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NUREG-0017, Revision 2 Calculation of Release of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized-Water Reactors
	ML20213C729
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Issue date:	07/31/2020
From:	Colameco D, Geelhood K, John Tomon; Office of Nuclear Regulatory Research, Pacific Northwest National Laboratory
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	Malone, Tina
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	NUREG-0017 R02
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NUREG-0017, Revision 2 Calculation of Release of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized-Water Reactors GALE-PWR 3.2 Code Office of Nuclear Regulatory Research

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NUREG-0017, Revision 2 Calculation of Release of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized-Water Reactors GALE-PWR 3.2 Code Manuscript Completed: February 2020 Date Published: July 2020 Prepared by J. J. Tomon K. J. Geelhood1 D. V. Colameco1 1

Pacific Northwest National Laboratory Richland, WA 99352 John Tomon, NRC Project Manager Office of Nuclear Regulatory Research

ABSTRACT This report is documentation for the release of the GALE-PWR 3.2 (Gaseous and Liquid EffluentsPressurized-Water Reactor 3.2) code. The GALE-PWR-3.2 code is a computerized mathematical model for calculating the releases of radioactive material in gaseous and liquid effluents (i.e., the gaseous and liquid source terms) from nuclear power plants (NPPs). The code is a tool that can be used to determine compliance with the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix I, Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the Criterion As Low as Is Reasonably Achievable for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents.

For domestic and international applicants submitting a new reactor application for review and approval by the U.S. Nuclear Regulatory Commission (NRC), the staff considers use of the following to be an acceptable method for demonstrating compliance with 10 CFR Part 50, Appendix I:

GALE86 code

DC/COL-ISG-05, Interim Staff Guidance on the Use of the GALE86 Code for Calculation of Routine Radioactive Releases in Gaseous and Liquid Effluents from Boiling-Water Reactors and Pressurized-Water Reactors To Support Design Certification and Combined License Applications, issued July 2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML081710299)

Regulatory Guide (RG) 1.112, Revision 1, Calculations of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Light-Water-Cooled Nuclear Power Reactors, issued March 2007 (ADAMS Accession No. ML070320241)

When the source term or reactor design parameters differ from those given in the GALE86 code, DC/COL-ISG-05, and RG 1.112, Revision 1, they should be described with sufficient detail, and the basis of the alternate method, model parameters, and assumptions should be provided, to allow the NRC to conduct an independent evaluation.

The purpose of the release of this version of the GALE-PWR code is to comprehensively verify the applicability of the current methodology described in NUREG-0017, Revision 1 (ADAMS Accession No. ML112720411), Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE code), issued April 1985, to both the current U.S. PWR facilities and proposed NPP designs. Additionally, the GALE-PWR 3.2 code includes a graphical user interface (GUI) to facilitate easier use of the code.

This report documents the updated GALE-PWR 3.2 code, which is in a self-contained executable and includes installation instructions and description of the various screen shots of the GALE-PWR 3.2 code GUI.

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TABLE OF CONTENTS ABSTRACT ................................................................................................................... iii TABLE OF CONTENTS.................................................................................................. v LIST OF FIGURES......................................................................................................... ix LIST OF TABLES ........................................................................................................ xiii ACKNOWLEDGMENTS ............................................................................................... xv ABBREVIATIONS....................................................................................................... xvii

1.0 INTRODUCTION

............................................................................................... 1-1 1.1 GALE-PWR 3.2 Code Update ...........................................................................1-1 1.1.1 Graphical User Interface........................................................................1-2 1.1.2 Reactor Coolant Source Term (ANS-18.1 Version) Options ..................1-2 1.1.3 GALE-PWR Fixed Modeling Parameters (GALE Version) Options ........1-3 1.2 Software Quality Assurance and Configuration Management Plans ..................1-3 2.0 DEFINITIONS AND TERMS.............................................................................. 2-1 2.1 Code Source Computation ................................................................................2-1 2.2 Definitions .........................................................................................................2-2 2.2.1 Plant Structures .....................................................................................2-2 2.2.2 Terminology ..........................................................................................2-4 2.3 Gaseous Source Terms ....................................................................................2-6 2.4 Liquid Source Terms .........................................................................................2-7 3.0 INSTALLATION AND USE ............................................................................... 3-1 3.1 Code Installation ...............................................................................................3-1 3.2 Code Execution and Use ..................................................................................3-2 3.2.1 Introductory GUI Inputs .........................................................................3-2 3.2.2 General Reactor Parameters GUI Inputs ...............................................3-7 v

3.2.3 Liquid GUI Inputs (Liquid Radwaste Treatment System Input Screen) ...............................................................................................3-18 3.2.4 Gaseous GUI Inputs (Gaseous Radwaste Treatment System Input Screen) ...............................................................................................3-46 vi

3.3 Viewing the Code Outputs ..............................................................................3-67 4.0 MODELING PARAMETERS ............................................................................. 4-1 4.1 GALE Version (Fixed Model Parameters) .........................................................4-1 4.1.1 GALE86 Fixed Modeling Parameters ....................................................4-1 vii

4.2 ANS-18.1 Version (Source Term Options) ......................................................4-57 4.2.1 ANSI/ANS-18.1-1999 Source Term Parameters ..................................4-58 4.2.2 ANSI/ANS-18.1-1984 Source Term Parameters ..................................4-82 4.2.3 ANSI/ANS-18.1-2016 Source Term Parameters ..................................4-89 4.3 PWRfixed-parameters.txt file...........................................................................4-96

5.0 REFERENCES

.................................................................................................. 5-1 viii

