page 306 of 374 TABLE 15A-2 (Sheet 2)

Release Location/

Time Interval /Q (hours) (Sec/meters3)

LOCA Mini-Purge1 & ECCS Leakage, FHA in FHB2 (Unit Vent Exhaust) 0 - Isolation (LOCA Mini-Purge & ECCS 1.90E-03 Leakage) 0-Isolation (FHA in FHB) 2.23E-03 Isolation - 2 6.86E-04 2-8 5.72E-04 8 - 24 2.32E-04 24 - 96 1.42E-04 96 - 720 9.57E-05 Letdown Line Break1 (Unit Vent Exhaust) 0-2 1.90E-03 2-8 1.58E-03 8 - 24 6.67E-04 24 - 96 3.90E-04 96 - 720 2.29E-04 LOCA RWST Backleakage (RWST Vent) 0 - Isolation 9.28E-04 Isolation - 2 7.47E-04 2-8 6.55E-04 8 - 24 2.71E-04 24 - 96 1.52E-04 96 - 720 9.17E-05 FHA in Containment (Emergency Personnel Access Hatch3) 0 - Isolation3 7.12E-03 Isolation - 2 8.61E-04 2-8 7.54E-04 8 - 24 3.22E-04 24 - 96 1.84E-04 96 - 720 1.43E-04 1

In this accident, the control room never isolates, so the normal intake receptor location is used for the entire accident.

2 The closest point of the FHB is used before isolation, since it has a higher /Q value.

3 Diffuse leakage through the containment wall is used before isolation instead, since it has a higher /Q value.

ULNRC-06636 Page 307 of 374 TABLE 15A-2 (Sheet 3)

Release Location/

Time Interval /Q (hours) (Sec/meters3)

Locked Rotor, SGTR (Closest ASD4) 0 - Isolation4 1.76E-02 Isolation - 2 1.74E-03 2-8 1.33E-03 8 - 24 6.50E-04 24 - 96 3.62E-04 96 - 720 2.96E-04 LOOP1 (MSSV) 0-2 1.76E-02 2-8 1.46E-02 8 - 24 6.74E-03 24 - 96 3.81E-03 96 - 720 3.05E-03 MSLB (Closest MSL Point5) 0 - Isolation5 1.76E-02 Isolation - 25 1.74E-03 2-8 1.56E-03 8 - 24 6.61E-04 24 - 96 3.83E-04 96 - 720 3.22E-04 4

The closest MSSV is used before isolation instead, since it has a higher /Q value.

5 The closest MSSV is used before isolation instead, since it has a higher /Q value. Additionally, the closest ASD is used for the first two hours instead, since it has a higher /Q value.

ULNRC-06636 Page 308 of 374 TABLE 15A-2 (Sheet 5)

LIMITING SHORT-TERM ATMOSPHERIC DISPERSION FACTORS (/QS) FOR WGDTR ANALYSIS

/Qs Applicable to: Waste Gas Decay Tank Failure (WGDTF, Section 15.7.1) and Radioactive Liquid Waste System Leak or Failure (Section 15.7.2)

Location Type/ /Q Time Interval (hrs) (sec/meter3)

Site boundary 1.5E-4 0-2 Low-population zone 0-8 1.5E-5 8-24 1.0E-5 24-96 4.6E-6 96-720 1.5E-6 Note that the WGDTR /Qs were not revised for the alternative source term (AST) analysis.

ULNRC-06636 Page 309 of 374 Insert T15A-3 TABLE 15A-3 CORE INVENTORY (Ci)

Isotope Core Activity (Ci) Isotope Core Activity (Ci) Isotope Core Activity (Ci)

Kr-85 9.677E+05 Cs-134 1.405E+07 Te-125m 1.590E+05 Kr-85m 2.469E+07 Cs-136 4.531E+06 Te-133m 9.382E+07 Kr-87 4.866E+07 Cs-137 1.008E+07 Ba-141 1.587E+08 Kr-88 6.507E+07 Ba-139 1.807E+08 Ba-137m 9.583E+06 Rb-86 1.834E+05 Ba-140 1.715E+08 Pd-109 3.214E+07 Sr-89 9.252E+07 La-140 1.777E+08 Rh-106 5.544E+07 Sr-90 7.220E+06 La-141 1.594E+08 Rh-103m 1.540E+08 Sr-91 1.151E+08 La-142 1.515E+08 Tc-101 1.680E+08 Sr-92 1.235E+08 Ce-141 1.620E+08 Eu-154 5.726E+05 Y-90 7.816E+06 Ce-143 1.492E+08 Eu-155 2.368E+05 Y-91 1.214E+08 Ce-144 1.213E+08 Eu-156 2.182E+07 Y-92 1.250E+08 Pr-143 1.460E+08 La-143 1.475E+08 Y-93 1.416E+08 Nd-147 6.421E+07 Nb-97 1.662E+08 Zr-95 1.651E+08 Np-239 1.907E+09 Nb-95m 1.894E+06 Zr-97 1.651E+08 Pu-238 2.539E+05 Pm-147 1.575E+07 Nb-95 1.659E+08 Pu-239 2.836E+04 Pm-148 1.720E+07 Mo-99 1.811E+08 Pu-240 3.890E+04 Pm-149 5.969E+07 Tc-99m 1.603E+08 Pu-241 1.171E+07 Pm-151 1.881E+07 Ru-103 1.541E+08 Am-241 1.130E+04 Pm-148m 3.619E+06 Ru-105 1.080E+08 Cm-242 3.128E+06 Pr-144 1.222E+08 Ru-106 4.835E+07 Cm-244 2.900E+05 Pr-144m 1.699E+06 Rh-105 9.707E+07 I-130 1.809E+06 Sm-153 4.468E+07 Sb-127 8.907E+06 Kr-83m 1.164E+07 Y-94 1.493E+08 Sb-129 2.816E+07 Xe-138 1.697E+08 Y-95 1.591E+08 Te-127 8.717E+06 Xe-131m 1.280E+06 Y-91m 6.664E+07 Te-127m 1.443E+06 Xe-133m 6.204E+06 Br-82 3.049E+05 Te-129 2.590E+07 Xe-135m 4.176E+07 Br-83 1.154E+07 Te-129m 4.971E+06 Cs-138 1.785E+08 Br-84 2.094E+07 Te-131m 1.886E+07 Cs-134m 4.054E+06 Am-242 6.242E+06 Te-132 1.390E+08 Rb-88 6.626E+07 Np-238 3.962E+07 I-131 9.758E+07 Rb-89 8.706E+07 Pu-243 4.014E+07 I-132 1.414E+08 Sb-124 7.037E+04 I-133 1.996E+08 Sb-125 7.418E+05 I-134 2.240E+08 Sb-126 4.691E+04 I-135 1.893E+08 Te-131 8.272E+07 Xe-133 1.995E+08 Te-133 2.890E+05 Xe-135 4.704E+07 Te-134 1.772E+08

ULNRC-06636 Page 310 of 374 Table 15A-3 (Sheet 2)

WGDTR Fuel and Rod Gap Activities Core Isotope Fuel Gap I-131 9.95E+7 9.95E+6 I-132 1.44E+8 1.44E+7 I-133 2.04E+8 2.04E+7 I-134 2.25E+8 2.25E+7 I-135 1.91E+8 1.91E+7 Kr-83m 1.27E+7 1.27E+6 Kr-85m 2.72E+7 2.72E+6 Kr-85 8.61E+5 2.58E+5 Kr-87 5.24E+7 5.24E+6 Kr-88 7.38E+7 7.38E+6 Kr-89 9.03E+7 9.03E+6 Xe-131m 1.12E+6 1.12E+5 Xe-133m 6.35E+6 6.35E+5 Xe-133 1.99E+8 1.99E+7 Xe-135m 3.96E+7 3.96E+6 Xe-135 4.38E+7 4.38E+6 Xe-137 1.78E+8 1.78E+7 Xe-138 1.70E+8 1.70E+7

• For WGDTR, the gap activity is assumed to be 10 percent of fuel activity for all isotopes except for Kr85; for Kr-85 it is assumed to be 30 percent of the fuel activity.

ULNRC-06636 Enclosure 6 Page 311 of 374 Insert T15A-4 Table 15A-4 DOSE CONVERSION FACTORS USED IN ALTERNATIVE SOURCE TERM (AST) ACCIDENT ANALYSIS FSAR sections re-analyzed for radiological consequences as part of the replacement steam generator program and, separately, the Alternative Source Term (AST) implementation include the following accidents:

15.1.5 STEAM SYSTEM PIPING FAILURE 15.2.6 LOSS OF NONEMERGENCY AC POWER TO THE STATION AUXILIARIES 15.3.3 REACTOR COOLANT PUMP SHAFT SEIZURE (LOCKED ROTOR) 15.4.8 SPECTRUM OF ROD CLUSTER CONTROL ASSEMBLY EJECTION ACCIDENTS 15.6.2 BREAK IN INSTUMENT LINE OR OTHER LINES FROM REACTOR COOLANT PRESSURE BOUNDARY THAT PENETRATE CONTAINMENT 15.6.3 STEAM GENERATOR TUBE FAILURE Additional FSAR sections re-analyzed for AST radiological consequences include the following accidents:

15.6.5 LOSS-OF-COOLANT ACCIDENTS RESULTING FROM A SPECTRUM OF POSTULATED PIPING BREAKS WITHIN THE REACTOR COOLANT PRESSURE BOUNDARY 15.7.4 FUEL HANDLING ACCIDENTS The following DCFs are based on Federal Guidance Reports 11 (Reference 8) and 12 (Reference 9).

Inhalation Inhalation Air Submersion Air Submersion Contaminated Contaminated CEDE* CEDE* EDE* EDE* Ground Ground Isotope Sv/Bq Rem/Ci (Sv-m3)/(Bq-sec) Rem-m3/Ci-sec EDE* EDE*

(Sv-m2)/(Bq-sec) Rem-m2/Ci-sec Kr-85m N/A N/A 7.48E-15 2.77E-02 1.52E-16 5.63E-04 Kr-85 N/A N/A 1.19E-16 4.40E-04 2.64E-18 9.78E-06 Kr-87 N/A N/A 4.12E-14 1.52E-01 7.32E-16 2.71E-03 Kr-88 N/A N/A 1.02E-13 3.77E-01 1.74E-15 6.44E-03 Xe-131m N/A N/A 3.89E-16 1.44E-03 2.06E-17 7.63E-05 Xe-133m N/A N/A 1.37E-15 5.07E-03 4.07E-17 1.51E-04 Xe-133 N/A N/A 1.56E-15 5.77E-03 4.61E-17 1.71E-04 Xe-135m N/A N/A 2.04E-14 7.55E-02 4.24E-16 1.57E-03 Xe-135 N/A N/A 1.19E-14 4.40E-02 2.42E-16 8.96E-04 Xe-138 N/A N/A 5.77E-14 2.13E-01 1.03E-15 3.81E-03 I-130 7.14E-10 2.64E+03 1.04E-13 3.85E-01 2.1E-15 7.78E-03 I-131 8.89E-09 3.29E+04 1.82E-14 6.73E-02 3.76E-16 1.39E-03 I-132 1.03E-10 3.81E+02 1.12E-13 4.14E-01 2.21E-15 8.19E-03 Table 15A-4 (Sheet 2)

ULNRC-06636 Page 312 of 374 Inhalation Inhalation Air Submersion Air Submersion Contaminated Contaminated CEDE* CEDE* EDE* EDE* Ground Ground Isotope Sv/Bq Rem/Ci (Sv-m3)/(Bq-sec) Rem-m3/Ci-sec EDE* EDE*

(Sv-m2)/(Bq-sec) Rem-m2/Ci-sec I-133 1.58E-09 5.85E+03 2.94E-14 1.09E-01 5.97E-16 2.21E-03 I-134 3.55E-11 1.31E+02 1.30E-13 4.81E-01 2.53E-15 9.37E-03 I-135 3.32E-10 1.23E+03 7.98E-14 2.95E-01 1.47E-15 5.44E-03 Cs-134 1.25E-08 4.63E+04 7.57E-14 2.80E-01 1.52E-15 5.63E-03 Cs-136 1.98E-09 7.33E+03 1.06E-13 3.92E-01 2.09E-15 7.74E-03 Cs-137 8.63E-09 3.19E+04 7.74E-18 2.86E-05 2.85E-19 1.06E-06 Cs-138 2.74E-11 1.01E+02 1.21E-13 4.48E-01 2.19E-15 8.11E-03 Rb-86 1.79E-09 6.62E+03 4.81E-15 1.78E-02 9.31E-17 3.45E-04 Te-127m 5.81E-09 2.15E+04 1.47E-16 5.44E-04 1.13E-17 4.19E-05 Te-127 8.60E-11 3.18E+02 2.42E-16 8.95E-04 5.18E-18 1.92E-05 Te-129m 6.47E-09 2.39E+04 1.55E-15 5.74E-03 3.78E-17 1.40E-04 Te-129 2.42E-11 8.95E+01 2.75E-15 1.02E-02 6.01E-17 2.23E-04 Te-131m 1.73E-09 6.40E+03 7.01E-14 2.59E-01 1.37E-15 5.07E-03 Te-132 2.55E-09 9.44E+03 1.03E-14 3.81E-02 2.28E-16 8.44E-04 Sb-127 1.63E-09 6.03E+03 3.33E-14 1.23E-01 6.76E-16 2.50E-03 Sb-129 1.74E-10 6.44E+02 7.14E-14 2.64E-01 1.38E-15 5.11E-03 Sr-89 1.12E-08 4.14E+04 7.73E-17 2.86E-04 2.27E-18 8.41E-06 Sr-90 3.51E-07 1.30E+06 7.53E-18 2.79E-05 2.84E-19 1.05E-06 Sr-91 4.49E-10 1.66E+03 3.45E-14 1.28E-01 6.77E-16 2.51E-03 Sr-92 2.18E-10 8.07E+02 6.79E-14 2.51E-01 1.25E-15 4.63E-03 Ba-139 4.64E-11 1.72E+02 2.17E-15 8.03E-03 4.59E-17 1.70E-04 Ba-140 1.01E-09 3.74E+03 8.58E-15 3.17E-02 1.8E-16 6.67E-04 Ru-103 2.42E-09 8.95E+03 2.25E-14 8.33E-02 4.63E-16 1.71E-03 Ru-105 1.23E-10 4.55E+02 3.81E-14 1.41E-01 7.69E-16 2.85E-03 Ru-106 1.29E-07 4.77E+05 0 0.00E+00 0.00E+00 0.00E+00 Rh-105 2.58E-10 9.55E+02 3.72E-15 1.38E-02 7.62E-17 2.82E-04 Mo-99 1.07E-09 3.96E+03 7.28E-15 2.69E-02 1.47E-16 5.44E-04 Tc-99m 8.80E-12 3.26E+01 5.89E-15 2.18E-02 1.21E-16 4.48E-04 Ce-141 2.42E-09 8.95E+03 3.43E-15 1.27E-02 7.38E-17 2.73E-04

ULNRC-06636 Page 313 of 374 Table 15A-4 (Sheet 3)

Inhalation Inhalation Air Submersion Air Submersion Contaminated Contaminated CEDE* CEDE* EDE* EDE* Ground Ground Isotope Sv/Bq Rem/Ci (Sv-m3)/(Bq-sec) Rem-m3/Ci-sec EDE* EDE*

(Sv-m2)/(Bq-sec) Rem-m2/Ci-sec Ce-143 9.16E-10 3.39E+03 1.29E-14 4.77E-02 2.79E-16 1.03E-03 Ce-144 1.01E-07 3.74E+05 8.53E-16 3.16E-03 2.03E-17 7.52E-05 Pu-238 1.06E-04 3.92E+08 4.88E-18 1.81E-05 8.38E-19 3.10E-06 Pu-239 1.16E-04 4.29E+08 4.24E-18 1.57E-05 3.67E-19 1.36E-06 Pu-240 1.16E-04 4.29E+08 4.75E-18 1.76E-05 8.03E-19 2.97E-06 Pu-241 2.23E-06 8.25E+06 7.25E-20 2.68E-07 1.93E-21 7.15E-09 Np-239 6.78E-10 2.51E+03 7.69E-15 2.85E-02 1.63E-16 6.04E-04 Y-90 2.28E-09 8.44E+03 1.90E-16 7.03E-04 5.32E-18 1.97E-05 Y-91 1.32E-08 4.88E+04 2.60E-16 9.62E-04 5.74E-18 2.13E-05 Y-92 2.11E-10 7.81E+02 1.30E-14 4.81E-02 2.53E-16 9.37E-04 Y-93 5.82E-10 2.15E+03 4.80E-15 1.78E-02 9.12E-17 3.38E-04 Nb-95 1.57E-09 5.81E+03 3.74E-14 1.38E-01 7.48E-16 2.77E-03 Zr-95 6.39E-09 2.36E+04 3.60E-14 1.33E-01 7.23E-16 2.68E-03 Zr-97 1.17E-09 4.33E+03 9.02E-15 3.34E-02 1.74E-16 6.44E-04 La-140 1.31E-09 4.85E+03 1.17E-13 4.33E-01 2.16E-15 8.00E-03 La-142 6.84E-11 2.53E+02 1.44E-13 5.33E-01 2.46E-15 9.11E-03 Nd-147 1.85E-09 6.85E+03 6.19E-15 2.29E-02 1.39E-16 5.15E-04 Pr-143 2.19E-09 8.10E+03 2.10E-17 7.77E-05 7.01E-19 2.60E-06 Am-241 1.20E-04 4.44E+08 8.18E-16 3.03E-03 2.75E-17 1.02E-04 Cm-242 4.67E-06 1.73E+07 5.69E-18 2.11E-05 9.56E-19 3.54E-06 Cm-244 6.70E-05 2.48E+08 4.91E-18 1.82E-05 8.78E-19 3.25E-06

• CEDE: Committed Effective Dose Equivalent for inhalation of radioactive materials EDE: Effective Dose Equivalent for cloudshine or submergence in a semi-infinite cloud, or shine from a contaminated ground surface

ULNRC-06636 Page 314 of 374 Insert T15A-5 TABLE 15A-5 (Sheet 2)

Isotope Specific Activity Ci/gm Class 2 Br-83 4.00E-02 Br-84 2.17E-02 Class 32 Rb-86 7.87E-04 Rb-88 1.86E+00 Cs-134 2.31E-01 Cs-136 1.20E-01 Cs-137 1.67E-01 Class 62 Co-58 1.78E-02 Co-60 2.22E-03 Sr-89 3.25E-03 Sr-90 9.26E-05 Sr-91 0.006027 Y-90 1.11E-05 Y-91m 0.003336 Y-91 0.000593 Y-93 0.000315 Zr-95 0.000556 Nb-95 0.000464 Tc-99m 0.4448

ULNRC-06636 Page 315 of 374 TABLE 15A-5 (Sheet 3)

Isotope Specific Activity Ci/gm Ru-103 0.000417 Ru-106 9.26E-05 Rh-103m 0.000417 Rh-106 9.26E-05 Te-125m 0.000269 Te-127m 0.002591 Te-127 0.007873 Te-129m 0.01301 Te-129 0.01479 Te-131m 0.02313 Te-131 0.010197 Te-132 0.2502 Ba-137m 0.147896 Ba-140 0.002035 La-140 0.00139 Ce-141 0.000648 Ce-143 0.00037 Ce-144 0.000306 Pr-143 0.000464 Pr-144 0.000306

ULNRC-06636 Page 316 of 374 Insert Appendix 15B Appendix 15B: Regulatory Guide 1.183, Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at nuclear Power Plants - Conformance Tables Note: In Tables 15B-1 through 15B-7. The text shown in the RG Position columns is taken from Regulatory Guide 1.183. Therefore, references to footnotes, tables, and numbered references may be found in the regulatory guide.

Table 15B-1 Conformance with Regulatory Guide 1.183, Revision 0 Main Sections Table 15B-2 Conformance with Regulatory Guide 1.183, Revision 0, Appendix A (Loss of Coolant Accident)

Table 15B-3 Conformance with Regulatory Guide 1.183, Revision 0, Appendix B (Fuel Handling Accident)

Table 15B-4 Conformance with Regulatory Guide 1.183, Revision 0, Appendix E (PWR Main Steam Line Break Accident)

Table 15B-5 Conformance with Regulatory Guide 1.183, Revision 0, Appendix F (PWR Steam Generator Tube Rupture Accident)

Table 15B-6 Conformance with Regulatory Guide 1.183, Revision 0, Appendix G (PWR Locked Rotor Accident)

Table 15B-7 Conformance with Regulatory Guide 1.183, Revision 0, Appendix H (PWR Rod Ejection Accident)

ULNRC-06636 Page 317 of 374 Table 15B-1 Conformance with Regulatory Guide 1.183, Revision 0 - Main Sections RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section

3. ACCIDENT SOURCE TERM 3.1 Fission Product Inventory Conforms The inventory of fission products in the reactor core and available for release to the containment was based on The inventory of fission products in the reactor core and available for release the maximum full power operation with a core thermal to the containment should be based on the maximum full power operation of power of 3565 MWt. A 2% calorimetric uncertainty is the core with, as a minimum, current licensed values for fuel enrichment, applied as a multiplier on the total core inventory fuel burnup, and an assumed core power equal to the current licensed rated resulting from the ORIGEN runs.

thermal power times the ECCS evaluation uncertainty. The period of irradiation should be of sufficient duration to allow the activity of dose- Core design parameters (enrichment, burnup, and MTU significant radionuclides to reach equilibrium or to reach maximum values. loading) are based on cycles 19-22 to model a bounding The core inventory should be determined using an appropriate isotope cycle.

generation and depletion computer code such as ORIGEN 2 or ORIGEN-ARP. Core inventory factors (Ci/MWt) provided in TIDI 14844 and used in some analysis computer codes were derived for low burnup, low enrichment fuels and should not be used with higher burnup and higher enrichment fuels.

3.1 For the DBA LOCA, all fuel assemblies in the core are assumed to be Conforms For the DBA LOCA, all fuel assemblies were assumed affected and the core average inventory should be used. For DBA events that to be affected and the core average inventory was used.

do not involve the entire core, the fission product inventory of each of the damaged fuel rods is determined by dividing the total core inventory by the A peaking factor of 1.65 was used for DBA events that term based upon full power, core average number of fuel rods in the core. To do not involve the entire core (fuel handling accident, account for differences in power level conditions. The FHA source term is rod ejection, locked rotor), with fission product across the core, radial peaking factors from the facility's core operating inventories for damages fuel rods determined by limits derived from the core source term, the report (COLR) or technical multiplying the total core inventory by the fraction of specifications should be applied in determining number of damaged fuel damaged rods.

rods, and a the inventory of the damaged rods.

3.1 No adjustment to the fission product conservative assembly peaking factor Conforms No adjustments for less than full power were made inventory should be made for events postulated to occur during power which in any analysis.

corresponds to the maximum fuel operations at less than full rated power or those postulated to occur at the rod peaking factor permitted at the beginning For the fuel handling accident, 76-hours of radioactive of core life. decay after shutdown was modeled.

ULNRC-06636 Page 318 of 374 Table 15B-1 Conformance with Regulatory Guide 1.183, Revision 0 - Main Sections RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section 3.2 Release Fractions Conforms For the LOCA event, the core inventory release fractions, by radionuclide groups, for the gap release and The core inventory release fractions, by radionuclide groups, for the gap early in-vessel damage phases in Table 2 were utilized.

release and early in-vessel damage phases for DBA LOCAs are listed in Table 2 for PWRs. These fractions are applied to the equilibrium core inventory described in Regulatory Position 3.1.

Table 2 PWR Core Inventory Fraction Released Into Containment Gap Early Release In-Vessel Phase Phase Total Noble Gases 0.05 0.95 1.0 Halogens 0.05 0.35 0.4 Alkali Metals 0.05 0.25 0.3 Tellurium Metals 0.00 0.05 0.05 Ba, Sr 0.00 0.02 0.02 Noble Metals 0.00 0.0025 0.0025 Cerium Group 0.00 0.0005 0.0005 Lanthanides 0.00 0.0002 0.0002

ULNRC-06636 Page 319 of 374 Table 15B-1 Conformance with Regulatory Guide 1.183, Revision 0 - Main Sections RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section 3.2 For non-LOCA events, the fractions of the core inventory assumed to be in Conforms For non-LOCA events, the fraction of the core inventory the gap for the various radionuclides are given in Table 3. The release assumed to be in the gap by radionuclide group in Table fractions from Table 3 are used in conjunction with the fission product 3 was utilized in conjunction with the maximum core inventory calculated with the maximum core. radial peaking factor of 1.65. The control rod ejection accident was evaluated per Footnote 11 (the gap Table 311 fractions are assumed to be 10% for iodines and noble Non-LOCA Fraction of Fission Product Inventory in Gap gases).

Group Fraction To account for possible damage to an assembly with I-131 0.08 high burnup and rod power and to address Footnote 11, Kr-85 0.10 the fuel handling accident used conservatively high gap Other Noble Gases 0.05 fractions of 12% for I-131, 30% for Kr-85, 10% for all Other Halogens 0.05 other iodines and noble gases, and 17% for Alkali Metals.

Alkali Metals 0.12 These gap fractions were obtained from NUREG/CR-5009.

Footnote 11: The release fractions listed here have been determined to be acceptable for use with currently approved LWR fuel with a peak burnup up to 62,000 MWD/MTU provided that the maximum linear heat generation rate does not exceed 6.3 kw/ft peak rod average power for burnups exceeding 54 GWD/MTU. As an alternative, fission gas release calculations performed using NRC-approved methodologies may be considered on a case-by-case basis. To be acceptable, these calculations must use a projected power history that will bound the limiting projected plant-specific power history for the specific fuel load.

ULNRC-06636 Page 320 of 374 Table 15B-1 Conformance with Regulatory Guide 1.183, Revision 0 - Main Sections RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section 3.3 Table 4 tabulates the onset and duration of each sequential release phase for Conforms The Table 4 PWR onset and durations for the DBA DBA LOCAs at PWRs. The specified onset is the time following the LOCA releases were utilized in the analysis.

initiation of the accident (i.e., time = 0). The early in-vessel phase immediately follows the gap release phase. The activity released from the Note that the gap release was modeled beginning at core during each release phase should be modeled as increasing in a linear 30 seconds and ending in the first half hour in order to fashion over the duration of the phase. For non-LOCA DBAs in which fuel model the early in-vessel release beginning at 0.5 hr.

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

Table 4 LOCA Release Phases PWRs BWRs Phase Onset Duration Onset Duration Gap Release 30 sec 0.5 hr 2 min 0.5 hr Early In-Vessel 0.5 hr 1.3 hr 0.5 hr 1.5 hr 3.3 For facilities licensed with leak-before-break methodology, the onset of Not No additional delays in gap release were assumed for the gap release phase may be assumed to be 10 minutes. A licensee may Applicable the DBA analyses.

propose an alternative time for the onset of the gap release phase, based on facility-specific calculations using suitable analysis codes or on art 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.

ULNRC-06636 Page 321 of 374 Table 15B-1 Conformance with Regulatory Guide 1.183, Revision 0 - Main Sections RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section 3.4 Radionuclide Composition Conforms The Table 5 elements in each radionuclide group were utilized in OBA analyses.

Table 5 lists the elements in each radionuclide group that should be considered in design basis analyses. Note that since RADTRAD is limited to modeling 63 nuclides, certain nuclides which were deemed to be Table 5 insignificant from a dose perspective were not included.

Radionuclide Groups Group Elements Noble Gases Xe, Kr Halogens I, Br Alkali Metals Cs, Rb Tellurium Group Te, Sb, Se, Ba, Sr Noble Metals Ru, Rh, Pd, Mo, Tc, Co Lanthanides La, Zr, Nd, Eu, Nb, Pm, Pr Sm, Y, Cm, Am Cerium Ce, Pu, Np 3.5 Chemical Form Conforms For releases from the reactor coolant system (RCS) to the containment, 95% of the iodine released was Of the radioiodine released from the reactor coolant system (RCS) to the assumed to be cesium iodide (CsI), 4.85% elemental containment in a postulated accident, 95% of the iodine released should be iodine, and 0.15% organic iodide.

assumed to be cesium iodide (CsI), 4.85% elemental iodine, and 0.15%

organic iodide. This includes releases from the gap and the fuel pellets. With Fission products were assumed to be in particulate form the exception of elemental and organic iodine and noble gases, fission with the exception of elemental and organic iodine and products should be assumed to be in particulate form. The same chemical noble gases, form is assumed in releases from fuel pins in FHAs and from releases from the fuel pins through the RCS in DBAs other than FHAs or LOCAs.

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

ULNRC-06636 Page 322 of 374 Table 15B-1 Conformance with Regulatory Guide 1.183, Revision 0 - Main Sections RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section 3.6 Fuel Damage in Non-LOCA DBAs Conforms The amount of fuel damage caused by non-LOCA design basis events was analyzed. The conservatively calculated The amount of fuel damage caused by non-LOCA design basis events values were reflected in the rod ejection and locked rotor should be analyzed to determine, for the case resulting in the highest DBA analyses.

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

4. DOSE CALCULATION METHODOLOGY 4.1 Offsite Dose Consequences 4.1.1 The dose calculations should determine the TEDE. TEDE is the sum of the Conforms The dose calculations determined the TEDE and committed effective dose equivalent (CEDE) from inhalation and the deep consider all radionuclides that are significant with regard dose equivalent (DDE) from external exposure. The calculation of these two to dose consequences.

components of the TEDE should consider all radionuclides, including progeny from the decay of parent radionuclides that are significant with Progeny was not included in the dose calculations regard to dose consequences and the released radioactivity. consistent with previously approved submittals, including:

Point Beach Units 1 & 2 - April 2011 (ADAMS Accession Number MLl 10240054)

Arkansas Nuclear One, Unit 2 - April 2011 (ADAMS Accession Number ML l 10980197) 4.1.2 The exposure-to-CEDE factors for inhalation of radioactive material should Conforms CEDE Conversion factors for isotopes were taken from be derived from the data provided in ICRP Publication 30, Limits for Table 2.1 of Federal Guidance Report 11, "Limiting Intakes of Radionuclides by Workers (Reference 19). Table 2.1 of Federal Values of Radionuclide Intake and Air Concentration Guidance Report 11, Limiting Values of Radionuclide Intake and Air and Dose Conversion Factors for Inhalation, Concentration and Dose Conversion Factors for Inhalation, Submersion, and Submersion, and Ingestion."

Ingestion (Reference 20), provides tables of conversion factors acceptable to the NRC staff. The factors in the column headed effective yield doses corresponding to the CEDE.

ULNRC-06636 Page 323 of 374 Table 15B-1 Conformance with Regulatory Guide 1.183, Revision 0 - Main Sections RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section 4.1.3 For the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the breathing rate of persons offsite should be assumed Conforms The breathing rates provided were utilized to calculate to be 3.5 x 10-4 cubic meters per second. From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the the offsite dose consequences. For detem1ining a accident, the breathing rate should be assumed to be 1.8 x 10-4 cubic meters limiting 2-hour EAB dose, a constant breathing rate of per second. After that and until the end of the accident, the rate should be 3.5 x 10-4 cubic meters per second was used.

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

4.1.4 The DDE should be calculated assuming submergence in semi-infinite cloud Conforms EDE Conversion factors for isotopes were taken from assumptions with appropriate credit for attenuation by body tissue. The DDE Table III.1 of Federal Guidance Report 12, External is nominally equivalent to the effective dose equivalent (EDE) from external Exposure to Radionuclides in Air, Water, and Soil."

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

4.15 The TEDE should be determined for the most limiting person at the EAB. Conforms The TEDE was determined for the most limiting person The maximum EAB TEDE for any two-hour period following the start of at the EAB. The maximum two-hour TEDE was the radioactivity release should be determined and used in determining determined by calculating the postulated dose for a compliance with the dose criteria in 10 CFR 50.67. The maximum two-hour series of small time increments and performing a TEDE should be determined by calculating the postulated dose for a series "sliding" sum over the increments for successive two-of small time increments and performing a sliding sum over the hour periods. This was performed by the RADTRAD increments for successive two-hour periods. The maximum TEDE obtained computer code with constant inputs for atmospheric is submitted. The time increments should appropriately reflect the dispersion factors and breathing rates.

progression of the accident to capture the peak dose interval between the start of the event and the end of radioactivity release (see also Table 6).

4.16 TEDE should be determined for the most limiting receptor at the outer Conforms The TEDE was determined for the most limiting receptor boundary of the low population zone (LPZ) and should be used in at the outer boundary of the low population zone (LPZ).

determining compliance with the dose criteria in 10 CFR 50.67.