LIST OF FIGURES Figure 2-1 Principal structures at an operating PWR nuclear power generation station..... 2-4 Figure 3-1 Windows Explorer directory with the GALE-PWR 3.2 files ............................... 3-1 Figure 3-2 Introductory Screen.......................................................................................... 3-3 Figure 3-3 Introductory ScreenGALE version options .................................................... 3-4 Figure 3-4 Introductory ScreenANS-18.1 version options .............................................. 3-5 Figure 3-5 Introductory Screenfile output/input and unit options .................................... 3-6 Figure 3-6 General Reactor Parameters Screen ............................................................... 3-7 Figure 3-7 Read from File Screen ..................................................................................... 3-8 Figure 3-8 General Reactor Parameters Screen with default values from the PWRGALE.in file.............................................................................................. 3-9 Figure 3-9 Basic structure of a PWR ............................................................................... 3-10 Figure 3-10 Name of Reactor Allowable Range Screen .................................................... 3-10 Figure 3-11 Thermal Power Level Allowable Range Screen ............................................. 3-11 Figure 3-12 Mass of Coolant in Primary System Allowable Range Screen ........................ 3-11 Figure 3-13 Primary Letdown Rate Allowable Range Screen ............................................ 3-12 Figure 3-14 Letdown Cation Demineralizer Flowrate Allowable Range Screen ................. 3-13 Figure 3-15 Number of Steam Generators Allowable Range Screen................................. 3-13 Figure 3-16 Total Steam Flow Allowable Range Screen ................................................... 3-14 Figure 3-17 Mass of Liquid in Each Steam Generator Allowable Range Screen ............... 3-14 Figure 3-18 Cutaway views of Westinghouse and Combustion Engineering U-tube steam generators ............................................................................... 3-15 Figure 3-19 Cutaway view of Babcock and Wilcox once-through steam generator............ 3-16 Figure 3-20 U-Tube Total Blowdown Rate Allowable Range Screen ................................. 3-17 Figure 3-21 Condensate Demineralizer Regeneration Time Allowable Range Screen ...... 3-17 Figure 3-22 Fraction of Feedwater through Condensate Demineralizers Allowable Range Screen ................................................................................ 3-18 Figure 3-23 Liquid Radwaste Treatment System Screen (Shim Bleed Tab) ...................... 3-18 Figure 3-24 Shim Bleed Liquid Stream Flow Rate Allowable Range Screen ..................... 3-19 Figure 3-25 Shim Bleed Decontamination Factor Allowable Range Screens..................... 3-20 Figure 3-26 Waste Collection and Processing Allowable Range Screens ......................... 3-22 Figure 3-27 Waste Collection Time, and Processing and Discharge Time Calculation Screen ......................................................................................... 3-23 Figure 3-28 Waste Collection Time and Processing and Discharge Time Calculation Allowable Range Screens .............................................................................. 3-24 Figure 3-29 Example of liquid waste processing system flow stream scheme ................... 3-26 Figure 3-30 Liquid Radwaste Treatment System Screen (Equipment Drain Waste Tab) .................................................................................................... 3-27 Figure 3-31 Equipment Drain Waste Stream Flow Rate Allowable Range Screen ............ 3-27 Figure 3-32 Equipment Drain Waste Activity of Inlet Stream Allowable Range Screen ..... 3-28 Figure 3-33 Equipment Drain Waste Screen ..................................................................... 3-29 Figure 3-34 Equipment Drain Waste Decontamination Factor Allowable Range Screens .............................................................................................. 3-30 Figure 3-35 Liquid Radwaste Treatment System Screen (Clean Waste Tab) .................... 3-31 Figure 3-36 Clean Waste Stream Flow Rate Allowable Range Screen ............................. 3-32 Figure 3-37 Clean Waste Activity of Inlet Stream Allowable Range Screen....................... 3-32 Figure 3-38 Clean Waste (deaerated or tritiated) Screen .................................................. 3-33 Figure 3-39 Clean Waste Decontamination Factor Allowable Range Screens .................. 3-34 Figure 3-40 Liquid Radwaste Treatment System Screen (Dirty Waste Tab) ...................... 3-35 ix

Figure 3-41 Dirty Waste Stream Flow Rate Allowable Range Screen ............................... 3-36 Figure 3-42 Dirty Waste Activity of Inlet Stream Allowable Range Screen ........................ 3-36 Figure 3-43 Miscellaneous Dirty Waste (aerated or non-tritiated) Screen .......................... 3-37 Figure 3-44 Dirty Waste Decontamination Factor Allowable Range Screens .................... 3-38 Figure 3-45 Liquid Radwaste Treatment System Screen (Blowdown Waste Tab) ............. 3-39 Figure 3-46 Blowdown Waste Fraction of Steam Processed Allowable Range Screen ..... 3-40 Figure 3-47 Blowdown Waste Decontamination Factor Allowable Range Screens............ 3-41 Figure 3-48 Liquid Radwaste Treatment System Screen (Regenerant Waste Tab)........... 3-42 Figure 3-49 Regenerant Waste Stream Flow Rate Allowable Range Screen .................... 3-42 Figure 3-50 Regenerant Waste Decontamination Factor Allowable Range Screens ......... 3-44 Figure 3-51 Liquid Radwaste Treatment System Screen (Detergent Waste Tab).............. 3-45 Figure 3-52 Detergent Waste Partition Factor Allowable Range Screen ........................... 3-45 Figure 3-53 Gaseous Radwaste Treatment System Screen .............................................. 3-46 Figure 3-54 Gaseous Radwaste Treatment System Screen with Letdown System field options.................................................................................................... 3-47 Figure 3-55 Gaseous Radwaste Treatment System Screen with holdup time for fission gases stripped from the primary coolant field ...................................... 3-48 Figure 3-56 Holdup Time for Xe Allowable Range Screen ................................................ 3-49 Figure 3-57 Holdup Time for Kr Allowable Range Screen ................................................. 3-49 Figure 3-58 Fill Time of Decay Tanks for the Gas Stripper Allowable Range Screen ........ 3-50 Figure 3-59 Gaseous Radwaste Treatment System Screen with Iodine field .................... 3-51 Figure 3-60 Fraction of Iodine Released from Blowdown Tank Vent Allowable Range Screen ................................................................................................ 3-51 Figure 3-61 Diagram of turbine condenser (a) and steam-jet air ejector (b)....................... 3-52 Figure 3-62 Percent of Iodine Removed from Air Ejector Release Allowable Range Screen ................................................................................................ 3-52 Figure 3-63 Gaseous Radwaste Treatment System Screen with Waste Gas System Particulate Release Tab options .................................................................... 3-53 Figure 3-64 Gaseous Radwaste Treatment System Screen with Fuel Handling Building Tab options ...................................................................................... 3-54 Figure 3-65 NUREG-0017, Revision 1 Charcoal Efficiency Information Screen ................ 3-55 Figure 3-66 Regulatory Guide 1.140, Revision 3 Charcoal Efficiency Information Screen......................................................................................... 3-56 Figure 3-67 Fuel Handling Building Charcoal Adsorber Removal Efficiency Allowable Range Screen ................................................................................ 3-57 Figure 3-68 Gaseous Radwaste Treatment System Screen with Auxiliary Building Tab options .................................................................................................... 3-58 Figure 3-69 Auxiliary Building Charcoal Adsorber Removal Efficiency Allowable Range Screen ................................................................................................ 3-58 Figure 3-70 Gaseous Radwaste Treatment System Screen with Containment Building Tab options ...................................................................................... 3-59 Figure 3-71 Containment Building Free Volume Allowable Range Screen ........................ 3-60 Figure 3-72 Containment Building Charcoal Adsorber Removal Efficiency Allowable Range Screen ................................................................................ 3-60 Figure 3-73 Containment Building Internal Cleanup System Flow Rate Allowable Range Screen ................................................................................................ 3-61 Figure 3-74 Gaseous Radwaste Treatment System Screen with Containment High Volume Purge Tab options .................................................................... 3-62 Figure 3-75 Containment High Volume Purge Charcoal Adsorber Removal Efficiency Allowable Range Screen ................................................................................ 3-62 x

Figure 3-76 Containment High Volume Purge Number of Purges Allowable Range Screen ................................................................................................ 3-63 Figure 3-77 Gaseous Radwaste Treatment System Screen with Containment Low Volume Purge Tab options ..................................................................... 3-64 Figure 3-78 Containment Low Volume Purge Charcoal Adsorber Removal Efficiency Allowable Range Screen ................................................................................ 3-64 Figure 3-79 Containment Low Volume Purge Continuous Purge Rate Allowable Range Screen ................................................................................................ 3-65 Figure 3-80 GALE code completion dialog box ................................................................. 3-66 Figure 3-81 Windows Explorer directory with GALE-PWR 3.2 input and output files ......... 3-66 Figure 3-82 PWRGE.out file .............................................................................................. 3-68 Figure 3-83 PWRLE.out file .............................................................................................. 3-69 Figure 3-84 GALE PWR Output 3.2.xls file........................................................................ 3-70 Figure 4-1 Krypton and xenon K values as a function of reciprocal temperature ............. 4-28 Figure 4-2 Effect of moisture content on the dynamic adsorption coefficient ................... 4-29 Figure 4-3 Charcoal moisture as a function of relative humidity ...................................... 4-29 Figure 4-4 Removal paths for the reference PWR with U-tube generators ...................... 4-69 Figure 4-5 Removal paths for the reference PWR with once-through steam generators...................................................................................................... 4-70 Figure 4-6 PWRfixed-parameters.txt file.......................................................................... 4-96 Figure 4-7 Example of a modified PWRfixed-parameters.txt file...................................... 4-99 xi

LIST OF TABLES Table 3-1 Seven PWR liquid radwaste inlet stream tabs and their sources .................... 3-19 Table 3-2 Summary of typical decontamination factors for PWR liquid waste treatment streams ......................................................................................... 3-21 Table 4-1 Plant capacity factors at operating PWRs ........................................................ 4-2 Table 4-2 Radioiodine releases from building ventilation systems before treatment ......... 4-4 Table 4-3 Annual radioiodine normalized releases from containment ventilation systems ........................................................................................................... 4-5 Table 4-4 Annual radioiodine normalized releases from auxiliary building ventilation systems ........................................................................................................... 4-6 Table 4-5 Annual radioiodine normalized releases from refueling area ventilation systems ........................................................................................................... 4-7 Table 4-6 Annual radioiodine normalized releases from turbine building ventilation systems ........................................................................................................... 4-8 Table 4-7 Fraction of radioiodine appearing in each chemical form from PWR building ventilation systems ........................................................................... 4-10 Table 4-8 Particulate release rate for gaseous effluents ................................................ 4-11 Table 4-9 Measured releases upstream of HEPA filterscontainment .......................... 4-12 Table 4-10 Measured releases upstream of HEPA filtersauxiliary building ................... 4-13 Table 4-11 Measured releases upstream of HEPA filtersfuel pool area ........................ 4-14 Table 4-12 Measured releases upstream of HEPA filterswaste gas system ................. 4-15 Table 4-13 Annual radioiodine normalized releases from main condenser air ejector exhaust .............................................................................................. 4-18 Table 4-14 PWR containment purging and venting experience ........................................ 4-20 Table 4-15 Removal efficiencies for charcoal adsorbers for iodine removal ..................... 4-23 Table 4-16 Data from the R.E. Ginna containment building atmosphere test ................... 4-24 Table 4-17 Waste gas system input flow to pressurized storage tanks for PWRs without recombiners ....................................................................................... 4-26 Table 4-18 Waste gas system input flow to pressurized storage tanks for PWRs with recombiners ................................................................................................... 4-26 Table 4-19 Dynamic adsorption coefficient for xenon and krypton ................................... 4-27 Table 4-20 PWR liquid wastes ......................................................................................... 4-31 Table 4-21 Calculated annual release of radioactive materials in untreated detergent waste ............................................................................................. 4-32 Table 4-22 Radionuclide distribution of detergent waste .................................................. 4-33 Table 4-23 Tritium release data from PWRs with Zircaloy-clad fuels................................ 4-36 Table 4-24 Tritium release rate from operating PWRs ..................................................... 4-40 Table 4-25 Total tritium release data from operating PWRs (percent) .............................. 4-41 Table 4-26 Distribution of tritium release gaseous effluents ............................................. 4-42 Table 4-27 Decontamination factors for demineralizers ................................................... 4-43 Table 4-28 Decontamination factors for evaporators........................................................ 4-44 Table 4-29 Reverse osmosis decontamination factors for R.E. Ginna.............................. 4-46 Table 4-30 Reverse osmosis decontamination factors for Point Beach ............................ 4-47 Table 4-31 Reverse osmosis decontamination factors for H.B. Robinson 2 ..................... 4-47 Table 4-32 Expected reverse osmosis decontamination factors for specific nuclides ....... 4-48 Table 4-33 Frequency and extent of unplanned liquid radwaste releases from operating plants ............................................................................................. 4-50 Table 4-34 Carbon-14 release data from operating PWRs............................................... 4-53 Table 4-35 Distribution of carbon-14 released in gaseous effluents ................................. 4-54 xiii

Table 4-36 Summary of argon-41 releases from operating PWRs ................................... 4-54 Table 4-37 Radionuclide concentrations in PWRs with U-tube steam generators

- ANSI/ANS-18.1-1999 .................................................................................. 4-59 Table 4-38 Radionuclide concentrations in PWRs with once-through steam generators

- ANSI/ANS-18.1-1999 .................................................................................. 4-61 Table 4-39 Parameters used to describe the reference PWR with U-tube steam generators...................................................................................................... 4-63 Table 4-40 Parameters used to describe the reference PWR with once-through steam generators...................................................................................................... 4-64 Table 4-41 Values used in determining adjustment factors for PWRs .............................. 4-65 Table 4-42 Adjustment factors for PWRs with U-tube steam generators

- ANSI/ANS-18.1-1999 .................................................................................. 4-67 Table 4-43 Adjustment factors for PWRs with once-through steam generators

- ANSI/ANS-18.1-1999 .................................................................................. 4-68 Table 4-44 Sample case parameters versus the range of values shown in Table 4-39 .... 4-71 Table 4-45 Monthly average primary/secondary leakage ................................................. 4-75 Table 4-46 Moisture carryover in recirculating U-tube steam generators ......................... 4-81 Table 4-47 Radionuclide concentrations in PWRs with U-tube steam generators

- ANSI/ANS-18.1-1984 .................................................................................. 4-83 Table 4-48 Radionuclide concentrations in PWRs with once-through steam generators - ANSI/ANS-18.1-1984 ................................................................ 4-85 Table 4-49 Adjustment factors for PWRs with U-tube steam generators

- ANSI/ANS-18.1-1984 .................................................................................. 4-87 Table 4-50 Adjustment factors for PWRs with once-through steam generators

- ANSI/ANS-18.1-1984 .................................................................................. 4-88 Table 4-51 Radionuclide concentrations in PWRs with U-tube steam generators

- ANSI/ANS-18.1-2016 .................................................................................. 4-90 Table 4-52 Radionuclide concentrations in PWRs with once-through steam generators

- ANSI/ANS-18.1-2016 .................................................................................. 4-92 Table 4-53 Adjustment factors for PWRs with U-tube steam generators

- ANSI/ANS-18.1-2016 .................................................................................. 4-94 Table 4-54 Adjustment factors for PWRs with once-through steam generators

- ANSI/ANS-18.1-2016 .................................................................................. 4-95 Table 4-55 PWRfixed-parameters.txt file modifiable parameters ...................................... 4-97 xiv

ACKNOWLEDGMENTS The improvements to the GALE-PWR 3.2 code would not have been possible without the technical guidance, feedback, and help of the U.S. Nuclear Regulatory Commission (NRC) staff from the Office of New Reactors, Office of Nuclear Reactor Regulations, and the Office of Nuclear Regulatory Research. Specifically, NRC staff: Dr. Richard Clement, Dr. Elijah Dickson, Zachary Gran, and John Tomon. Additionally, the authors thank the staff of Pacific Northwest National Laboratory for its support and efforts in updating the GALE-PWR 3.2 code.