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

ULNRC-06636 Page 324 of 374 Table 15B-1 Conformance with Regulatory Guide 1.183, Revision 0 - Main Sections RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section 4.2 Control Room Dose Consequences 4.2.1 The TEDE analysis should consider all sources of radiation that will cause Conforms The TEDE analysis considered all significant sources of exposure to control room personnel. The applicable sources will vary from radiation that would cause exposure to Control Room facility to facility, but typically will include: personnel. For Callaway, the limiting Control Room

• Contamination of the control room atmosphere by the intake or dose included:

infiltration of the radioactive material contained in the radioactive

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

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

• Contamination of the control room atmosphere by

• Radiation shine from the external radioactive plume released from the intake or infiltration of airborne radioactive the facility, material from the Control Building,

• Radiation shine from radioactive material in the reactor

• Radiation shine from the external radioactive plume containment, released from the facility,

• Radiation shine from radioactive material in systems and

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

• Radiation shine from radioactive material in Control Room recirculation filters and radioactive material in the Control Building.

• Radiation shine from radioactive material deposited on the ground which operators encounter in transit to and from the Control Room.

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

4.2.3 The models used to transport radioactive material into and through the Conforms The models used to transport radioactive material into control room, and the shielding models used to determine radiation dose and through the Control Room, and the shielding models rates from external sources, should be structured to provide suitably used to deternine radiation dose rates from external conservative estimates of the exposure to control room personnel. sources, were developed to provide suitably conservative estimates of the exposure to Control Room personnel.

ULNRC-06636 Page 325 of 374 Table 15B-1 Conformance with Regulatory Guide 1.183, Revision 0 - Main Sections RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section 4.2.4 Credit for engineered safety features that mitigate airborne radioactive Conforms Credit for engineered safety features that mitigate material within the control room may be assumed. Such features may airborne radioactive material within the Control Room include control room isolation or pressurization, or intake or recirculation and Control Building were assumed as appropriate. Note filtration. Refer to Section 6.5.1, ESF Atmospheric Cleanup System, of the that no credit for Control Room isolation was modeled SRP (Reference 3) and Regulatory Guide 1.52, Design, Testing, and for events that rely solely on radiation monitors.

Maintenance Criteria for Postaccident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants (Reference 25), for guidance. The control room design is often optimized for the DBA LOCA and the protection afforded for other accident sequences may not be as advantageous. In most designs, control room isolation is actuated by engineered safeguards feature (ESF) signals or radiation monitors (RMs). In some cases, the ESF signal is effective only for selected accidents, placing reliance on the RMs for the remaining accidents. Several aspects of RMs can delay the control room isolation, including the delay for activity to build up to concentrations equivalent to the alarm setpoint and the effects of different radionuclide accident isotopic mixes on monitor response.

4.2.5 Credit should generally not be taken for the use of personal protective Conforms Credit was not taken for the use of personnel protective equipment or prophylactic drugs. Deviations may be considered on a case- equipment or prophylactic drugs.

by-case basis.

4.2.6 The dose receptor for these analyses is the hypothetical maximum exposed Conforms The occupancy factors and breathing rate were utilized individual who is present in the control room for 100% of the time during to determine the doses to the hypothetical maximum the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the event, 60% of the time between 1 and 4 days, and exposed individual who is present in the Control Room.

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 10-4 cubic Control Room X/Q values were determined utilizing the meters per second ARCON96 computer code which does not incorporate occupancy factors. Occupancy factors were included in the RADTRAD computer code for the dose evaluations.

ULNRC-06636 Page 326 of 374 Table 15B-1 Conformance with Regulatory Guide 1.183, Revision 0 - Main Sections RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section 4.2.7 Control room doses should be calculated using dose conversion factors Conforms The DDE from photons was corrected for the difference identified in Regulatory Position 4.1 above for use in offsite dose analyses. between finite cloud geometry in the Control Room and The DDE from photons may be corrected for the difference between finite the semi-infinite cloud assumption used in calculating cloud geometry in the control room and the semi-infinite cloud assumption the dose conversion factors by the given equation. This used in calculating the dose conversion factors. The following expression correction was performed by the RADTRAD computer may be used to correct the semi-infinite cloud dose, DDE, to a finite cloud code.

dose, DDEfinite, where the control room is modeled as a hemisphere that has a volume, V, in cubic feet, equivalent to that of the control room (Reference 22).

4.3 Other Dose Consequences Conforms Exception - The current TID-14844 accident source term will remain the licensing basis for equipment The guidance provided in Regulatory Positions 4.1 and 4.2 should be used, qualification and NUREG-0737 evaluations other than as applicable, in re-assessing the radiological analyses identified in Control Room and Technical Support Center doses.

Regulatory Position 1.3.1, such as those in NUREG-0737 (Reference 2).

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

4.4 Acceptance Criteria Conforms The DBAs were updated for consistency with the TEDE criterion in Table 6 for offsite doses and in 10 CFR The radiological criteria for the EAB, the outer boundary of the LPZ, and for 50.67(b)(2)(iii) for the Control Room and Technical the control room are in 10 CFR 50.67. These criteria are stated for Support Center doses.

evaluating reactor accidents of exceedingly low probability of occurrence and low risk of public exposure to radiation, e.g., a large-break LOCA. The control room criterion applies to all accidents. For events with a higher probability of occurrence, postulated EAB and LPZ doses should not exceed the criteria tabulated in Table 6. The acceptance criteria for the various NUREG-0737 (Reference 2) items generally reference General Design

ULNRC-06636 Page 327 of 374 Table 15B-1 Conformance with Regulatory Guide 1.183, Revision 0 - Main Sections RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section Criteria (GDC 19) from Appendix A to 10 CFR Part 50 or specify criteria derived from GDC-19. These criteria are generally specified in terms of whole body dose, or its equivalent to any body organ. For facilities applying for, or having received, approval for the use of AST, the applicable criteria should be updated for consistency with the TEDE criterion in 10 CFR 50.67(b)(2)(iii).

5. ANALYSIS ASSUMPTIONS AND METHODOLOGY 5.1 General Considerations 5.1.1 The evaluations required by 10 CFR 50.67 are re-analyses of the design Conforms The OBA analyses were prepared, reviewed, and basis safety analyses and evaluations required by 10 CFR 50.34; they are maintained per 10 CFR 50 Appendix B and the guidance considered to be a significant input to the evaluations required by 10 CFR consistent with RG 1 .183.

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.

ULNRC-06636 Page 328 of 374 Table 15B-1 Conformance with Regulatory Guide 1.183, Revision 0 - Main Sections RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section 5.1.2 Credit may be taken for accident mitigation features that are classified as Conforms Credit was taken for Engineered Safeguard Features with safety-related, are required to be operable by technical specifications, are failure assumptions to maximize the calculated doses.

powered by emergency power sources, and are either automatically actuated Assumptions regarding the occurrence and timing of a or, in limited cases, have actuation requirements explicitly addressed in loss of offsite power were also selected with the emergency operating procedures. The single active component failure that objective of maximizing the postulated radiological results in the most limiting radiological consequences should be assumed. consequences.

Assumptions regarding the occurrence and timing of a loss of offsite power should be selected with the objective of maximizing the postulated radiological consequences.

5.1.3 The numeric values that are chosen as inputs to the analyses required by 10 Conforms The numeric values that were chosen as inputs to the CFR 50.67 should be selected with the objective of determining a analyses required by 10 CFR 50.67 were selected with conservative postulated dose. In some instances, a particular parameter may the objective of determining a conservative postulated be conservative in one portion of an analysis but be nonconservative in dose.

another portion of the same analysis. For a range of values, the value that resulted in a conservative postulated dose was used.

5.1.4 Licensees should ensure that analysis assumptions and methods are Conforms Licensee has ensured that analysis assumptions and compatible with the ASTs and the TEDE criteria. methods are compatible with the AST and the TEDE criteria.

5.2 Accident-Specific Assumptions Conforms Licensee has analyzed the DBAs that are affected by the specific proposed applications of an AST.

The appendices to this regulatory guide provide accident-specific assumptions that are acceptable to the staff for performing analyses that are required by 10 CFR 50.67. The DBAs addressed in these attachments were selected from accidents that may involve damage to irradiated fuel. This guide does not address DBAs with radiological consequences based on technical specification reactor or secondary coolant-specific activities only.

The inclusion or exclusion of a particular DBA in this guide should not be interpreted as indicating that an analysis of that DBA is required or not required. Licensees should analyze the DBAs that are affected by the specific proposed applications of an AST.

ULNRC-06636 Page 329 of 374 Table 15B-1 Conformance with Regulatory Guide 1.183, Revision 0 - Main Sections RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section 5.3 Meteorological Assumptions Conforms The re-calculation of atmospheric dispersion factors was performed for the EAB and LPZ using the NRC Atmospheric dispersion values (X/Q) for the EAB, the LPZ, and the control computer code PAVAN according to the guidance of room that were approved by the staff during initial facility licensing or in RG 1.145 and for the control room and TSC intakes with subsequent licensing proceedings may be used in performing the new release points using the NRC computer code radiological analyses identified by this guide. Methodologies that have been ARCON96 according to the guidance of RG 1.194.

used for determining x/Q values are documented in Regulatory Guides 1.3 and 1.4, Regulatory Guide 1.145, "Atmospheric Dispersion Models for The meteorological data used in the calculation were Potential Accident Consequence Assessments at Nuclear Power Plants," and collected in accordance with Callaway site-specific the paper, "Nuclear Power Plant Control Room Ventilation System Design measurements program and RG 1.23.

for Meeting General Criterion 19"(Refs. 6, 7, 22, and 28).

References 22 and 28 should be used if the FSAR X/Q values are to be revised or if values are to be determined for new release points or receptor distances. Fumigation should be considered where applicable for the EAB and LPZ. For the EAB, the assumed fumigation period should be timed to be included in the worst 2-hour exposure period. The NRC computer code PAVAN (Ref. 29) implements Regulatory Guide 1.145 (Ref. 28) and its use is acceptable to the NRC staff. The methodology of the NRC computer code ARCON96 (Ref. 26) is generally acceptable to the NRC staff for use in determining control room X/Q values. Meteorological data collected in accordance with the site-specific meteorological measurements program described in the facility FSAR should be used in generating accident X/Q values. Additional guidance is provided in Regulatory Guide 1.23, "Onsite Meteorological Programs" (Ref. 30). All changes in X/Q analysis methodology should be reviewed by the NRC staff.

6. Assumptions for Evaluating the Radiation Doses for Equipment Not applicable An AST assessment was not performed for equipment Qualification qualification. The TID-14844 assumptions will continue to be used as the radiation dose basis for equipment qualification, radiation zone maps, and shielding calculations.

ULNRC-06636 Enclosure 6 Page 330 of 374 Table 15B-2 Conformance with Regulatory Guide 1.183, Revision 0 Appendix A (Loss-of-Coolant Accident)

RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section Appendix Assumptions for Evaluating the Radiological Consequences of a A LWR Loss-of-Coolant Accident Source Term The inventory of fission products in the reactor core and available Appendix Acceptable assumptions regarding core inventory and the release of Conforms for release to the containment was based on the maximum full A1 radionuclides from the fuel are provided in Regulatory Position 3 of power operation with a core thermal power of 3636 MWt (102%

this guide.

of 3565 MWt nominal power).

Core design parameters (enrichment, burnup, and MTU loading) are based on the cycles 19 through 22 with conservative increases in enrichment and burnup. Margin is added to the EOC core inventory, calculated with ORIGEN-S, to account for potential core design differences in future cycles. For the DBA LOCA, all fuel assemblies were assumed to be affected and a conservatively bounding core inventory was used.

Appendix If the sump or suppression pool pH is controlled at values of 7 or Conforms The equilibrium pH in the sump stays above 7.

A2 greater, the chemical form of radioiodine released to the containment should be assumed to be 95% cesium iodide (CsI),

4.85% elemental iodine, and 0.15% organic iodide. Iodine species, including those from iodine re-evolution, for sump or suppression pool pH values less than 7 will be evaluated on a case-by-case basis. Evaluations of pH should consider the effect of acids and bases created during the LOCA event, e.g., radiolysis products.

With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form.

Appendix The radioactivity released from the fuel should be assumed to mix Conforms Based on relative volumes, the release from the fuel is split A 3.1 instantaneously and homogeneously throughout the free air volume between the sprayed and unsprayed regions in containment.

of the primary containment in PWRs as it is released. This While operation of 2 of 4 containment air coolers promotes distribution should be adjusted if there are internal compartments mixing between the two regions, the exchange rate is that have limited ventilation exchange. .... The release into the conservatively limited to two turnovers of the unsprayed region containment should be assumed to terminate at the end of the per hour. This is in accordance with section A 3.3 (below).

early in-vessel phase. The release to containment is assumed to terminate at the end of the early in-vessel phase.

ULNRC-06636 Enclosure 6 Page 331 of 374 Table 15B-2 Conformance with Regulatory Guide 1.183, Revision 0 Appendix A (Loss-of-Coolant Accident)

RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section Appendix Reduction in airborne radioactivity in the containment by natural Conforms Natural deposition is not credited in this analysis.

A 3.2 deposition within the containment may be credited. Acceptable Spray removal coefficients are calculated in accordance with models for removal of iodine and aerosols are described in Chapter Chapter 6.5.2 of the SRP.

6.5.2, Containment Spray as a Fission Product Cleanup System, of the Standard Review Plan (SRP), NUREG-0800 (Reference A-1) and in NUREG/CR-6189, A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments (Reference A-2). The latter model is incorporated into the analysis code RADTRAD (Reference A-3). The prior practice of deterministically assuming that a 50% plateout of iodine is released from the fuel is no longer acceptable to the NRC staff as it is inconsistent with the characteristics of the revised source terms.

ULNRC-06636 Enclosure 6 Page 332 of 374 Table 15B-2 Conformance with Regulatory Guide 1.183, Revision 0 Appendix A (Loss-of-Coolant Accident)

RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section Appendix Reduction in airborne radioactivity in the containment by In accordance with Position 5.1.2, Containment spray is A 3.3 containment spray systems that have been designed and are Conforms - an ESF system, maintained in accordance with Chapter 6.5.2 of the SRP may be

- classified as safety related, credited. Acceptable models for the removal of iodine and aerosols

- required to be operable by technical specifications are described in Chapter 6.5.2 of the SRP and NUREG/CR-5966.

- powered by emergency power sources, and This simplified model is incorporated into the analysis code

- automatically actuated.

RADTRAD.

The evaluation of the containment sprays should address areas The mixing rate between sprayed and unsprayed regions in within the primary containment that are not covered by the spray containment is conservatively assumed to be two turnovers of the drops. The mixing rate attributed to natural convection between unsprayed regions per hour.

sprayed and unsprayed regions of the containment building, provided that adequate flow exists between these regions, is assumed to be two turnovers of the unsprayed regions per hour, The spray removal coefficient for particulate iodine is reduced by unless other rates are justified. The containment building a factor of 10 when a DF of 50 is reached.

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 (DF) for elemental iodine based on the maximum iodine activity in the primary containment atmosphere when the sprays actuate, divided by the activity of iodine remaining at some time after decontamination. The SRP also states that the particulate iodine removal rate should be reduced by a factor of 10 when a DF of 50 is reached. The reduction in the removal rate is not required if the removal rate is based on the calculated time-dependent airborne aerosol mass.

There is no specified maximum DF for aerosol removal by sprays.

Appendix Reduction in airborne radioactivity in the containment by in- Not used.

A 3.4 containment recirculation filter systems may be credited if these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. A-5 and A-6). The filter media loading caused by the increased aerosol release associated with the revised source term should be addressed.

ULNRC-06636 Enclosure 6 Page 333 of 374 Table 15B-2 Conformance with Regulatory Guide 1.183, Revision 0 Appendix A (Loss-of-Coolant Accident)

RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section Appendix Reduction in airborne radioactivity in the containment by Not A 3.5 suppression pool scrubbing in BWRs should generally not be Applicable credited. However, the staff may consider such reduction on an individual case basis. The evaluation should consider the relative timing of the blowdown and the fission product release from the fuel, the force driving the release through the pool, and the potential for any bypass of the suppression pool (Reference 7). Analyses should consider iodine re-evolution if the suppression pool liquid pH is not maintained greater than 7.