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ABBREVIATIONS ANS American Nuclear Society ANSI American National Standards Institute AOO anticipated operational occurrence Ar argon B&W Babcock and Wilcox BWR boiling-water reactor C carbon CE Combustion Engineering CFR Code of Federal Regulations Ci curie Ci/yr curies per year cm3/g cubic centimeters per gram Co cobalt Cs cesium DCD design certification document DF decontamination factor EPA U.S. Environmental Protection Agency EPRI Electric Power Research Institute

°F Fahrenheit Fortran Formula Translation (formerly FORTRAN)

FSAR final safety analysis report ft3 cubic feet ft3/d cubic feet per day ft3/min cubic feet per minute ft3/yr cubic feet per year gal gallons gal/d gallons per day gal/min gallons per minute GALE Gaseous and Liquid Effluents computer code GALE-PWR 3.2 Gaseous and Liquid EffluentsPressurized-Water Reactor computer code, version 3.2 GUI graphical user interface GWd gigawatt day h hour H-3 tritium HEPA high-efficiency particulate air (filter)

HOI hypoiodous acid I iodine ISG interim staff guidance kg kilogram kg/h kilograms per hour Kr krypton lb pound lb/h pounds per hour L/d liters per day L/min liters per minute LWR light water reactor m3 cubic meter xvii

m3/min cubic meters per minute MTU metric ton of uranium MWd megawatt day MWe megawatt electric MWt megawatt thermal

µCi/cm3 microcuries per cubic centimeter

µCi/g microcuries per gram

µCi/ml microcuries per milliliter NPP nuclear power plant NRC U.S. Nuclear Regulatory Commission NSS nuclear steam system NUREG U.S. Nuclear Regulatory Commission technical report designation ORNL Oak Ridge National Laboratory PC partition coefficient PCA primary coolant activity PNNL Pacific Northwest National Laboratory PWR pressurized-water reactor PWRGE pressurized-water reactor gaseous effluent model PWRLE pressurized-water reactor liquid effluent model RCTS reactor coolant treatment system Rb rubidium RESAR reference safety analysis report RG regulatory guide SI International System of Units SMR small modular reactor SQAP software quality assurance plan Sr strontium STP standard temperature and pressure URC ultrasonic resin cleaner WGPS waste gas processing system Xe xenon Zn zinc xviii

1.0 INTRODUCTION

In promulgating Appendix I, Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the Criterion As Low as Is Reasonably Achievable for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents, to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, [Ref. 1], the U.S. Nuclear Regulatory Commission (NRC) indicated its desire for improving the calculation models used by the staff to determine conformance with the requirements of the regulation. To conform to the requirements of Appendix I to 10 CFR Part 50, the NRC developed the GALE-PWR (Gaseous and Liquid Effluents Pressurized-Water Reactor) code as a computerized mathematical model for calculating the releases of radioactive material in gaseous and liquid effluents from pressurized-water reactors (PWRs). The original NUREG-0017, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized-Water Reactors (PWR-GALE Code), issued April 1976 [Ref. 2] and its Revision 1, issued April 1985 [Ref. 3], documented this version of the GALE-PWR code, referred to as GALE-PWR 86 (GALE86).

The NRC staff considers the following as an acceptable method for demonstrating compliance with 10 CFR Part 50, Appendix I:

GALE86 code

DC/COL-ISG-05, Interim Staff Guidance on the Use of the GALE86 Code for Calculation of Routine Radioactive Releases in Gaseous and Liquid Effluents from Boiling-Water Reactors and Pressurized-Water Reactors to Support Design Certification and Combined License Applications, issued July 2008 [Ref. 4]

Regulatory Guide (RG) 1.112, Revision 1, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Light-Water-Cooled Nuclear Power Reactors, issued March 2007 [Ref. 5]

When the source term or reactor design parameters differ from those given in the GALE86 code, DC/COL-ISG-05, and RG 1.112, Revision 1, they should be described with sufficient detail, and the basis of the alternate method, model parameters, and assumptions should be provided, to allow the staff to conduct an independent evaluation. Additionally, the licensing guidance of RG 1.112, Revision 1, and NUREG-0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, issued March 2007

[Ref. 6], reference the GALE86 code.

This report describes the GALE-PWR 3.2 code and the work Pacific Northwest National Laboratory (PNNL) performed on the GALE86 code. The GALE-PWR 3.2 code is an update to the GALE86 code, and it maintains direct traceability back to the GALE86 code. The GALE-PWR 3.2 code is available for download from the NRCs Radiation Protection Computer Code Analysis and Maintenance Program (RAMP) Web site (https://ramp.nrc-gateway.gov/).

1.1 GALE-PWR 3.2 Code Update The GALE-PWR 3.2 code version updates the GALE86 version. The GALE86 version of the code based on NUREG-0017, Revision 1, contained two modulesthe pressurized-water reactor liquid effluent (PWRLE86) and the pressurized-water reactor gaseous effluent 1-1

(PWRGE86), based on the reactor coolant source term (tables and adjustment equations) described in American National Standards Institute/American Nuclear Society (ANSI/ANS)-18.1-1984, Radioactive Source Term for Normal Operation of Light Water Reactors [Ref. 7].

The GALE-PWR 3.2 code version maintains the input/output functionality of previous versions including the Fortran card-based format input file. A new graphical user interface (GUI) function allowing for direct creation of the input file was added to simplify the data input process.

In one technical change made to GALE-PWR 3.2, the iodine isotopes iodine-132, iodine-134, and iodine-135 were added to the PWRGE code assuming the primary and secondary coolant activities given in the appropriate ANSI/ANS-18.1 tables. The decay constants for these isotopes were taken from the Isotope Generation and Depletion Code (ORIGEN) database in the PWRLE code. The release relative to the primary coolant activities from various buildings was assumed to be the same for all iodine isotopes consistent with the previous treatment of iodine-131 and iodine-133.

1.1.1 Graphical User Interface The GALE-PWR 3.2 code was designed and developed to be a flexible tool providing the same functionality of the GALE86 code with additional features available to the user. In the original version of the GALE86 code, the user was required to enter data using the Fortran card-based format. A Visual Basic GUI, developed for the GALE-PWR 3.2 code, allows easier input of plant-specific parameters; Chapter 3 describes this GUI. Additionally, with the development of the GALE-PWR 3.2 GUI, other features were now easier to add to the original GALE86 code.

These additional features include the options to select the specific reactor coolant source term and the option to modify some of the GALE86 modeling parameters using the PWR fixed parameter text file.

In GALE-PWR 3.2, the values for the reactor coolant source term and GALE-PWR fixed modeling parameters are defaulted to the ANSI/ANS-18.1-1999 source term values and the NUREG-0017, Revision 1, GALE86 code fixed modeling parameters. These GALE-PWR 3.2 default values are consistent with the guidance in DC/COL-ISG-05; RG 1.112, Revision 1; and NUREG-0800. Additionally, because the updates to the GALE-PWR 3.2 source codes did not involve changes in the GALE86 model formulations, the revised source code had the same functionality as the previous version with differences in the outputs reflecting only reactor coolant source term and the GALE-PWR fixed modeling parameter options available to the user in GALE-PWR 3.2.

1.1.2 Reactor Coolant Source Term (ANS-18.1 Version) Options As stated above, GALE-PWR 3.2 gives the user the option of selecting the reactor coolant source term (ANS-18.1 Version) desired for the GALE-PWR 3.2 calculations. The GALE-PWR 3.2 code provides three options for the reactor coolant source terms, including the default option of ANSI/ANS-18.1-1999, which is consistent with the guidance in DC/COL-ISG-05; RG 1.112, Revision 1; and NUREG-0800. Section 4.2.1 describes the parameters used by GALE-PWR 3.2 when the default reactor coolant source term option is selected.

The other reactor coolant source term options available in the GALE-PWR 3.2 code are ANSI/ANS-18.1-1984, Radioactive Source Term for Normal Operation of Light Water Reactors

[Ref. 8] and ANSI/ANS-18.1-2016 [Ref. 9]. Section 4.2.2 describes the parameters used by 1-2

GALE-PWR 3.2 when the ANSI/ANS-18.1-1984 reactor coolant source term option is selected.

Section 4.2.3 describes the parameters used by GALE-PWR 3.2 when the ANSI/ANS-18.1-2016 reactor coolant source term option is selected. When the source term or reactor design parameters differ from those given in the GALE86 code, DC/COL-ISG-05, and RG 1.112, Revision 1, they should be described with sufficient detail, and the basis of the alternate method, model parameters, and assumptions should be provided, to allow the NRC to conduct an independent evaluation.

1.1.3 GALE-PWR Fixed Modeling Parameters (GALE Version) Options The GALE-PWR 3.2 includes the fixed modeling parameters form the GALE86 code, which is described in NUREG-0017, Revision 1, and is consistent with the guidance in DC/COL-ISG-05 and RG 1.112, Revision 1. Section 4.1.1 describes the parameters used by GALE-PWR 3.2 with the GALE86 fixed modeling parameters.

The GALE-PWR 3.2 code also comes with a PWR fixed parameter text file, which allows the entry of user-defined values for certain GALE-PWR fixed modeling parameters by means of a text file. Section 4.3 describes the use of the PWRfixed-parameter.txt file. Finally, the GALE-PWR 3.2 code output files have been modified to identify any changes made to the default GALE-PWR 3.2 inputs (GALE86 and ANSI/ANS-18.1-1999).

1.2 Software Quality Assurance and Configuration Management Plans In addition to the updates described above, PNNL-24249, Software Quality Assurance Plan:

Support for the Gaseous and Liquid Effluent (GALE) Computer Code Project, issued April 2015

[Ref. 10], documents a GALE software quality assurance plan. PNNL-24250, Support for the Gaseous and Liquid Effluent (GALE) Computer Code Project: Configuration Management &

Maintenance Plan, issued April 2015 [Ref. 11], documents the GALE code configuration management and maintenance plan. The NRC has defined three levels of software, per NUREG/BR-0167, Software Quality Assurance Program and Guidelines, issued February 1993 [Ref. 12]:

(1) Level 1technical application software used in a safety decision by the NRC (2) Level 2technical or nontechnical application software not used in a safety decision by the NRC (3) Level 3technical or nontechnical application software not used in a safety decision and having local or limited use by the NRC The quality assurance documents are written to conform to the Level 2 requirements. Code development on GALE-PWR 3.2 has proceeded under the SQAP and code CMMP. PNNL-28818, GALE-3.2 Verification Report, issued June 2019 [Ref. 13], describes the work done to verify proper implementation of the new coding.

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1 2.0 DEFINITIONS AND TERMS This chapter describes the GALE-PWR 3.2 code source computations that estimate airborne radionuclide emission rates and waterborne radionuclide effluent rates. It also presents guidance on defining these input parameters based on specific plant characteristics.

2.1 Code Source Computation The average quantity of radioactive material released to the environment from a nuclear power reactor during normal operation, including AOOs is called the source term. The calculations performed by the GALE-PWR 3.2 code using the default values are based on (1)

ANSI/ANS-18.1-1999 tables and adjustment factors, (2) the release and transport mechanisms that result in the appearance of radioactive material in liquid and gaseous waste streams, (3) plant-specific design features used to reduce the quantities of radioactive materials ultimately released to the environment, and (4) NUREG-0017, Revision 1, information received on the operation of nuclear power plants (NPPs).

In a PWR, primary coolant water circulates through the reactor core where it removes the heat from the fuel elements. In the steam generators, heat from the pressurized primary coolant water is transferred to the secondary coolant water to form steam. The steam expands through the turbine and is then condensed and returned to the steam generators. The primary coolant water flows back to the reactor core. The principal mechanisms that affect the concentrations of radioactive materials in the primary coolant are (1) fission product leakage to the coolant from defects in the fuel cladding and fission product generation in tramp uranium, (2) corrosion products activated in the core, (3) radioactivity removed in the reactor coolant treatment systems (RCTS), and (4) activity removed because of primary coolant leakage. The following paragraphs briefly describe these mechanisms.

The primary coolant is continuously purified by passing a side stream through filters and demineralizers in the RCTS. It is necessary to maintain the purity of the primary coolant to prevent fouling of heat transfer surfaces and to keep releases to the environment as low as is reasonably achievable. Chemicals are added to the primary coolant to inhibit corrosion, improve fuel economy, or both. Lithium hydroxide is added for pH control to reduce corrosion.

Water decomposes into oxygen and hydrogen as a result of radiolysis. The control of oxygen concentration in the primary coolant is important for corrosion control. Hydrogen, added to the primary coolant as dissolved free hydrogen, tends to force the net reaction toward the recombination of hydrogen and oxygen to water at an overall rate sufficient to maintain low primary coolant oxygen concentrations.

Boron is added to the primary coolant as a neutron absorber (shim control). As the fuel cycle progresses, boron is removed from the primary coolant through the RCTS loop (shim bleed).

The shim bleed is processed through an evaporator, and the boron in the evaporator bottoms is either reused or packaged as solid waste. The evaporator distillate may be recycled to the reactor coolant system as makeup water or discharged to the environment.

Radioactive gases stripped from the primary coolant by degasification are normally collected in pressurized storage tanks and held for radioactive decay before recycling or release to the environment. Alternative treatment methods include charcoal delay systems and cryogenic distillation.