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

Appendix The primary containment should be assumed to leak at the peak Conforms From Technical Specification 5.5.16.c., the maximum allowable A 3.7 pressure technical specification leak rate for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. For containment leakage rate, La, at Pa, shall be 0.20% of the PWRs, the leak rate may be reduced after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 50% containment air weight per day.

of the technical specification leak rate. .. . Leakage from After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, this is reduced to 0.10% per day.

subatmospheric containments is assumed to terminate when the containment is brought to and maintained at a subatmospheric condition as defined by technical specifications.

Appendix If the primary containment is routinely purged during power Conforms. Only the Containment Mini-purge may be in use during power A 3.8 operations, releases via the purge system prior to containment operation.

isolation should be analyzed and the resulting doses summed with 100% of the RCS maximum equilibrium activity is released to the postulated doses from other release paths. The purge release containment at initiation of the LOCA.

evaluation should assume that 100% of the radionuclide inventory in the reactor coolant system liquid is released to the containment at The mini-purge isolation valves automatically close within 11 the initiation of the LOCA. This inventory should be based on the seconds, well before the onset of the gap release.

technical specification reactor coolant system equilibrium activity.

Iodine spikes need not be considered. If the purge system is not isolated before the onset of the gap release phase, the release fractions associated with the gap release and early in-vessel phases should be considered as applicable.

ULNRC-06636 Enclosure 6 Page 334 of 374 Table 15B-2 Conformance with Regulatory Guide 1.183, Revision 0 Appendix A (Loss-of-Coolant Accident)

RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section Appendix Assumptions on ESF System Leakage A 5.0 Appendix With the exception of noble gases, all the fission products released Conforms In combination with item A 5.3 below, only iodine is released to A 5.1 from the fuel to the containment (as defined in Tables 1 and 2 of the environment from the ESF system leakage.

this guide) should be assumed to instantaneously and homogeneously mix in the primary containment sump water (in PWRs) at the time of release from the core. In lieu of this deterministic approach, suitably conservative mechanistic models for the transport of airborne activity in containment to the sump water may be used. Note that many of the parameters that make spray and deposition models conservative with regard to containment airborne leakage are non-conservative with regard to the buildup of sump activity.

Appendix The leakage should be taken as two times the sum of the Conforms The operational limit of 1 gpm is doubled to 2 gpm as the basis for A 5.2 simultaneous leakage from all components in the ESF recirculation ECCS leakage to the Aux. Building. Instead of waiting the full systems above which the technical specifications, or licensee 11.8 minutes as the earliest time to begin recirculation, commitments to item III.D.1.1 of NUREG-0737 (Reference A-8), recirculation is conservatively assumed to start just after control would require declaring such systems inoperable. The leakage room isolation at 62 seconds.

should be assumed to start at the earliest time the recirculation flow occurs in these systems and end at the latest time the releases from these systems are terminated. Consideration should also be given to design leakage through valves isolating ESF recirculation Isolation valve seat leakage to the RWST is analyzed as a separate systems from tanks vented to atmosphere, e.g., emergency core case with a total of 4 gpm of back-leakage to the RWST (3 gpm cooling system (ECCS) pump miniflow return to the refueling below water line, 1 gpm above water line).

water storage tank.

Appendix With the exception of iodine, all radioactive materials in the Conforms The release from leakage to the environment is limited to iodine.

A 5.3 recirculating liquid should be assumed to be retained in the liquid phase.

ULNRC-06636 Enclosure 6 Page 335 of 374 Table 15B-2 Conformance with Regulatory Guide 1.183, Revision 0 Appendix A (Loss-of-Coolant Accident)

RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section Appendix If the temperature of the leakage exceeds 212°F, the fraction of Conforms With a maximum sump temperature of approximately 265F after A 5.4 total iodine in the liquid that becomes airborne should be assumed the beginning of recirculation, the calculated flashing fraction is equal to the fraction of the leakage that flashes to vapor. This flash less than 10%.

fraction, FF, should be determined using a constant enthalpy, h, process, based on the maximum time-dependent temperature of the sump water circulating outside the containment:

where, hf1 is the enthalpy of liquid at system design temperature and pressure; hf2 is the enthalpy of liquid at saturation conditions (14.7 psia, 212ºF); and hfg is the heat of vaporization at 212ºF.

Appendix If the temperature of the leakage is less than 212°F or the calculated Conforms The analysis assumes that 10% of the iodine activity in the leakage A 5.5 flash fraction is less than 10%, the amount of iodine that becomes becomes airborne and is available for filtration by the Aux.

airborne should be assumed to be 10% of the total iodine activity in Building vent/exhaust system.

the leaked fluid, unless a smaller amount can be justified based on The back-leakage into the RWST represents a much more the actual sump pH history and area ventilation rates.

controlled and well-defined environment that allows the ultimate release from the leakage to be more directly evaluated. As a very conservative treatment of RWST liquid, the analysis assumes the RG 1.183 conservative airborne release of 10% of the liquid activity for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the event to cover any potential flashing or elemental iodine regeneration within the piping. After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a very conservative airborne release of 8% of the iodine is assumed despite the calculated flashing fraction of 5.5% (based on sump saturated liquid enthalpy). In accordance with NUREG/CR-5950, the temperature and pH dependent iodine re-evolution from the liquid space inside the tank and incoming sump leakage is conservatively bounded by an 8% release of sump iodine directly to the airspace. Additionally, the vent at the top of the tank restricts the ventilation rate of the vapor space.

ULNRC-06636 Enclosure 6 Page 336 of 374 Table 15B-2 Conformance with Regulatory Guide 1.183, Revision 0 Appendix A (Loss-of-Coolant Accident)

RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section Appendix The radioiodine that is postulated to be available for release to the Conforms 97% elemental and 3% organic iodine was specified as the A 5.6 environment is assumed to be 97% elemental and 3% organic. appropriate chemical form.

Reduction in release activity by dilution or holdup within buildings, or by ESF ventilation filtration systems, may be credited where applicable. Filter systems used in these applications should be Compliance with RG 1.52 is presented in FSAR Table 9.4-2.

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

Appendix Assumption on Containment Purging Limited The containment is not purged post-LOCA, but because use of the A 7.0 applicability Mini-purge system is allowed during normal operation and The radiological consequences from post-LOCA primary because of the possibility of the accident occurring at this time, containment purging as a combustible gas or pressure control operation of Mini-purge is analyzed as a possible contribution to measure should be analyzed. If the installed containment purging the LOCA dose. See item A 3.8.

capabilities are maintained for purposes of severe accident management and are not credited in any design basis analysis, radiological consequences need not be evaluated. If the primary containment purging is required within 30 days of the LOCA, the results of this analysis should be combined with consequences postulated for other fission product release paths to determine the No credit is taken for the filtration portion of the Mini-purge total calculated radiological consequences from the LOCA.

system because these filters are not classified as safety related.

Reduction in the amount of radioactive material released via ESF filter systems may be taken into account provided that these systems meet the guidance in Regulatory Guide 1.52 (Reference A-

5) and Generic Letter 99-02 (Reference A-6). Compliance with Reg. Guide 1.52 for (other) ESF filters is presented in FSAR Table 9.4-2.

ULNRC-06636 Enclosure 6 Page 337 of 374 Table 15B-3 Conformance with Regulatory Guide 1.183, Revision 0 Appendix B (Fuel Handling Accident)

RG Section Regulatory Guide 1.183 Position Analysis Basis of Conformance Appendix Assumptions for Evaluating the Radiological Consequences of a B

Fuel Handling Accident Source Term Appendix Acceptable assumptions regarding core inventory and the release of The inventory of fission products in the reactor core and Conforms B1 radionuclides from the fuel are provided in Regulatory Position 3 of available for release to the containment was based on the this guide. maximum full power operation with a core thermal power of 3637 MWt (102% of 3565 MWt nominal power). Core design parameters (enrichment, burnup, and MTU loading) are based on the cycles 19 through 22 with conservative increases in enrichment and burnup. Margin is added to the EOC core inventory, calculated with ORIGEN-S, to account for potential core design differences in future cycles.

ULNRC-06636 Enclosure 6 Page 338 of 374 Table 15B-3 Conformance with Regulatory Guide 1.183, Revision 0 Appendix B (Fuel Handling Accident)

RG Section Regulatory Guide 1.183 Position Analysis Basis of Conformance Appendix The number of fuel rods damaged during the accident should be based The number of damaged fuel rods is carried over from the Conforms B 1.1 on a conservative analysis that considers the most limiting case. This current design basis FHA analyses.

based upon analysis should consider parameters such as the weight of the dropped consideratio heavy load or the weight of a dropped fuel assembly (plus any For the FHA in containment, 1.2 fuel assemblies were assumed n of Section attached handling grapples), the height of the drop, and the to be affected and a conservatively bounding source term was 5.1.4 as compression, torsion, and shear stresses on the irradiated fuel rods. used.

discussed in Damage to adjacent fuel assemblies, if applicable (e.g., events over Basis of the reactor vessel), should be considered. For the FHA in the Fuel Handling Building, 1.0 fuel assemblies Compliance were assumed to be affected and a conservatively bounding source term was used.

Per Reg. Guide 1.183, Section 5.1.4 Applicability of Prior Licensing Basis, the prior FHA design basis may continue as the facilitys design basis if it is unrelated to the use of the AST or unaffected by the AST. The prior design basis for the number of damaged fuel rods is not directly affected by implementation of the AST and is compatible with the characteristics and the revised dose calculation methodology of the AST or the TEDE criteria.

Additionally, retention of the facilitys design basis for this FHA analysis does not introduce any assumption that is inconsistent with the internally consistent assumptions that comprise the AST methodology as specified in Section 5.2 of Reg Guide 1.183.

Use of this prior licensing/design basis fuel failure value is in accordance with Section 5.1.4 of the main body of Reg. Guide 1.183 and therefore remains applicable as the basis for AST.

Appendix The fission product release from the breached fuel is based on Conforms The release fractions specified in Regulatory Position 3.2 are B 1.2 Regulatory Position 3.2 of this guide and the estimate of the number applied to damaged fuel that meets the requirements specified.

of fuel rods breached. All the gap activity in the damaged rods is For high burnup fuel, higher release fractions from NUREG/CR-assumed to be instantaneously released. Radionuclides that should be 5009 and Reg. Guide 1.25 are applied.

considered include xenons, kryptons, halogens, cesiums, and rubidiums.

ULNRC-06636 Enclosure 6 Page 339 of 374 Table 15B-3 Conformance with Regulatory Guide 1.183, Revision 0 Appendix B (Fuel Handling Accident)

RG Section Regulatory Guide 1.183 Position Analysis Basis of Conformance Appendix The chemical form of radioiodine released from the fuel to the spent Conforms In accordance with the specified instant and complete B 1.3 fuel pool should be assumed to be 95% cesium iodide (CsI), 4.85 dissociation and re-evolution of the CsI, the effective chemical percent elemental iodine, and 0.15 percent organic iodide. The CsI form specified for the radioiodine is 99.85 percent elemental released from the fuel is assumed to completely dissociate in the pool (95% CsI + 4.85% elemental) and 0.15 percent organic.

water. Because of the low pH of the pool water, the iodine re-evolves as elemental iodine. This is assumed to occur instantaneously. The NRC staff will consider, on a case-by-case basis, justifiable mechanistic treatment of the iodine release from the pool.

Water Depth Appendix If the depth of water above the damaged fuel is 23 feet or greater, the A minimum of 23 of water is required by Technical B2 decontamination factors for the elemental and organic species are 500 Specifications during fuel movement. As noted in Item 8 and 1, respectively, giving an overall effective decontamination factor Conforms ofNRC Regulatory Isssue Summary (RIS) 2006-04, Experience of 200 (i.e., 99.5% of the total iodine released from the damaged rods with Implementation of Alternative Source Terms, the is retained by the water). This difference in decontamination factors decontamination factor discussion in Reg. Guide 1.183 is for elemental (99.85%) and organic iodine (0.15%) species results in misleading; the elemental iodine decontamination factor to be the iodine above the water being composed of 57% elemental and used is 285. The organic decontamination factor is 1.0. As 43% organic species. If the depth of water is not 23 feet, the described, this gives an overall effective decontamination factor decontamination factor will have to be determined on a case-by-case of 200.

method (Ref. B-1).

Noble Gases Appendix The retention of noble gases in the water in the fuel pool or reactor Conforms No holdup or scrubbing of noble gases is credited.

B3 cavity is negligible (i.e., decontamination factor of 1). Particulate radionuclides are assumed to be retained by the water in the fuel pool or reactor cavity (i.e., infinite decontamination factor).

Fuel Handling Accidents within the Fuel Building Appendix The radioactive material that escapes from the fuel pool to the fuel Conforms A two hour release period is specified for activity escaping the B 4.1 building is assumed to be released to the environment over a 2-hour fuel pool. Holdup and dilution within the FHB are not credited.

time period.

ULNRC-06636 Enclosure 6 Page 340 of 374 Table 15B-3 Conformance with Regulatory Guide 1.183, Revision 0 Appendix B (Fuel Handling Accident)

RG Section Regulatory Guide 1.183 Position Analysis Basis of Conformance Appendix A reduction in the amount of radioactive material released from the Not Filtration of releases from the FHB are not credited.

B 4.2 fuel pool by engineered safety feature (ESF) filter systems may be Applicable taken into account provided these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. B-2, B-3).

Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system1 should be determined and accounted for in the radioactivity release analyses.

Appendix The radioactivity release from the fuel pool should be assumed to be Conforms Holdup and dilution in the FHB are not credited.

B 4.3 drawn into the ESF filtration system without mixing or dilution in the fuel building. If mixing can be demonstrated, credit for mixing and dilution may be considered on a case-by-case basis. This evaluation should consider the magnitude of the building volume and exhaust rate, the potential for bypass to the environment, the location of exhaust plenums relative to the surface of the pool, recirculation ventilation systems, and internal walls and floors that impede stream flow between the surface of the pool and the exhaust plenums.

Fuel Handling Accidents within Containment Appendix If the containment is isolated during fuel handling operations, no Not During refueling, open containment penetrations are allowed B 5.1 radiological consequences need to be analyzed. Applicable (under administrative controls).

Appendix If the containment is open during fuel handling operations, but Not Automatic isolation of the containment is not credited.

B 5.2 designed to automatically isolate in the event of a fuel handling Applicable accident, the release duration should be based on delays in radiation detection and completion of containment isolation. If it can be shown that containment isolation occurs before radioactivity is released to the environment,1 no radiological consequences need to be analyzed.

Appendix If the containment is open during fuel handling operations (e.g.,

Conforms A two hour release period is specified for the activity escaping B 5.3 personnel air lock or equipment hatch is open), the radioactive the cavity pool.

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

ULNRC-06636 Enclosure 6 Page 341 of 374 Table 15B-3 Conformance with Regulatory Guide 1.183, Revision 0 Appendix B (Fuel Handling Accident)

RG Section Regulatory Guide 1.183 Position Analysis Basis of Conformance Appendix A reduction in the amount of radioactive material released from the Not ESF filtration in the containment is not credited.

B 5.4 containment by ESF filter systems may be taken into account Applicable provided that these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. B-2 and B-3). Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system should be determined and accounted for in the radioactivity release analyses.

Appendix Credit for dilution or mixing of the activity released from the reactor Not Dilution and mixing within the containment are not credited.

B 5.5 cavity by natural or forced convection inside the containment may be Applicable considered on a case-by-case basis. Such credit is generally limited to 50% of the containment free volume. This evaluation should consider the magnitude of the containment volume and exhaust rate, the potential for bypass to the environment, the location of exhaust plenums relative to the surface of the reactor cavity, recirculation ventilation systems, and internal walls and floors that impede stream flow between the surface of the reactor cavity and the exhaust plenums.

ULNRC-06636 Enclosure 6 Page 342 of 374 Table 15B-4 Conformance with Regulatory Guide 1.183, Revision 0 Appendix E (PWR Main Steam Line Break)

RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section Appendix Assumptions for Evaluating the Radiological Consequences of a PWR Main Steam Line Break Accident E

Source Term Appendix Assumptions acceptable to the NRC staff regarding core inventory Not Assuming that the highest worth control rod is stuck at its fully E1 and the release of radionuclides from the fuel are provided in Applicable withdrawn position, no fuel damage was postulated to occur during Regulatory Position 3 of this regulatory guide. The release from the the MSLB.

breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached. The fuel damage estimate should assume that the highest worth control rod is stuck at its fully withdrawn position.

Appendix If no or minimal fuel damage is postulated for the limiting event, Conforms The initial RCS coolant activity is such that the DE I-133 and DE E2 the activity released should be the maximum coolant activity Xe-133 values are at the maximum allowed by technical allowed by the technical specifications. Two cases of iodine spiking specifications.

should be assumed.