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Because of leakage through valve stems and pump shaft seals, some coolant escapes into the containment and the auxiliary buildings. A portion of the leakage evaporates, thus contributing to the gaseous source term, and a fraction remains as liquid, becoming part of the liquid source term. The relative amount of leakage entering the gaseous and liquid phases depends on the temperature and pressure at the point where the leakage occurs. Most of the noble gases enter the gas phase, whereas radioiodine partitions into both phases.

Leakage of primary coolant into the secondary coolant in the steam generator is the only source of radioactivity in the secondary coolant system. Water or steam leakage from the secondary system provides significant inputs to the liquid and gaseous radwaste treatment systems.

Steam leakage may be significant to the gaseous source term since the radioactivity released remains in the gas phase.

In a recirculating U-tube steam generator, the nonvolatile radionuclides leaking from the primary coolant concentrate in the liquid phase in the steam generator. The steam generator blowdown rate and condensate demineralizer flow rate control the degree of concentration.

Since there is no liquid reservoir in a once-through steam generator, the primary coolant leakage boils to steam when it enters the secondary side of the steam generator. A condensate demineralizer system maintains secondary coolant purity, and there is no steam generator blowdown. The condensate demineralizer flow rate controls the concentration of radioactivity in the secondary coolant.

Sources of radioactive wastes from the secondary system are the offgases from the turbine condenser, vent gases from the turbine gland seal, liquid and vent gases from the steam generator blowdown, and liquid and gaseous leaks into the turbine building. Liquid wastes also originate from the chemical regeneration of condensate demineralizers in feedwater or condensate systems.

This document describes both the plant-specific input parameters and code-defined parameters that are used in the PWR source term calculations. The plant-specific input parameters are values used in the models that depend on the design and operations of the facility. This document explains the input data required and acceptable means for obtaining these data.

Subsequent chapters address the required source term parameters developed for use with the GALE-PWR 3.2 code and explain the basis for each parameter, documentation of the GALE-PWR 3.2 code models (i.e., computer programs), and information needed to generate source terms that an applicant is required to submit with an application. Data input and output files are provided for base-case runs of the liquid and gaseous effluent computer programs.

2.2 Definitions 2.2.1 Plant Structures The GALE-PWR 3.2 code release rates are based on the sum of estimated annual emissions in different structural components of the plant. Figure 2-1 is a schematic diagram showing containment structure (reactor building), containment building, turbine building, auxiliary building, radwaste building, and fuel handling building. This building nomenclature represents the typical names of the buildings in most current PWRs; however, different PWR designs may use different terminology and different building names (i.e., compound building, fuel building, and reactor building). It may be necessary to compare the functions of the building in a specific design to those described in this section, especially since the outputs will be tailored to the 2-2

buildings described below. As noted in the following discussion, these structures have various sources of radionuclide releases during both operational and shutdown activities. Additionally, depending on the plant design, the auxiliary or reactor building and radwaste building may house shared components of the offgas-recombiner system.

The main buildings are described below.

Containment Building: The airtight building that houses a nuclear reactor and its pressurizer, reactor coolant pumps, steam generator, and other equipment or piping that might otherwise release fission products to the atmosphere in the event of an accident. Such buildings are usually constructed of steel and surrounded by a steel-reinforced concrete structure. The area between the steel and concrete building is called the annulus.

Auxiliary or Reactor Building: A building at an NPP, which is frequently located adjacent to the reactor containment structure and houses most of the auxiliary and safety systems associated with the reactor, such as radioactive waste systems, chemical and volume control systems, and the emergency cooling water system.

Turbine Building: A building that houses the turbine, generator, condenser, and condensate and feedwater systems. Some PWRs in the United States have a structure without the traditional roof and walls.

Fuel Handling Building: A building separate from the containment that is used to store spent fuel assemblies in steel racks in a large storage pool. Casks for shipping or onsite dry storage of spent fuel assemblies will be loaded (or unloaded) in this pool. A new fuel storage area is provided for receipt of new assemblies and storage before they go into the containment and subsequently into the reactor during refueling.

Radwaste Building: A building that houses various systems for processing liquid, solid, and gaseous radioactive wastes generated by the plant.

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Turbine Building Fuel Handling Auxiliary Radwaste Building Building Building Figure 2-1 Principal structures at an operating PWR nuclear power generation station 2.2.2 Terminology The following definitions apply to terms used in this report:

Activation Gases: Gases (including oxygen, nitrogen, and argon) that are radioactive as a result of irradiation in the core.

Anticipated Operational Occurrence (AOO): Unplanned releases of radioactive materials from miscellaneous actions such as equipment failure, operator error, and administrative error that are not consequential enough to be considered an accident. Note that in the context of the GALE-PWR 3.2 code, this definition is different than the one in Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50.

Chemical Waste Stream: Normally, liquids that contain relatively high concentrations of decontaminants, regenerants, or chemical compounds other than detergents. These liquids originate primarily from resin regenerants and laboratory wastes.

Clean Waste System: Normally, tritiated, nonaerated, low-conductivity liquids consisting primarily of liquid waste collected from equipment leaks and drains and certain valve and pump seal leakoffs. These liquids originate from systems containing primary coolant and are normally reused as primary coolant makeup water.

Decontamination Factor (DF): The ratio of the initial amount of a radionuclide in a stream (specified in terms of concentration or activity of radioactive materials) to the final amount of that radionuclide in a stream following treatment by a given process.

Detergent Waste Stream: Liquids that contain detergent, soaps, or similar organic materials.

These liquids consist principally of laundry, personnel shower, and equipment decontamination wastes that normally have low radioactivity content.

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Dirty Waste Stream (Floor Drains): Normally nontritiated, aerated, high conductivity, and nonprimary-coolant quality liquids collected from building sumps and floor and sample station drains. These liquids are not readily amenable for reuse as primary coolant makeup water.

Effective Full-Power Days: The number of days a plant would have to operate at 100-percent licensed power to produce the integrated thermal power output during a calendar year; that is, Integrated Thermal Power Pi Ti Effective Full Power Days = = Ptotal (2-1)

Licensed Power Level where Pi = the ith power level, in megawatt thermal (MWt);

Ptotal = the licensed power level, in MWt; and Ti = the time of operation at power level Pi, in days.

Fission Product: A radionuclide produced either by fission or by subsequent radioactive decay or neutron activation of the radionuclides formed in the fission process.

Gaseous Effluent Stream: Processed gaseous wastes containing radioactive materials resulting from the operation of a nuclear power reactor.

Liquid Effluent Stream: Processed liquid wastes containing radioactive materials resulting from the operation of a nuclear power reactor.

Partition Coefficient (PC): The ratio of the concentration of a radionuclide in the gas phase to the concentration of a radionuclide in the liquid phase when the liquid and gas are at equilibrium.

Partition Factor: The ratio of the quantity of a radionuclide in the gas phase to the total quantity in both the liquid and gas phases when the liquid and gas are at equilibrium.

Plant Capacity Factor: The ratio of the average net power to the rated power capacity.