Since no fuel damage occurs, two cases of iodine spiking (pre-accident and accident-initiated) were modeled.

Appendix A reactor transient has occurred prior to the postulated main steam Conforms The pre-accident iodine spike was modeled with a primary coolant E 2.1 line break (MSLB) and has raised the primary coolant iodine iodine concentration of 60 Ci/gm DE I-131, consistent with the concentration to the maximum value (typically 60 Ci/gm DE I- Technical Specification limit.

131) permitted by the technical specifications (i.e., a pre-accident iodine spike case).

Appendix The primary system transient associated with the MSLB causes an Conforms The accident-initiated concurrent iodine spike was modeled with a E 2.2 iodine spike in the primary system. The increase in primary coolant spike factor of 500 on the appearance rate and spike duration of 8 iodine concentration is estimated using a spiking model that hours. The initial activity was based on 1.0 Ci/gm DE I-131, assumes that the iodine release rate from the fuel rods to the consistent with the Technical Specification limit.

primary coolant (expressed in curies per unit time) increases to a value 500 times greater than the release rate corresponding to the iodine concentration at the equilibrium value (typically 1.0 Ci/gm DE I-131) specified in technical specifications (i.e., concurrent iodine spike case). A concurrent iodine spike need not be considered if fuel damage is postulated. The assumed iodine spike duration should be 8 hours9.259259e-5 days <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.

ULNRC-06636 Enclosure 6 Page 343 of 374 Table 15B-4 Conformance with Regulatory Guide 1.183, Revision 0 Appendix E (PWR Main Steam Line Break)

RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section Appendix The activity released from the fuel should be assumed to be Conforms No fuel failures are postulated. The activity from pre-accident and E3 released instantaneously and homogeneously through the primary accident-initiated iodine spikes was modeled to be released coolant. instantaneously and homogenously throughout the primary coolant.

Appendix The chemical form of radioiodine released from the fuel should be Conforms No fuel failures are postulated. Iodine chemical fractions for steam E4 assumed to be 95% cesium iodide (CsI), 4.85% elemental iodine, generator releases to the environment (97% elemental and 3%

and 0.15% organic iodide. Iodine releases from the steam organic) were modeled in the analysis.

generators to the environment should be assumed to be 97%

elemental and 3% organic. These fractions apply to iodine released as a result of fuel damage and to iodine released during normal operations, including iodine spiking.

Transport Appendix For facilities that have not implemented alternative repair criteria Conforms The Technical Specification in-identified leakage of 1 gpm was E 5.1 (see Reference E-1, DG-1074), the primary-to-secondary leak rate assumed to be entirely to the faulted steam generator and released in the steam generators should be assumed to be the leak rate directly to the environment.

limiting condition for operation specified in the technical specifications. For facilities with traditional generator specifications (both per generator and total of all generators), the leakage should be apportioned between affected and unaffected steam generators in such a manner that the calculated dose is maximized.

Appendix The density used in converting volumetric leak rates (e.g., gpm) to Conforms A density of 1.0 gm/cc (62.4 lbm/ft3) was used.

E 5.2 mass leak rates (e.g., lbm/hr) should be consistent with the basis of the parameter being converted. The ARC leak rate correlations are generally based on the collection of cooled 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).

Appendix The primary-to-secondary leakage should be assumed to continue Conforms The primary-to-secondary leakage was terminated at 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> when E 5.3 until the primary system pressure is less than the secondary system the reactor coolant system was cooled to 212°F.

pressure, or until the temperature of the leakage is less than 100°C (212°F). The release of radioactivity from unaffected steam The release of radioactivity from the unaffected steam generators generators should be assumed to continue until shutdown cooling is was terminated at 7.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> when shutdown cooling is available in operation and releases from the steam generators have been (350°F) to remove decay heat.

terminated.

ULNRC-06636 Enclosure 6 Page 344 of 374 Table 15B-4 Conformance with Regulatory Guide 1.183, Revision 0 Appendix E (PWR Main Steam Line Break)

RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section Appendix All noble gas radionuclides released from the primary system are Conforms The noble gas radionuclides released from the primary system are E 5.4 assumed to be released to the environment without reduction or released to the environment without reduction or mitigation.

mitigation.

Appendix The transport model described in this section should be utilized for Conforms The transport model was utilized in the analysis. See items 5.5.1 E 5.5 iodine and particulate releases from the steam generators. This thru 5.5.4, below, for additional discussion.

model is shown in Figure E-1 and summarized below:

Appendix A portion of the primary-to-secondary leakage will flash to vapor, Conforms All primary-to-secondary leakage was assumed to be to the faulted E 5.5.1 based on the thermodynamic conditions in the reactor and steam generator. The faulted steam generator was assumed to secondary coolant. blowdown to dryout conditions and the primary-to secondary

• During periods of steam generator dryout, all of the leakage was modeled as a release to the environment with no primary-to-secondary leakage is assumed to flash to vapor mitigation.

and be released to the environment with no mitigation.

• With regard to the unaffected steam generators used for There is no primary-to-secondary leakage to the unaffected steam plant cooldown, the primary-to-secondary leakage can be generators.

assumed to mix with the secondary water without flashing during periods of total tube submergence.

ULNRC-06636 Enclosure 6 Page 345 of 374 Table 15B-4 Conformance with Regulatory Guide 1.183, Revision 0 Appendix E (PWR Main Steam Line Break)

RG Regulatory Guide 1.183 Position Analysis Basis of Conformance Section Appendix The leakage that immediately flashes to vapor will rise through the Not See Item 5.5.1 above. All leakage is assumed to be to the faulted E 5.5.2 bulk water of the steam generator and enter the steam space. Credit Applicable generator and released directly to the environment with no may be taken for scrubbing in the generator, using the models in mitigation.

NUREG-0409, Iodine Behavior in a PWR Cooling System Following a Postulated Steam Generator Tube Rupture Accident (Reference E-2), during periods of total submergence of the tubes.

Appendix The leakage that does not immediately flash is assumed to mix with Not See Item 5.5.1 above. All leakage is assumed to be to the faulted E 5.5.3 the bulk water. Applicable generator and released directly to the environment with no mitigation.

Appendix The radioactivity in the bulk water is assumed to become vapor at a Conforms A partition coefficient of 100 was modeled for iodine and E 5.5.4 rate that is the function of the steaming rate and the partition particulates. The moisture carryover from the steam generators for coefficient. A partition coefficient for iodine of 100 may be particulate retention is 0.1%. This is equivalent to a partition factor assumed. The retention of particulate radionuclides in the steam of 1000 and therefore, assuming a partition factor of 100 for generators is limited by the moisture carryover from the steam particulates is conservative.

generators.

Appendix Operating experience and analyses have shown that for some steam Not See Item 5.5.1 above. All primary-to-secondary leakage was E 5.6 generator designs, tube uncovery may occur for a short period Applicable assumed to be to the faulted steam generator. The faulted steam following any reactor trip (Reference E-3). The potential impact of generator was assumed to blowdown to dryout conditions and the tube uncovery on the transport model parameters (e.g., flash primary-to secondary leakage was modeled as a release to the fraction, scrubbing credit) needs to be considered. The impact of environment with no mitigation.

emergency operating procedure restoration strategies on steam generator water levels should be evaluated.

Table 15B-5 Conformance with Regulatory Guide 1.183, Revision 0 Appendix F (PWR Steam Generator Tube Rupture Accident)

RG Section Regulatory Guide 1.183 Position Analysis Basis of Conformance Appendix Assumptions for Evaluating the Radiological Consequences of a F PWR Steam Generator Tube Rupture Accident Source Term

ULNRC-06636 Enclosure 6 Page 346 of 374 Table 15B-5 Conformance with Regulatory Guide 1.183, Revision 0 Appendix F (PWR Steam Generator Tube Rupture Accident)

RG Section Regulatory Guide 1.183 Position Analysis Basis of Conformance Appendix Assumptions acceptable to the NRC staff regarding core inventory Conforms The inventory of fission products in the reactor core and available F1 and the release of radionuclides from the fuel are in Regulatory for release from the fuel into the RCS was based on the 1% fuel Position 3 of this guide. The release from the breached fuel is based defects described in the FSAR and scaled up according to DE I-131 on Regulatory Position 3.2 of this guide and the estimate of the and DE Xe-133 Technical Specifications number of fuel rods breached.

Appendix If no or minimal fuel damage is postulated for the limiting event, Conforms No fuel damage is postulated. The maximum coolant activities F2 the activity released should be the maximum coolant activity allowed by Tech Spec 3.4.16 [1 Ci/gm DE I-131 (equilibrium), 60 allowed by technical specification. Two cases of iodine spiking Ci/gm DE I-131 (transient) and 225 Ci/gm DE XE-133] are used should be assumed. in the iodine spiking cases.

Appendix A reactor transient has occurred prior to the postulated steam Conforms The pre-accident iodine spike case uses the maximum coolant F 2.1 generator tube rupture (SGTR) and has raised the primary coolant activities allowed by Tech Spec 3.4.16 (60 Ci/gm DE I-131 iodine concentration to the maximum value (typically 60 Ci/gm (transient) and 225 Ci/gm DE XE-133) as the source term released DE I-131) permitted by the technical specifications (i.e., a to the primary coolant.

preaccident iodine spike case).

Appendix The primary system transient associated with the SGTR causes an Conforms The concurrent iodine spike case uses the equilibrium coolant F 2.2 iodine spike in the primary system. The increase in primary coolant activities allowed by Tech Spec 3.4.16 (1 Ci/gm DE I-131 and iodine concentration is estimated using a spiking model that 225 Ci/gm DE XE-133) and applies the 335 spiking factor to the assumes that the iodine release rate from the fuel rods to the determined release rate for the 8-hour duration as indicated. The primary coolant (expressed in curies per unit time) increases to a effects of decay, purification from letdown, and leakage are value 335 times greater than the release rate corresponding to the accounted for over the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> duration when determining the iodine iodine concentration at the equilibrium value (typically 1.0 Ci/gm release rates.

DE I-131) specified in technical specifications (i.e., concurrent iodine spike case). A concurrent iodine spike need not be considered if fuel damage is postulated. The assumed iodine spike duration should be 8 hours9.259259e-5 days <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.

Appendix The activity released from the fuel, if any, should be assumed to be Conforms No activity released from damaged fuel. Source term is the F3 released instantaneously and homogeneously through the primary maximum coolant activity allowed by Tech Spec 3.4.16 per Item F coolant. 2 above.

Appendix Iodine releases from the steam generators to the environment Conforms The chemical form of the iodine from steam generator releases are F4 should be assumed to be 97% elemental and 3% organic. 97% elemental and 3% organic.

ULNRC-06636 Enclosure 6 Page 347 of 374 Table 15B-5 Conformance with Regulatory Guide 1.183, Revision 0 Appendix F (PWR Steam Generator Tube Rupture Accident)

RG Section Regulatory Guide 1.183 Position Analysis Basis of Conformance Appendix The primary-to-secondary leak rate in the steam generators should The primary-to-secondary leak rate is based on Callaway FSAR SP F 5.1 be assumed to be the leak rate limiting condition for operation Conforms Section 15.3.3.3.1.2, Rev. OL-18, 500 lbm/hr (equivalent to 1gpm specified in the technical specifications. The leakage should be according to LCO 3.4.13).

apportioned between affected and unaffected steam generators in The leakage is apportioned equally between the SGs since all noble such a manner that the calculated dose is maximized. gases are released without reduction or mitigation and the partition coefficient in bulk water is the same for all SGs. The flashing fraction in the ruptured SG is also conservatively applied to the intact SGs to ensure the maximum activity release is modeled.

Appendix The density used in converting volumetric leak rates (e.g., gpm) to Not used. The primary-to-secondary leak rate is based on Callaway Technical F 5.2 mass leak rates (e.g., lbm/hr) should be consistent with the basis of Specific 3.4.13 Basis Safety Analysis Limit, 1 gpm. FSAR SP surveillance tests used to show compliance with leak rate technical Section 15.3.3.3.1.2, Rev. OL-18, Item d shows a leak rate of 1 specifications. These tests are typically based on cool liquid. gpm (500 lbm/hr).which is already in mass leak rate form.

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).

Appendix The primary-to-secondary leakage should be assumed to continue Conforms The release of radioactivity is based on steam release analyses F 5.3 until the primary system pressure is less than the secondary system which have been biased to maximize the radiological consequences.

pressure, or until the temperature of the leakage is less than 100 °C The steam release analysis documents that shutdown cooling is (212 °F). The release of radioactivity from the unaffected steam established at 5.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> for the SGTR with a stuck open ADV and generators should be assumed to continue until shutdown cooling is at 6.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> with overfill.

in operation and releases from the steam generators have been terminated.

Appendix The release of fission products from the secondary system should Conforms The secondary system release is through the MSSVs, ASDs, F 5.4 be evaluated with the assumption of a coincident loss of offsite TDAFP exhaust with no credit for a partition coefficient when the power. secondary system release is through the condenser.

Appendix All noble gas radionuclides released from the primary system are Conforms The noble gas radionuclides released in the model are released F 5.5 assumed to be released to the environment without reduction or without reduction or mitigation.

mitigation.

ULNRC-06636 Enclosure 6 Page 348 of 374 Table 15B-5 Conformance with Regulatory Guide 1.183, Revision 0 Appendix F (PWR Steam Generator Tube Rupture Accident)

RG Section Regulatory Guide 1.183 Position Analysis Basis of Conformance Appendix The transport model described in Regulatory Positions 5.5 and 5.6 Conforms. The transport model described in Regulatory Positions 5.5 and 5.6 F 5.6 of Appendix E should be utilized for iodine and particulates. of Appendix E for iodine and particulates was considered as appropriate for the SGTR Accident.

Additionally, it should be noted that within NL17006A060, PG&E Letter DCL-16-124 there is a discussion on page 94 of 221 that The effect of SG tube uncovery in intact SGs (for SGTR and non-SGTR events), has been evaluated for potential impact on dose consequences as part of a Westinghouse Owners Group (WOG)

Program and demonstrated to be insignificant; For the SGTR (Overfill and stuck open ADV), this work for Callaway accommodates 0.83 hours9.606481e-4 days <br />0.0231 hours <br />1.372354e-4 weeks <br />3.15815e-5 months <br /> of SG tube uncovery. This approach is very conservative relative to NRC acceptance of previous Westinghouse Owners Group work that shows no (complete) uncovery of SG tubes. More specifically, in response to NRC Information Notice 88-31, WOG letter OG-92-25, dated 3 92 summarized work on this subject and submitted WCAP-13247 to the NRC.

The NRC response to the WOG submittal is a letter from Robert C.

Jones, dated 3-10-93, which concludes that the Westinghouse analyses demonstrate that the effects of partial steam generator tube uncovery on the iodine release for SGTR and non-SGTR events is negligible.

In the context of more recent events, the License Amendment Request to implement AST at Diablo Canyon referenced the Westinghouse Owners Group work in saying that the effect of SG tube uncovery in SGTR and non-SGTR events has been evaluated and demonstrated to be insignificant. See pages 94 and 104 of 221 of ADAMS accession number ML17006A060. NRC approval of this approach is reflected in acceptance of (only) an iodine partition coefficient of 100 in the associated Safety Evaluation released as an attachment to the NRC letter dated 4-27-17. See pages 63 and 89 of 115 of ADAMS accession number ML17012A246.

ULNRC-06636 Enclosure 6 Page 349 of 374 Table 15B-6 Conformance with Regulatory Guide 1.183, Revision 0 Appendix G (PWR Locked Rotor Accident)

RG Section Regulatory Guide 1.183 Position Analysis Basis of Conformance Appendix Assumptions for Evaluating the Radiological Consequences of a G PWR Locked Rotor Accident Source Term The inventory of fission products in the reactor core and available Appendix Assumptions acceptable to the NRC staff regarding core inventory Conforms for release from the fuel into the RCS was based on the maximum and the release of radionuclides from the fuel are in Regulatory G1 full power operation with a core thermal power of 3636 MWt Position 3 of this regulatory guide. The release from the breached (102% of 3565 MWt nominal power).

fuel is based on Regulatory Position 3.2 of this guide and the Core design parameters (enrichment, burnup, and MTU loading) estimate of the number of fuel rods breached.

are based on the cycles 19 through 22 with conservative increases in enrichment and burnup. Margin is added to the EOC core inventory, calculated with ORIGEN-S, to account for potential core design differences in future cycles. For the DBA Locked Rotor Accident, 5% of the fuel assemblies were assumed to be affected, damaged, and 35 rods per damaged assembly were assumed to be affected by Footnote 11.

Appendix If no fuel damage is postulated for the limiting event, a radiological Conforms For the DBA Locked Rotor Accident, 5% of the fuel assemblies analysis is not required as the consequences of this event are were assumed to be affected, damaged, and 35 rods per damaged G2 bounded by the consequences projected for the main steam line assembly were assumed to be affected by Footnote 11.

break outside containment.

Appendix The activity released from the fuel should be assumed to be Conforms The radionuclides from the damaged assemblies start in the RCS released instantaneously and homogeneously throughout the compartment and not the fuel compartment, therefore being G3 primary coolant. instantaneously and homogeneously distributed throughout the primary coolant.

Appendix The chemical form of radioiodine released from the fuel should be Conforms The chemical form of the iodine from steam generator releases are assumed to be 95% cesium iodide (CsI), 4.85 percent elemental 97% elemental and 3%.

G4 iodine, and 0.15 percent organic iodine. Iodine releases from the steam generators to the environment should be assumed to be 97%

elemental and 3% organic. These fractions apply to iodine released as a result of fuel damage and to iodine released during normal operations, including iodine spiking.

ULNRC-06636 Enclosure 6 Page 350 of 374 Table 15B-6 Conformance with Regulatory Guide 1.183, Revision 0 Appendix G (PWR Locked Rotor Accident)

RG Section Regulatory Guide 1.183 Position Analysis Basis of Conformance The primary-to-secondary leak rate is based on Callaway Technical Appendix The primary-to-secondary leak rate in the steam generators should Specific 3.4.13 Basis Safety Analysis Limit, 1 gpm. FSAR SP be assumed to be the leak-rate-limiting condition for operation G 5.1 Conforms Section 15.3.3.3.1.2, Rev. OL-18, Item d shows a leak rate of 1 specified in the technical specifications. The leakage should be gpm (500 lbm/hr).

apportioned between the steam generators in such a manner that the calculated dose is maximized.

Appendix The density used in converting volumetric leak rates (e.g., gpm) to Not used. The primary-to-secondary leak rate is based on Callaway G 5.2 mass leak rates (e.g., lbm/hr) should be consistent with the basis of Technaical Specific 3.4.13 Basis Safety Analysis Limit, 1 gpm.

surveillance tests used to show compliance with leak rate technical FSAR SP Section 15.3.3.3.1.2, Rev. OL-18, Item d shows a leak specifications. These tests are typically based on cool liquid. rate of 1 gpm (500 lbm/hr), which is already in mass leak rate form.

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).

Appendix The primary-to-secondary leakage should be assumed to continue Conforms The release of radioactivity is based on steam release analyses until the primary system pressure is less than the secondary system which have been biased to maximize the radiological consequences.

G 5.3 pressure, or until the temperature of the leakage is less than 100 °C The steam release analysis documents that shutdown cooling is (212 °F). The release of radioactivity should be assumed to established at 7.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> for the locked rotor accident.

continue until shutdown cooling is in operation and releases from the steam generators have been terminated.

Appendix The release of fission products from the secondary system should Conforms The secondary system release is through the MSSVs, ASDs, be evaluated with the assumption of a coincident loss of offsite TDAFP exhaust with no credit for a partition coefficient when the G 5.4 power. secondary system release is through the condenser.

Appendix All noble gas radionuclides released from the primary system are Conforms The noble gas radionuclides released in the model are released assumed to be released to the environment without reduction or without reduction or mitigation.

G 5.5 mitigation.

ULNRC-06636 Enclosure 6 Page 351 of 374 Table 15B-6 Conformance with Regulatory Guide 1.183, Revision 0 Appendix G (PWR Locked Rotor Accident)

RG Section Regulatory Guide 1.183 Position Analysis Basis of Conformance The transport model described in assumptions 5.5 and 5.6 of The transport model described in Regulatory Positions 5.5 and 5.6 Appendix Conforms.

Appendix E should be utilized for iodine and particulates. of Appendix E for iodine and particulates was considered as G 5.6 appropriate for the Locked Rotor Accident.

Additionally, it should be noted that within NL17006A060, PG&E Letter DCL-16-124 there is a discussion on page 94 of 221 that The effect of SG tube uncovery in intact SGs (for SGTR and non-SGTR events), has been evaluated for potential impact on dose consequences as part of a Westinghouse Owners Group (WOG)

Program and demonstrated to be insignificant; This work for Callaway accommodates 2.48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of SG tube uncovery. This approach is very conservative relative to NRC acceptance of previous Westinghouse Owners Group work that shows no (complete) uncovery of SG tubes. More specifically, in response to NRC Information Notice 88-31, WOG letter OG-92-25, dated 3-31-92 summarized work on this subject and submitted WCAP-13247 to the NRC.

The NRC response to the WOG submittal is a letter from Robert C.

Jones, dated 3-10-93, which concludes that the Westinghouse analyses demonstrate that the effects of partial steam generator tube uncovery on the iodine release for SGTR and non-SGTR events is negligible.

In the context of more recent events, the License Amendment Request to implement AST at Diablo Canyon referenced the Westinghouse Owners Group work in saying that the effect of SG tube uncovery in SGTR and non-SGTR events has been evaluated and demonstrated to be insignificant. See pages 94 and 104 of 221 of ADAMS accession number ML17006A060. NRC approval of this approach is reflected in acceptance of (only) an iodine partition coefficient of 100 in the associated Safety Evaluation released as an attachment to the NRC letter dated 4-27-17. See pages 63 and 89 of 115 of ADAMS accession number ML17012A246.

ULNRC-06636 Enclosure 6 Page 352 of 374 Table 15B-7 Conformance with Regulatory Guide 1.183, Revision 0 Appendix H (PWR Rod Ejection Accident)

RG 1.183 Regulatory Guide 1.183 Position Analysis Basis of Conformance Section Appendix H Assumptions for Evaluating the Radiological Consequences of a PWR Rod Ejection Accident Source Term Appendix Assumptions acceptable to the NRC Staff regarding core inventory are in Conforms The inventory of fission products in the reactor core and available H1 Regulatory Position 3 of this guide. For the rod ejection accident, the for release to the containment was based on the maximum full power release from the breached fuel is based on the estimate of the number of operation with a core thermal power of 3636 MWt (102% of 3565 fuel rods breached and the assumption that 10% of the core inventory of the MWt nominal power).

noble gases and iodines is in the fuel gap. The release attributed to fuel Core design parameters (enrichment, burnup, and MTU loading) are melting is based on the fraction of the fuel that reaches or exceeds the based on the cycles 19 through 22 with conservative increases in initiation temperature for fuel melting and the assumption that 100% of the enrichment and burnup. Margin is added to the EOC core inventory, noble gases and 25% of the iodines contained in that fraction are available calculated with ORIGEN-S, to account for potential core design for release from containment. For the secondary system release pathway, differences in future cycles. For the DBA CRE, all fuel assemblies 100% of the noble gases and 50% of the iodines in that fraction are released were assumed to be affected and a conservatively bounding core to the reactor coolant. inventory was used.

Appendix If no fuel damage is postulated for the limiting event, a radiological Conforms As fuel damage is postulated for a rod ejection event at Callaway, H2 analysis is not required as the consequences of this event are bounded by this analysis is performed to confirm the dose consequences of a rod the consequences projected for the loss-of-coolant accident (LOCA), main ejection are bounded by a LOCA.

steam line break, and steam generator tube rupture.

ULNRC-06636 Enclosure 6 Page 353 of 374 Table 15B-7 Conformance with Regulatory Guide 1.183, Revision 0 Appendix H (PWR Rod Ejection Accident)

RG 1.183 Regulatory Guide 1.183 Position Analysis Basis of Conformance Section Appendix Two release cases are to be considered. In the first, 100% of the activity Conforms RADTRAD does not allow the user to specify more than one NIF H3 released from the fuel should be assumed to be released instantaneously file for a given plant scenario file; therefore, the primary to and homogeneously through the containment atmosphere. In the second, secondary leakage case requires a third case to analyze the release of 100% of the activity released from the fuel should be assumed to be the equilibrium activity contained in the secondary coolant prior to completely dissolved in the primary coolant and available for release to the the accident. According to Appendix A Item 7.3, the noble gases secondary system. leaked form primary to secondary coolant must be released directly to environment without mitigation. RADTRAD does not have an effective mechanism for releasing the noble gases without reduction without affecting the iodine inventories; therefore, the primary to secondary leakage case is divided into two cases to analyze the full source term applicable to a primary to secondary leakage case. In total, four RADTRAD-NAI cases are analyzed to produce the dose consequences of the two release cases defined by RG 1.183 Appendix A Item 3.

• Containment Leakage

• Primary to Secondary Noble Gas Leakage

• Primary to Secondary Non-Noble Gas Leakage (Alkalis and Halogens)

• Secondary Equilibrium Initial Activity Leakage Appendix The chemical form of radioiodine released to the containment atmosphere Conforms The iodine chemical fractions for release to containment are 95%

H4 should be assumed to be 95% cesium iodide (CsI), 4.85% elemental iodine, CsI, 4.85% elemental, and 0.15% organic. All fission products, with and 0.15% organic iodide. If containment sprays do not actuate or are the exception of elemental and organic iodine and noble gases, are terminated prior to accumulating sump water, or if the containment sump assumed to be in particulate form.

pH is not controlled at values of 7 or greater, the iodine species should be Leakage and decay are the only removal processes modeled in evaluated on an individual case basis. Evaluations of pH should consider containment. With no iodine being held in the sump water, sump pH the effect of acids created during the rod ejection accident event, e.g., has no impact.

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.

Appendix Iodine releases from the steam generators to the environment should be Conforms The primary to secondary leakage cases and secondary initial H5 assumed to be 97% elemental and 3% organic. activity case implement the composition: 97% elemental iodine, and 3% organic iodide.

Appendix Transport from Containment H 6.0

ULNRC-06636 Enclosure 6 Page 354 of 374 Table 15B-7 Conformance with Regulatory Guide 1.183, Revision 0 Appendix H (PWR Rod Ejection Accident)

RG 1.183 Regulatory Guide 1.183 Position Analysis Basis of Conformance Section Appendix A reduction in the amount of radioactive material available for leakage Conforms Containment spray and natural deposition are available for credit at H 6.1 from the containment that is due to natural deposition, containment sprays, Callaway but are conservatively neglected in this analysis.

recirculating filter systems, dual containments, or other engineered safety Neglecting these engineered safety features provides margin for the features may be taken into account. Refer to Appendix A to this guide for potential that containment spray does not actuate during a rod guidance on acceptable methods and assumptions for evaluating these ejection accident.

mechanisms.

Appendix The containment should be assumed to leak at the leak rate incorporated in Conforms From Technical Specification 5.5.16.c., the maximum allowable H 6.2 the technical specifications at peak 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 />, containment leakage rate, La, at Pa, shall be 0.20% of the and at 50% of this leak rate for the remaining duration of the accident. Peak containment air weight per day.

accident pressure is the maximum pressure defined in the technical specifications for containment leak testing. Leakage from subatmospheric After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, this is reduced to 0.10% per day.

containments is assumed to be terminated when the containment is brought to a subatmospheric condition as defined in technical specifications.

Appendix H Transport from Secondary System 7.0 Appendix A leak rate equivalent to the primary-to-secondary leak rate limiting Conforms The primary-to-secondary leak rate is based on Callaway FSAR SP H 7.1 condition for operation specified in the technical specifications should be Section 15.4.8.3.1.2 and LCO 3.4.13, 1 gpm.

assumed to exist until shutdown cooling is in operation and releases from the steam generators have been terminated. The release of radioactivity is based on steam release analyses which have been biased to maximize the radiological consequences. The steam release analysis documents that shutdown cooling is established at 7.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> for the Rod ejection accident.

Appendix The density used in converting volumetric leak rates (e.g., gpm) to mass Conforms The primary-to-secondary leak rate based on Callaway FSAR SP H 7.2 leak rates (e.g., lbm/hr) should be consistent with the basis of surveillance Section 15.4.8.3.1.2, 1 gpm, is converted to 8.34 lbm/min as shown tests used to show compliance with leak rate technical specifications. These below.

tests typically are based on cooled liquid. The facilitys instrumentation used to determine leakage typically is located on lines containing cool 0.13373 62.4 1 ( )( ) = 8.34 /

liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 3 lbm/ft3).

Appendix All noble gas radionuclides released to the secondary system are assumed Conforms The noble gas radionuclides released in the models are released H 7.3 to be released to the environment without reduction or mitigation. without reduction or mitigation.

Appendix The transport model described in assumptions 5.5 and 5.6 of Appendix E Conforms Use of Assumptions 5.5 and 5.6 of Appendix E is documented H 7.4 should be utilized for iodine and particulates.

ULNRC-06636 Page 355 of 374 Chapter 15 Replacement Figures Figure 15.6-3a: Replaced with Tab A in spreadsheet SGTR_FSAR_Plots.xlsx Figure 15.6-3b: Replaced with Tab B in spreadsheet SGTR_FSAR_Plots.xlsx Figure 15.6-3c: Replaced with Tab C in spreadsheet SGTR_FSAR_Plots.xlsx Figure 15.6-3d: Replaced with Tab D in spreadsheet SGTR_FSAR_Plots.xlsx Figure 15.6-3e: Replaced with Tab E in spreadsheet SGTR_FSAR_Plots.xlsx Figure 15.6-3f: Replaced with Tab F in spreadsheet SGTR_FSAR_Plots.xlsx Figure 15.6-3g: Replaced with Tab G in spreadsheet SGTR_FSAR_Plots.xlsx Figure 15.6-3h: Replaced with Tab H in spreadsheet SGTR_FSAR_Plots.xlsx Figure 15.6-3i: Replaced with Tab I in spreadsheet SGTR_FSAR_Plots.xlsx Figure 15.6-3j: Replaced with Tab J in spreadsheet SGTR_FSAR_Plots.xlsx Figure 15.6-3k: Replaced with Tab K in spreadsheet SGTR_FSAR_Plots.xlsx Figure 15.6-3l: Replaced with Tab L in spreadsheet SGTR_FSAR_Plots.xlsx Figure 15.6-3m: Replaced with Tab M in spreadsheet SGTR_FSAR_Plots.xlsx Figure 15.6-3n: Replaced with Tab N in spreadsheet SGTR_FSAR_Plots.xlsx Figure 15.6-3o: Replaced with Tab O in spreadsheet SGTR_FSAR_Plots.xlsx Figure 15.6-3p: Replaced with Tab P in spreadsheet SGTR_FSAR_Plots.xlsx Figure 15.6-33a: Replaced with Tab A in spreadsheet Overfill_FSAR_Plots.xlsx Figure 15.6-33b: Replaced with Tab B in spreadsheet Overfill_FSAR_Plots.xlsx Figure 15.6-33c: Replaced with Tab C in spreadsheet Overfill_FSAR_Plots.xlsx Figure 15.6-33d: Replaced with Tab D in spreadsheet Overfill_FSAR_Plots.xlsx Figure 15.6-33e: Replaced with Tab E in spreadsheet Overfill_FSAR_Plots.xlsx Figure 15.6-33f: Replaced with Tab F in spreadsheet Overfill_FSAR_Plots.xlsx Figure 15.6-33g: Replaced with Tab G in spreadsheet Overfill_FSAR_Plots.xlsx Figure 15.6-33h: Replaced with Tab H in spreadsheet Overfill_FSAR_Plots.xlsx Figure 15.6-33i: Replaced with Tab I in spreadsheet Overfill_FSAR_Plots.xlsx Figure 15.6-33j: Replaced with Tab J in spreadsheet Overfill_FSAR_Plots.xlsx Figure 15.6-33k: Replaced with Tab K in spreadsheet Overfill_FSAR_Plots.xlsx Figure 15.6-33l: Replaced with Tab L in spreadsheet Overfill_FSAR_Plots.xlsx Figure 15.6-33m: Replaced with Tab M in spreadsheet Overfill_FSAR_Plots.xlsx Figure 15.6-33n: Replaced with Tab N in spreadsheet Overfill_FSAR_Plots.xlsx Figure 15A-1 Figure 15A-2 Figure 15A-3

ULNRC-06636 Enclosure 6 Page 356 of 374 HOLD UP IN LOCA holdup, FHA (2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> CONTAINMENT release) and RE (holdup)

A No holdup: Radwaste CONTROL ROOM STEAM VENT & F SAFETY VALVES OR RADWASTE BLDG SGTR, MSLB (partial holdup),

A LR, LOOP and RE ATMOSPHERE E

FUEL TECH. SUPPORT BUILDING (NO FHA (2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> release) CENTER HOLDUP) B G LLB AUX.