Primary Coolant: The fluid circulated through the reactor to remove heat. The reactor coolant activity is considered to be constant over a range of power levels, coolant and cleanup flows, and coolant volumes. The radionuclide distributions and concentrations default values for the reactor coolant and main steam are based on ANSI/ANS-18.1-1999, but the GALE-PWR 3.2 code offers additional options to allow the user to select ANSI/ANS-18.1-1984 or ANSI/ANS-18.1-2016. In accordance with the recommendations of the standard, the GALE-PWR 3.2 code provides for adjusting reactor coolant concentrations if the plant is designed to parameters that are outside the ranges considered in the standard. The radionuclides are divided into the following categories:

noble gases

halogens (bromine, radioiodine)

cesium and rubidium

water activation products

tritium

other radionuclides (as listed in GALE-PWR 3.2 code-liquid effluent model output files)

Radioactive Halogens: The radionuclides of fluorine, chlorine, bromine, and radioiodine. The radioactive radionuclides of radioiodine are the key radionuclides considered in dose calculations.

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Radioactive Noble Gases: The GALE-PWR 3.2 code focuses on the radioactive radionuclides of argon, krypton, and xenon, which are characterized by their chemical inactivity. These are the noble gases that are generated by a nuclear reactor or those that are precursors to the production of specific radionuclides. The radioactive radionuclides of argon, krypton, and xenon are the key elements considered in dose calculations.

Radioactive Release Rate: The average quantity of radioactive material released to the environment from a nuclear power reactor during normal operation, including AOOs.

Regenerate Solutions Waste: The liquid solution from ion exchange columns or condensate polishers that are part of the demineralizer system.

Secondary Coolant: The coolant converted to steam by the primary coolant in a heat exchanger (steam generator) to power the turbine. The radionuclide concentrations in the secondary coolant are obtained as discussed above in the definition of primary coolant.

Source Term: The calculated average quantity of radioactive material released to the environment from a nuclear power reactor during normal operation, including AOOs. The source term is the isotopic distribution of radioactive materials used in evaluating the impact of radioactive releases on the environment. Normal operation includes routine outages for maintenance and scheduled refueling.

Steam Generator Blowdown: Liquid intentionally discharged from a steam generator to avoid concentration of impurities during continuing evaporation of steam and to maintain proper water chemistry. This is done only for PWRs with U-tube steam generators.

Tramp Uranium: The uranium present on the cladding of a fuel rod.

Turbine Building Floor Drains: Liquids of high conductivity and low-level radioactivity primarily resulting from secondary system leakage, steam trap drains, sampling system drainage, and maintenance and waste drains.

2.3 Gaseous Source Terms The following sources are considered in calculating the releases of radioactive materials (noble gases, radioactive particulates, and radioiodine) in gaseous effluents from normal operation, including AOOs:

(1) waste gas processing system (2) steam generator blowdown system (3) condenser air ejector exhaust (4) containment purge exhaust (5) ventilation exhaust air from the auxiliary and turbine buildings and the spent fuel pool area (6) steam leakage from the secondary system Depending on the plants design, the gaseous effluent sources listed above may be released from a single stack for the entire plant or via multiple stacks and vents by specific buildings.

The releases of radioactive materials in gaseous effluents from the following sources are 2-6

calculated to be less than 1 curie per year (Ci/yr) of noble gases and 1.0E-04 Ci/yr of iodine-131. Therefore, the following releases are considered negligible:

steam releases resulting from steam dumps to the atmosphere and low-power physics testing, and

ventilation air from buildings not included in item 5 above.

The model computation uses inputs to the waste gas processing system from both continuous stripping of the primary coolant during normal operation and from degassing the primary coolant for two cold shutdowns per year. For plants equipped with steam generator blowdown systems, the model considers radioiodine present in gases leaving the system vent. The GALE-PWR 3.2 code calculates the release rates of noble gases and radioiodine to building atmospheres based on coolant leakage rates to buildings. Radioiodine releases are related to the iodine-131 coolant concentrations for the PWR being evaluated. Particulate release rates are based on measurements at operating PWRs.

The GALE-PWR 3.2 code also calculates tritium releases through the ventilation exhaust systems. The annual quantity of tritium available for release is calculated by using a functional relationship derived from measured liquid and vapor tritium releases at operating PWRs and the integrated thermal power output during the calendar year in which the releases occur. This relationship expresses total tritium as a function of power output. It is assumed that the tritium releases through the ventilation exhaust systems are the total tritium available for release minus the tritium calculated to be released through the liquid pathway.

Later sections provide radioiodine and particulate DFs for removal equipment and parameters for calculating holdup times for noble gases and for calculating tritium, argon-41, and carbon-14 releases.

2.4 Liquid Source Terms The following sources are considered in calculating the release of radioactive materials in liquid effluents from normal operation, including AOOs:

processed water generated from the boron recovery system to maintain plant water balance or for tritium control

processed liquid waste discharged from the dirty waste or miscellaneous waste systems

processed liquid waste discharged from the steam generator blowdown treatment system

processed liquid waste discharged from the chemical waste and condensate demineralizer regeneration system

liquid waste discharged from the turbine building floor drain and sumps

detergent wastes The radioactivity input to the liquid radwaste treatment system is based on the flow rates of the liquid waste streams and their radioactivity levels, expressed as a fraction of the primary coolant activity (PCA). The default values for the PCAs are based on the recommendations of ANSI/ANS-18.1-1999; however, these values can vary based on the ANSI/ANS-18.1 version (1984, 1999, or 2016) selected by the user.

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Radionuclide removal by the liquid radwaste treatment system is based on the following parameters:

decay during collection and processing, and

removal by the proposed treatment systems (e.g., filtration, ion exchange, evaporation, reverse osmosis, plateout).

For PWRs using a deep-bed condensate demineralizer, the inventory of radionuclides collected on the demineralizer resins is calculated by considering the flow rate of condensate at main steam activity that is processed through the demineralizers and radionuclide removal using the DFs given in Sections 4.1.1 and 4.1.2. The activity on the condensate demineralizer resins will also include the steam generator blowdown activity if the blowdown is recycled to the condensate demineralizers. The radioactivity content of the demineralizer regenerant solution is obtained by considering that all radioactivity is removed from the resins at the interval dictated by the regeneration frequency.

Sections 4.1.1 and 4.1.2 describe the methods for calculating collection and processing times and the DFs for radwaste treatment equipment. The liquid radioactive source terms are adjusted to compensate for equipment downtime and AOOs.

For plants using an onsite laundry, a standard detergent waste source term, adjusted for the treatment provided, is added to the adjusted source term.

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3.0 INSTALLATION AND USE This section describes the GALE-PWR 3.2 executable and data files required for operation.

Additionally, this section contains information on how to install the code, how to run the code executable file, and how to view the code output files.

3.1 Code Installation GALE-PWR 3.2 is a Fortran-based computer code used to calculate the effluents released from a PWR during normal operations. GALE-PWR 3.2 can be installed on any computer running Windows operating systems 7.0, 8.0, and 10.0. To install the code, the user should copy the three data files (i.e., actinides.data, fission-products.data, and light-elements.data), the existing PWR input file, the PWR fixed parameter text file, the PWR output Excel file and the executable file to the computer and set up a working directory as shown in Figure 3-1.