BUILDING (NO HOLDUP) C Filter LOCA ECCS recirc EMER. OP. FACILITY D H RWST Post LOCA backleakage (HOLDUP) pathway R

REV. OL-10 6/20 CALLAWAY PLANT FIGURE 15A-1 RELEASE PATHWAYS

ULNRC-06636 Page 357 of 374 Allowable Inleakage Values 6000 5000 Unacceptable Control Building Unfiltered Inleakage (cfm) 4000 3000 Acceptable 2000 1000 0

0 10 20 30 40 50 60 Control Room Unfiltered Inleakage (cfm)

REV. 17 6/20 CALLAWAY PLANT FIGURE 15A-2 ALLOWABLE INLEAKAGE VALUES

ULNRC-06636 Enclosure 6 Page 358 of 374 Figure 15A-3 Control Room Model F 24 Control Room Unfiltered Inleakage F 37 AND F 27 CR Exhaust to Environment Function of Control Building Half-Train Failure CB leakage to CR F 19 Dotted Line - RM 1512 / 1501 Enclosure F 16 Cntrl Bldg Filtered Intake CRFU Balance <-- CR Return to Filter Unit F 26 Presurization Filter Pathway

<-- CR Return to AC Plenum F 31 900cfm for 30 min, then 450cfm CR AC Balance F 21 F 18 CRFU CRFU Filter 18,000 cfm +/- 10% F 23 Control Room CR AC Plenum 400 +/- 40cfm cfm Intake Pathway minus 350+/11-%

2000cfm +/-200cfm Environment Cntrl Bldg Unfiltered Inleakage F 17 F 22 F 25 350cfm +/- 35cfm Function of Equipment Room F 28 ER to CRFU Intake: 300cfm +/- 30cfm F 20 F 30 Balance ER Balance Normal flow to ER F 41 F 29 Normal flow to CR F 39 F 40 Normal flow to CB F 36 Normal CB/ER/CR Supply flow from Env Normal CB/CR/ER HVAC Supply Plenum F 38 Normal CB/ER/CR Exhaust/Outleakage to Env

ULNRC-06636 Page 359 of 374 Insert AddendumChapter2LOT 2.3-87 Joint Frequency Distribution (in percent of total hours) for Stability Class A 2.3-88 Joint Frequency Distribution (in percent of total hours) for Stability Class B 2.3-89 Joint Frequency Distribution (in percent of total hours) for Stability Class C 2.3-90 Joint Frequency Distribution (in percent of total hours) for Stability Class D 2.3-91 Joint Frequency Distribution (in percent of total hours) for Stability Class E 2.3-92 Joint Frequency Distribution (in percent of total hours) for Stability Class F 2.3-93 Joint Frequency Distribution (in percent of total hours) for Stability Class G 2.3-94 Wind Direction Occurrence Frequency 2.3-95 Wind Speed Occurrence Frequency 2.3-96 Exclusion Area Boundary and Low Population Zone /QS 2.3-97 Release/receptor pairs and inputs for /Q calculation 2.3-98 Calculated /Q values and associated events Insert Addendum2.3.4.4 2.3.4.4 Alternative Source Term Short-Term Diffusion Estimates The alternative source term (AST) methodology implements the guidance of NRC Regulatory Guide (RG) 1.183 (2000).

2.3.4.4.1 Short-term (Accident) Atmospheric Dispersion Factors for the Exclusion Area Boundary and the Low Population Zone For AST Conservative values of atmospheric dispersion factors at the exclusion area boundary (EAB) and the low population zone (LPZ) were calculated for appropriate time periods using meteorological data collected onsite. Four consecutive years of hourly measured site-specific meteorological data from January 1, 2013 to December 31st, 2016 were used in the evaluations. Meteorological data used are described in the joint frequency distributions of wind speed, wind direction, and of the atmospheric stability class presented Table 2.3-87 through 2.3-95.

2.3.4.4.1.1 Methodology The methodology used for this calculation is consistent with Regulatory Guide (RG) 1.145 as implemented by the PAVAN computer code (Bander, 1982). Using joint frequency distributions of wind direction and wind speed by atmospheric stability, the PAVAN computer code provides relative air concentration (/Q) values as functions of direction for various time periods at the site boundary and LPZ. Three procedures for calculation of /Qs are utilized for the EAB and LPZ; a direction-dependent approach, a direction-independent approach, and an overall site /Q approach. The /Q calculations are based on the theory that material released to the atmosphere will be normally distributed (Gaussian) about the plume centerline. A straight-line trajectory is assumed between the point of release and all distances for which /Q values are calculated.

ULNRC-06636 Page 360 of 374 The theory and implementing equations employed by the PAVAN computer code are documented in Bander (1982).

2.3.4.4.1.2 Calculations/PAVAN Computer Code Input Data The minimum EAB distance assumed for all directions is 1200 meters from the midpoint of the Unit 1 reactor building and the canceled Unit 2 reactor building. The LPZ distance is taken as 2.5 miles from the midpoint of the Unit 1 reactor building and the canceled Unit 2 reactor building in all directions.

Two PAVAN cases were executed for the offsite short term /Q determination. The first case (RB) simulated a release from the midpoint between the operating Unit 1 containment/reactor building and disabled Unit 2 containment/RB. The second case (RWST) simulated a release from the RWST. The cases differ based on the building wake effects (i.e. building area) and the release heights.

All of the releases were considered ground level releases because the highest possible release elevation is from the plant stack at 217.4 ft. From Section 1.3.2 of RG 1.145, a release is only considered a stack release if the release point is at a level higher than two and one-half times the height of adjacent solid structures. For the Callaway plant, the elevation of the top of the Unit 1 containment is 140.5 ft.

Therefore, the highest possible release point is not 2.5 times higher than the adjacent containment buildings, and thus all releases were considered ground level releases. As such, the release height was set equal to 10.0 meters as required by Table 3.1 of Bander (1982). The building cross-sectional areas used for the building wake term were 1,526 m2 for the RB release and 171 m2 for the RWST release. The area of the containment was calculated to be conservatively small in that the height used in the area calculation was from the highest roof elevation of a nearby building to the elevation of the bottom of the containment dome. The area for the RWST release conservatively used the smaller RWST profile, since it is significantly distant from the containment structure and therefore the dominant wakes are those of the RWST.

The tower heights at which the wind speeds were measured are 10 m and 60 m above plant grade. The wind speed units are given in meters per second; therefore, the PAVAN variable UCOR was set equal to -

1 to keep the wind speeds in meters per second. The maximum wind speed in each wind speed category was chosen to match the raw joint frequency distribution data, which conforms to the wind speed bins in Table 1 of RG 1.23. The maximum wind speed values are 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 mps. The maximum windspeed in each windspeed category was chosen to match the recommendation of RIS-2006-4 (USNRC, 2006).

2.3.4.4.1.3 Results PAVAN computer runs for the EAB and LPZ boundary distances were performed using the data discussed previously. Per Section 4 of RG 1.145, the maximum /Q for each distance was determined and compared to the 5% overall site value for the boundary under consideration. The maximum EAB and LPZ

/Qs that resulted from this comparison are provided in Table 2.3-96.

ULNRC-06636 Page 361 of 374 2.3.4.4.2 Short-term (Accident) Atmospheric Dispersion Factors For Onsite Receptors For AST Conservative values of atmospheric dispersion factors to the emergency control room intake and the normal operation control room intake were calculated for appropriate time periods using meteorological data collected onsite. Four consecutive years of hourly measured site-specific meteorological data from January 1, 2013 to December 31st, 2016 were used in the evaluations.

Meteorological data used are described in the joint frequency distributions of wind speed, wind direction, and of the atmospheric stability class presented in Table 2.3-87 through Table 2.3-95.

2.3.4.4.2.1 Methodology The ARCON96 computer code is used by the USNRC staff to review licensee submittals relating to control room habitability (Ramsdell, 1995). Therefore, the ARCON96 computer code was used to determine the relative concentrations (/Qs) for the control room air intakes and inleakage locations.

The ARCON96 computer code uses hourly meteorological data for estimating dispersion in the vicinity of buildings to calculate relative concentrations at control room air intakes that would be exceeded no more than five percent of the time. These concentrations are calculated for averaging periods ranging from two hour to 30 days in duration.

The theory and implementing equations employed by the ACRCON96 computer code are documented in Ramsdell (1995).

2.3.4.4.2.2 Calculations/ARCON Computer Code Input Data Four years of meteorological data (2013-2016) were used for the ARCON96 computer code runs.

A number of various release-receptor combinations were considered for the control room /Qs. These different cases were considered to determine the limiting release-receptor combinations for the various events. The case matrix for these combinations is provided in Table 2.3-98.

The distance and direction inputs for the ARCON96 runs may be found in Table 2.3-97. The distances were converted from feet to meters with a factor of 0.3048 m/ft. The distances in meters were then rounded down to the nearest tenth for conservatism. The elevation difference term was set to zero for each case since all elevation points are taken with respect to the same datum.

The lower and upper measurement heights for the meteorological data were entered as 10 m and 60.0 m, respectively, for each case. The mph option was selected for the wind speed units.

A ground level release was chosen for each scenario since none of the release points are 2.5 times taller than the closet solid structure as called out in Section 3.2.2 of RG 1.194 for stack releases. The top of the containment structures is at an elevation of 140.5 ft above grade. The highest release point is from the top of the plant stack at an elevation of 217.44 ft., which is not 2.5 times higher than the nearby containment structure. The vertical velocity, stack flow, and stack radius terms were all set equal to zero since each case is a ground level release. The vent release option was not selected for any of the scenarios.

The release heights and intake heights were determined as their respective elevations less the plant grade elevation of 1999.5 ft. No credit was taken for effective release height due to plume rise; therefore, for releases from the stacks, the release elevations were set equal to the stack top elevation.

ULNRC-06636 Page 362 of 374 For those cases that included a diffuse area source (i.e. RB wall), the release height was set at the vertical center of the projected plane (see RG 1.194, Section 3.2.4.5). The horizontal distance and direction between the release sources and receptor intakes were entered.

The only cases in this analysis that take credit for the building wake effect are the scenarios where the release is from the containment building. Some of the other scenarios have buildings between the release and receptor points, but for these cases the building wake was not credited for the sake of conservatism. Not crediting wakes was accomplished by setting the building area term equal to 0.01 m2 as stated in Table A-2 of RG 1.194. The building area used is a conservatively determined containment cross sectional area. The width used is equal to the inside diameter of the containment building plus the thickness of the wall, while the height is taken as the distance between the top of the cylinder potion of the containment structure and the highest auxiliary building roof elevation. This building cross-sectional area is equal to 1,526 m2.

All of the default values in the ARCON96 code were unchanged from the code default values with the following exceptions as recommended in Table A-2 of RG 1.194:

• A value of 0.2 is used for the surface roughness length, m, in lieu of the default value of 0.1, and

• A value of 4.3 is used for the averaging sector width constant, in lieu of the default value of 4.0.

The minimum wind speed was left at 0.5 m/s per the guidance instruction in Table A-2 of RG 1.194.

2.3.4.4.2.3 Results ARCON96 computer runs for the various release points and control room intake locations were performed using the data discussed previously. Per RG 1.194, the 95th percentile /Q values were determined. The resulting /Qs are listed in Table 2.3-98.

For plants with dual CR air emergency intakes, RG 1.194, Section 3.3.2.3 states the /Q values may be reduced (by a factor of 2 or 4) to credit the dilution by the flow of dual intakes or by operator action to make the proper intake selection (i.e. air intake not in the direction of the wind). RG 1.194 goes onto say that this protocol should be used only if the dual intakes are in different wind direction windows.

For Callaway, the two CR emergency air intakes (A and B) are two feet apart. Consequently, both emergency air intakes are within the same wind direction window. Therefore, no credit was taken for dual intake dilution.

ULNRC-06636 Page 363 of 374 Insert AddendumChapter2Tables Table 2.3-87 Joint Frequency Distribution (in percent of total hours) for Stability Class A Atmospheric Stability: Class A Period of Record: January 1, 2013 to December 31, 2016 (based on lower wind speed instrument)

Maximum Wind Direction Wind Speed (m/s) N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW Total 0.22 0.00 0.50 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.75 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 1.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 1.25 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 1.50 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 2.00 0.00 0.00 0.00 0.01 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.01 3.00 0.00 0.01 0.01 0.01 0.00 0.00 0.04 0.07 0.05 0.04 0.02 0.03 0.02 0.03 0.01 0.00 0.34 4.00 0.01 0.01 0.02 0.01 0.02 0.01 0.13 0.08 0.03 0.06 0.11 0.07 0.08 0.08 0.09 0.03 0.84 5.00 0.01 0.03 0.01 0.01 0.01 0.01 0.07 0.06 0.06 0.06 0.10 0.04 0.08 0.13 0.11 0.05 0.85 6.00 0.02 0.01 0.01 0.00 0.00 0.02 0.03 0.05 0.03 0.08 0.07 0.04 0.07 0.08 0.07 0.02 0.61 8.00 0.00 0.01 0.00 0.00 0.00 0.00 0.00 0.01 0.04 0.03 0.01 0.01 0.03 0.01 0.03 0.03 0.22 10.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.01 26.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Total 0.04 0.07 0.05 0.03 0.03 0.04 0.28 0.27 0.21 0.27 0.32 0.19 0.28 0.33 0.31 0.13 2.88

ULNRC-06636 Page 364 of 374 Table 2.3-88 Joint Frequency Distribution (in percent of total hours) for Stability Class B Atmospheric Stability: Class B Period of Record: January 1, 2013 to December 31, 2016 (based on lower wind speed instrument)

Maximum Wind Direction Wind Speed (m/s) N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW Total 0.22 0.00 0.50 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.75 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 1.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 1.25 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 1.50 0.01 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.02 2.00 0.01 0.00 0.01 0.01 0.00 0.00 0.01 0.01 0.01 0.01 0.00 0.00 0.00 0.00 0.00 0.00 0.07 3.00 0.03 0.04 0.05 0.03 0.01 0.04 0.10 0.14 0.13 0.07 0.07 0.05 0.05 0.05 0.05 0.03 0.94 4.00 0.06 0.04 0.06 0.05 0.03 0.03 0.11 0.15 0.09 0.15 0.13 0.05 0.07 0.11 0.12 0.13 1.37 5.00 0.05 0.02 0.01 0.03 0.02 0.02 0.04 0.07 0.08 0.13 0.07 0.05 0.07 0.06 0.09 0.11 0.93 6.00 0.04 0.02 0.01 0.01 0.00 0.01 0.01 0.03 0.11 0.06 0.05 0.01 0.03 0.04 0.03 0.03 0.47 8.00 0.00 0.01 0.00 0.00 0.00 0.00 0.00 0.02 0.04 0.03 0.03 0.02 0.01 0.01 0.01 0.04 0.24 10.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.01 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.02 26.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Total 0.19 0.14 0.13 0.12 0.07 0.10 0.27 0.43 0.47 0.45 0.36 0.18 0.24 0.27 0.30 0.34 4.06

ULNRC-06636 Page 365 of 374 Table 2.3-89 Joint Frequency Distribution (in percent of total hours) for Stability Class C Atmospheric Stability: Class C Period of Record: January 1, 2013 to December 31, 2016 (based on lower wind speed instrument)

Maximum Wind Direction Wind Speed (m/s) N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW Total 0.22 0.00 0.50 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.75 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 1.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.01 1.25 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.01 0.00 0.00 0.00 0.00 0.00 0.01 1.50 0.00 0.01 0.00 0.00 0.00 0.00 0.01 0.01 0.02 0.00 0.01 0.00 0.01 0.00 0.01 0.00 0.08 2.00 0.02 0.01 0.03 0.02 0.02 0.01 0.02 0.07 0.03 0.05 0.03 0.03 0.04 0.05 0.03 0.01 0.47 3.00 0.09 0.09 0.10 0.08 0.07 0.07 0.28 0.27 0.22 0.22 0.12 0.08 0.11 0.13 0.10 0.10 2.14 4.00 0.07 0.11 0.05 0.07 0.05 0.06 0.19 0.21 0.20 0.16 0.13 0.07 0.10 0.12 0.13 0.12 1.83 5.00 0.10 0.05 0.02 0.03 0.02 0.05 0.05 0.08 0.10 0.15 0.09 0.04 0.05 0.06 0.05 0.11 1.05 6.00 0.05 0.03 0.01 0.01 0.01 0.02 0.01 0.03 0.09 0.07 0.04 0.03 0.03 0.05 0.05 0.07 0.60 8.00 0.01 0.00 0.00 0.00 0.00 0.00 0.00 0.03 0.08 0.03 0.01 0.01 0.01 0.03 0.03 0.02 0.28 10.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.01 0.01 0.00 0.00 0.00 0.00 0.00 0.00 0.02 26.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Total 0.34 0.29 0.21 0.21 0.17 0.22 0.57 0.72 0.75 0.70 0.43 0.27 0.36 0.44 0.40 0.43 6.51

ULNRC-06636 Page 366 of 374 Table 2.3-90 Joint Frequency Distribution (in percent of total hours) for Stability Class D Atmospheric Stability: Class D Period of Record: January 1, 2013 to December 31, 2016 (based on lower wind speed instrument)

Maximum Wind Direction Wind Speed (m/s) N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW Total 0.22 0.01 0.50 0.00 0.00 0.00 0.00 0.00 0.01 0.00 0.00 0.01 0.00 0.00 0.00 0.00 0.01 0.00 0.00 0.03 0.75 0.01 0.01 0.01 0.01 0.01 0.01 0.00 0.00 0.01 0.01 0.01 0.01 0.01 0.02 0.01 0.01 0.14 1.00 0.05 0.02 0.02 0.04 0.02 0.02 0.06 0.03 0.04 0.03 0.03 0.04 0.03 0.02 0.04 0.01 0.53 1.25 0.03 0.03 0.05 0.02 0.02 0.03 0.06 0.04 0.02 0.02 0.03 0.02 0.02 0.03 0.04 0.02 0.49 1.50 0.07 0.09 0.10 0.12 0.12 0.14 0.19 0.11 0.12 0.07 0.09 0.06 0.10 0.11 0.11 0.08 1.71 2.00 0.21 0.25 0.17 0.23 0.20 0.30 0.49 0.42 0.25 0.17 0.21 0.14 0.30 0.31 0.26 0.23 4.13 3.00 0.58 0.61 0.59 0.61 0.60 0.81 1.35 0.98 0.63 0.47 0.49 0.34 0.49 0.73 0.81 0.79 10.87 4.00 0.83 0.57 0.44 0.39 0.40 0.50 0.73 0.71 0.63 0.56 0.45 0.25 0.37 0.67 0.67 0.97 9.14 5.00 0.81 0.39 0.21 0.14 0.17 0.27 0.21 0.40 0.53 0.39 0.30 0.18 0.35 0.50 0.56 0.58 6.00 6.00 0.45 0.16 0.07 0.05 0.05 0.05 0.03 0.18 0.38 0.21 0.17 0.09 0.25 0.30 0.33 0.36 3.13 8.00 0.19 0.06 0.06 0.03 0.01 0.01 0.00 0.09 0.52 0.14 0.10 0.08 0.20 0.17 0.14 0.23 2.04 10.00 0.01 0.00 0.02 0.00 0.00 0.00 0.00 0.00 0.05 0.01 0.01 0.01 0.02 0.00 0.00 0.03 0.16 26.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.01 0.00 0.00 0.00 0.00 0.01 Total 3.25 2.19 1.76 1.63 1.61 2.15 3.12 2.96 3.18 2.07 1.90 1.23 2.14 2.88 2.99 3.32 38.37

ULNRC-06636 Page 367 of 374 Table 2.3-91 Joint Frequency Distribution (in percent of total hours) for Stability Class E Atmospheric Stability: Class E Period of Record: January 1, 2013 to December 31, 2016 (based on lower wind speed instrument)

Maximum Wind Direction Wind Speed (m/s) N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW Total 0.22 0.07 0.50 0.01 0.01 0.01 0.01 0.03 0.01 0.03 0.01 0.01 0.01 0.00 0.01 0.01 0.01 0.01 0.01 0.17 0.75 0.02 0.01 0.02 0.02 0.03 0.02 0.03 0.02 0.02 0.03 0.02 0.02 0.02 0.03 0.03 0.02 0.35 1.00 0.07 0.06 0.05 0.06 0.07 0.06 0.11 0.09 0.06 0.05 0.06 0.05 0.07 0.10 0.09 0.04 1.10 1.25 0.03 0.04 0.05 0.04 0.03 0.05 0.12 0.05 0.05 0.03 0.05 0.07 0.05 0.11 0.09 0.06 0.94 1.50 0.08 0.13 0.11 0.12 0.11 0.17 0.43 0.15 0.09 0.06 0.10 0.12 0.12 0.17 0.14 0.12 2.22 2.00 0.26 0.19 0.23 0.23 0.27 0.36 0.93 0.51 0.22 0.16 0.20 0.23 0.17 0.27 0.36 0.25 4.86 3.00 0.45 0.36 0.34 0.27 0.38 0.48 1.33 1.81 0.88 0.61 0.53 0.38 0.52 0.48 0.52 0.59 9.92 4.00 0.21 0.10 0.07 0.08 0.13 0.19 0.39 0.94 1.17 0.65 0.43 0.24 0.43 0.20 0.19 0.27 5.70 5.00 0.03 0.02 0.02 0.01 0.02 0.04 0.07 0.38 0.71 0.29 0.13 0.09 0.15 0.09 0.05 0.06 2.18 6.00 0.00 0.00 0.01 0.00 0.00 0.02 0.01 0.15 0.41 0.13 0.03 0.01 0.05 0.03 0.03 0.02 0.91 8.00 0.00 0.00 0.01 0.00 0.00 0.00 0.01 0.07 0.19 0.05 0.01 0.01 0.01 0.00 0.00 0.00 0.35 10.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.01 0.00 0.00 0.00 0.01 0.00 0.00 0.00 0.00 0.02 26.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Total 1.16 0.92 0.91 0.82 1.07 1.40 3.46 4.19 3.82 2.09 1.57 1.24 1.62 1.50 1.52 1.45 28.82

ULNRC-06636 Page 368 of 374 Table 2.3-92 Joint Frequency Distribution (in percent of total hours) for Stability Class F Atmospheric Stability: Class F Period of Record: January 1, 2013 to December 31, 2016 (based on lower wind speed instrument)

Maximum Wind Direction Wind Speed (m/s) N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW Total 0.22 0.11 0.50 0.03 0.01 0.01 0.02 0.03 0.03 0.03 0.02 0.01 0.01 0.01 0.01 0.02 0.01 0.02 0.01 0.27 0.75 0.02 0.02 0.03 0.04 0.03 0.05 0.07 0.05 0.04 0.02 0.02 0.01 0.04 0.02 0.02 0.03 0.48 1.00 0.06 0.08 0.09 0.07 0.08 0.11 0.19 0.11 0.06 0.03 0.05 0.06 0.07 0.11 0.07 0.08 1.33 1.25 0.07 0.05 0.06 0.06 0.07 0.07 0.21 0.10 0.04 0.04 0.05 0.05 0.04 0.07 0.06 0.01 1.04 1.50 0.07 0.09 0.14 0.14 0.11 0.15 0.40 0.25 0.10 0.11 0.08 0.08 0.05 0.17 0.19 0.06 2.20 2.00 0.17 0.11 0.11 0.11 0.08 0.13 0.42 0.65 0.27 0.19 0.21 0.09 0.10 0.19 0.25 0.14 3.20 3.00 0.09 0.07 0.05 0.05 0.03 0.04 0.27 1.57 0.67 0.34 0.35 0.11 0.10 0.13 0.11 0.23 4.22 4.00 0.01 0.00 0.00 0.00 0.00 0.00 0.00 0.27 0.35 0.18 0.08 0.02 0.02 0.00 0.00 0.01 0.95 5.00 0.00 0.00 0.00 0.00 0.00 0.00 0.01 0.01 0.04 0.02 0.00 0.01 0.00 0.00 0.00 0.00 0.08 6.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 8.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 10.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 26.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Total 0.51 0.43 0.48 0.49 0.43 0.58 1.60 3.04 1.60 0.93 0.84 0.43 0.44 0.69 0.71 0.58 13.90

ULNRC-06636 Page 369 of 374 Table 2.3--93 Joint Frequency Distribution (in percent of total hours) for Stability Class G Atmospheric Stability: Class G Period of Record: January 1, 2013 to December 31, 2016 (based on lower wind speed instrument)

Maximum Wind Direction Wind Speed (m/s) N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW Total 0.22 0.19 0.50 0.02 0.05 0.03 0.02 0.02 0.01 0.04 0.04 0.04 0.01 0.01 0.01 0.02 0.02 0.01 0.02 0.38 0.75 0.03 0.04 0.03 0.03 0.03 0.04 0.06 0.05 0.05 0.03 0.02 0.03 0.02 0.01 0.01 0.03 0.51 1.00 0.10 0.09 0.09 0.05 0.05 0.04 0.12 0.17 0.08 0.03 0.05 0.03 0.03 0.07 0.08 0.04 1.13 1.25 0.04 0.05 0.04 0.03 0.01 0.02 0.03 0.11 0.04 0.02 0.02 0.02 0.01 0.05 0.06 0.03 0.60 1.50 0.07 0.07 0.10 0.03 0.02 0.02 0.08 0.17 0.07 0.06 0.03 0.03 0.02 0.07 0.08 0.09 1.00 2.00 0.07 0.07 0.04 0.02 0.01 0.01 0.05 0.23 0.11 0.05 0.04 0.01 0.02 0.04 0.08 0.08 0.91 3.00 0.01 0.01 0.00 0.00 0.00 0.00 0.02 0.33 0.09 0.04 0.02 0.00 0.01 0.01 0.05 0.03 0.62 4.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.08 0.03 0.01 0.01 0.00 0.00 0.00 0.00 0.00 0.12 5.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 6.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 8.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 10.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 26.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Total 0.35 0.38 0.33 0.17 0.14 0.14 0.40 1.19 0.52 0.24 0.19 0.12 0.13 0.27 0.38 0.32 5.47

ULNRC-06636 Page 370 of 374 Table 2.3-94 Wind Direction Occurrence Frequency Period of Record: January 1, 2013 to December 31, 2016 (based on lower wind speed instrument)

Wind Direction N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW Frequency 5.85 4.42 3.88 3.47 3.52 4.63 9.69 12.78 10.55 6.75 5.61 3.66 5.21 6.39 6.62 6.57 Table 2.3-95 Wind Speed Occurrence Frequency Period of Record: January 1, 2013 to December 31, 2016 (based on lower wind speed instrument)

Maximum Wind Speed (m/s) 0.22 0.50 0.75 1.00 1.25 1.50 2.00 3.00 4.00 5.00 6.00 8.00 10.00 26.00 Frequency 0.39 0.85 1.48 4.1 3.09 7.23 13.66 29.06 19.95 11.09 5.72 3.13 0.24 0.02

ULNRC-06636 Page 371 of 374 Table 2.3-96 Exclusion Area Boundary and Low Population Zone /Qs Exclusion Area Boundary RWST Vent 0 - 720 2.05E-04 Reactor Building/Other Onsite Release Locations 0 - 720 2.00E-04 Low Population Zone RWST Vent 0-2 6.87E-05 2-8 3.57E-05 8 - 24 2.57E-05 24 - 96 1.26E-05 96 - 720 4.54E-06 Reactor Building/Other Onsite Release Locations 0-2 6.87E-05 2-8 3.42E-05 8 - 24 2.42E-05 24 - 96 1.13E-05 96 - 720 3.83E-06

ULNRC-06636 Page 372 of 374 Table 2.3-97 Release/receptor pairs and inputs for /Q calculation Horizontal Release Height Above Plant Intake Height Above Plant Direction Looking at Source From Release Point Receptor Point Distance (m) Grade (m) Grade (m) Receptor (° from True North)

Stack/Plant Vent 'B' CB intake (Emergency) 70.9 66.3 5.5 338 Stack/Plant Vent CB intake (Normal) 31.9 66.3 22.5 8 RWST 'B' CB intake (Emergency) 93.8 16.5 5.5 19 RWST CB intake (Normal) 82.7 16.5 22.5 49 FHB Closest Point 'B' CB intake (Emergency) 73.6 5.5 5.5 2 FHB Closest Point CB intake (Normal) 52.2 22.5 22.5 41 Closest ASD Midpoint between Intakes 60.7 35.5 5.5 312 Closest ASD CB intake (Normal) 15.9 35.5 22.5 292 Closest MSSV CB intake (Normal) 14.5 34.8 22.5 316 Closest Main Steam Midpoint between Intakes 60.5 12.0 5.5 316 Line Point Closest Main Steam CB intake (Normal) 14.5 12.0 22.5 316 Line Point Emergency Personnel

'B' CB intake (Emergency) 88.4 4.3 5.5 328 Access Hatch Containment CB intake (Normal) 9.3 34.8 22.5 8 (Diffuse) (Note 1)

Containment CB Intake (Emergency) 47.6 34.8 5.5 339 (Diffuse) Midpoint Notes:

1. As stated in Item 3.4 of RG 1.194, ARCON96 should not be used when the horizontal distance is less than 10 meters. A conservative estimate was used by running two additional cases with the horizontal distance set arbitrarily to 10 meters and 20 meters, and extrapolating using a 1/r2 shape.

ULNRC-06636 Enclosure 6 Page 373 of 374 Table 2.3-98 Calculated /Q values and associated events Event(s) (Receptor Location)

Release Source Time (hours) /Q (sec/m3)

Control Room LOCA Containment Leakage, Rod Ejection (Diffuse Containment) 0 - Isolation 7.12E-03 Isolation - 2 7.49E-04 2-8 5.32E-04 8 - 24 2.29E-04 24 - 96 1.50E-04 96 - 720 9.56E-05 1

LOCA Mini-Purge & ECCS Leakage (Unit Vent Exhaust) 0 - Isolation 1.90E-03 Isolation - 2 6.86E-04 2-8 5.72E-04 8 - 24 2.32E-04 24 - 96 1.42E-04 96 - 720 9.57E-05 Letdown Line Break1 (Unit Vent Exhaust) 0-2 1.90E-03 2-8 1.58E-03 8 - 24 6.67E-04 24 - 96 3.90E-04 96 - 720 2.29E-04 FHA in the Fuel Handling Building (Unit Vent Exhaust )

2 0 - Isolation 2.23E-03 Isolation - 2 6.86E-04 2-8 5.72E-04 8 - 24 2.32E-04 24 - 96 1.42E-04 96 - 720 9.57E-05 1: In this accident, the control room never isolates, so the normal intake receptor location is used for the entire accident.

2: The Fuel Handling Buliding closest point is used before isolation instead, since it has a higher /Q value.

ULNRC-06636 Enclosure 6 Page 374 of 374 LOCA RWST Backleakage (RWST Vent) 0 - Isolation 9.28E-04 Isolation - 2 7.47E-04 2-8 6.55E-04 8 - 24 2.71E-04 24 - 96 1.52E-04 96 - 720 9.17E-05 FHA in Containment (Emergency Personnel Access Hatch3) 0 - Isolation3 7.12E-03 Isolation - 2 8.61E-04 2-8 7.54E-04 8 - 24 3.22E-04 24 - 96 1.84E-04 96 - 720 1.43E-04 Locked Rotor, SGTR (Closest ASD4) 0 - Isolation4 1.76E-02 Isolation - 2 1.74E-03 2-8 1.33E-03 8 - 24 6.50E-04 24 - 96 3.62E-04 96 - 720 2.96E-04 LOOP1 (MSSV) 0-2 1.76E-02 2-8 1.46E-02 8 - 24 6.74E-03 24 - 96 3.81E-03 96 - 720 3.05E-03 MSLB (Closest MSL Point5) 0 - Isolation5 1.76E-02 Isolation - 25 1.74E-03 2-8 1.56E-03 8 - 24 6.61E-04 24 - 96 3.83E-04 96 - 720 3.22E-04 3

Diffuse leakage through the containment wall is used before isolation instead, since it has a higher /Q value.

4

The closest MSSV is used before isolation instead, since it has a higher /Q value.

5: The closest MSSV is used before isolation instead, since it has a higher /Q value to the normal intake stack. The closest ASD is used for the first two hours after isolation, since it has a higher /Q value for that time frame.