Figure 3-1 Windows Explorer directory with the GALE-PWR 3.2 files The names and descriptions of the GALE-PWR 3.2 files to be copied to the working directory are given below:

actinides.dataThis data file is required for the calculation of actinide radionuclides.

The actinides have atomic numbers from 89 to 103 and represent the heavy elements in the fuel. Because of the radioactive decay of the actinides, the data file contains radionuclides with atomic numbers from 81 to 99.

fission-products.dataThis data file is required for the calculation of fission products.

These radionuclides have atomic numbers from 1 to 68 and represent the radionuclides created during the fission of the actinide fuel.

light-elements.dataSimilar to the fission product file, these data are required for the calculation of other elements in the reactor, including impurities in the fuel and other materials associated with the reactor. The atomic numbers range from 1 to 84 in this data file.

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PWRGALE.inThis is sample input for gaseous and liquid effluents from PWRs. This file, which comes with the GALE-PWR 3.2 code download, is a sample PWR input that provides the user with default values.

PWRfixed-parameters.txtThis text file allows the user to change selected fixed modeling GALE-PWR parameters. The user can open and edit this file in any text editor program (e.g., NotePad). The text file should start with $user and end with $end and comments can be included after any !. The default values in this text file are set to ANSI/ANS-18.1-1999 and the GALE86 fixed modeling parameters in NUREG-0017, Revision 1.

PWR GALE Output 3.2.xlsThis is an Excel file to read and display GALE-PWR 3.2 output from PWRs.

GALE_PWR.exeThis is the executable file that opens up the GALE-PWR 3.2 GUI.

This program allows for the generation of plant-specific inputs and the execution of the code.

GALE3.2.icoThis file provides the user with an icon to attach to a desktop shortcut.

The user can create a desktop icon for GALE-PWR 3.2 by right clicking on the GALE_PWR.exe file and selecting the option Create shortcut. The shortcut can then be placed on the desktop, and the icon file can be added by right clicking the desktop shortcut and selecting the properties option on the dropdown menu. Select the Change Icon button in the properties window and navigate to the GALE3.2.ico file shown in Figure 3-1.

Note: It is important to include all of the data files (.data extensions listed in Section 3.1) in the same directory as the input file. Otherwise, the code will give the user an error message stating that it cannot locate the .data files and will terminate the run.

3.2 Code Execution and Use To execute the GALE-PWR 3.2 code, either double click on the executable file, GALE_PWR.exe, or the desktop shortcut icon, and the Introductory Screen for the code will open as shown in Figure 3-2.

3.2.1 Introductory GUI Inputs The Introductory Screen (Figure 3-2) is the GUI screen from which the user selects the name of the input file to be used; the type of analysis to be performed (gaseous, liquid, or both); the GALE version and ANS-18.1 version options to be used; the name of the gaseous and liquid output files; and whether to use legacy inputs in the run.

Input File Name Field On the Introductory Screen, the user should specify the name of the input file to be used for the run. GALE-PWR 3.2 comes with a sample case, the PWRGALE.in file, which appears as the default value for Input File Name. Sections 3.2.2 through 3.2.4 discuss the plant-specific parameter values for this file. The user can also browse for an existing input file that has previously been created by selecting the Browse button in Figure 3-2. This button opens 3-2

Windows Explorer to the default working directory, which contains the GALE-PWR 3.2 executable file (Figure 3-1), but the user can navigate to the directory that contains the input file of interest. If the existing GALE-PWR 3.2 input file is not located in the default working directory noted in Figure 3-1, then all of the data files (.data file extensions listed in Section 3.1) should be in the same directory as the input file. Otherwise the code will give the user an error message stating that it cannot locate the data files and will terminate the run.

Figure 3-2 Introductory Screen Type of Analysis Section This section of the Introductory Screen allows the user the option of selecting the Type of Analysis to be performed by choosing either the Gas, Liquid, or both radio buttons. The default setting, as shown in Figure 3-2, is that both radio buttons are checked.

GALE Version and ANS-18.1 Version Options Figures 3-3 and 3-4 show the GALE version option and the ANS-18.1 version options available to the user. The user also has the option to manually change certain fixed modeling parameters 3-3

through the use of the PWRfixed-parameters.txt file listed in Figure 3-1 and discussed in Section 4.3. Section 4.0 discusses the GALE version option (GALE86), ANS-18.1 version options, and the PWRfixed-parameters.txt file in greater detail.

Note: It is important for the user to recognize that the default values for the GALE version option and the ANS-18.1 version option are the GALE86 code and ANSI/ANS-18.1-1999, which are consistent with the guidance in DC/COL-ISG-05; RG 1.112, Revision 1; and NUREG-0800.

Figure 3-3 Introductory ScreenGALE version options 3-4

Figure 3-4 Introductory ScreenANS-18.1 version options The default ANSI/ANS-18.1-1999 reactor coolant source term option is consistent with the guidance in DC/COL-ISG-05; RG 1.112, Revision 1; and NUREG-0800.

The Introductory Screen contains three additional sections that give the user options to control the Output files, Legacy Input, and English/SI Unit Setting for the code, as shown in Figure 3-5.

Output Files Section The Output Files section gives the user the option to choose the file names for the output files.

The default names for output files are PWRGE.out and PWRLE.out, which will be automatically written to the working directory described in Section 3.1 as shown in Figure 3-1.

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Figure 3-5 Introductory Screenfile output/input and unit options Legacy Input Section The Legacy Input section allows the user to read in legacy GALE-PWR input files (e.g., GALE86 input files) by selecting the Read Legacy File radio button. When the user selects the Read Legacy File radio button, the four input file fields in this section become active and used. As shown in Figure 3-5, the default setting for the Read Legacy File radio button is unchecked with the four input file fields greyed out and prepopulated with the PWRGE86.inp, PWRGE86.out, PWRLE86.inp, and PWRLE86.out in the respective fields. The user should note that when using this option, the legacy files should be located in the working directory described in Section 3.1 and shown in Figure 3-1.

English/SI Unit Setting Section This section gives the user the option to change the input and output units used in the code.

When the user selects the Initialize to all SI Units in Input and Output radio button, all of the 3-6

GALE-PWR inputs and outputs will be in the International System of Units (SI). As shown in Figure 3-5, the default setting for the Initialize to all SI Units in Input and Output radio button is unchecked.

Once the input and output files and the ANS-18.1 version are specified, the user should click the OK button to enter the plant-specific parameters for GALE-PWR 3.2 inputs. These plant-specific parameters are input via three GUI input screens: (1) General Reactor Parameters GUI, (2) Liquid GUI Inputs (Liquid Radwaste Treatment System Input Screen), and (3) Gaseous GUI Inputs (Gaseous Radwaste Treatment System Input Screen).

3.2.2 General Reactor Parameters GUI Inputs This section identifies the plant-specific parameters required and the acceptable ranges of those parameters, if applicable, to be entered in the General Reactor Parameters GUI shown in Figure 3-6. Users are able to read values in from previous input files entered on the Introductory Screen, as well as use the values identified in the sample case in the PWRGALE.in file discussed above. If the user has previously specified an existing input file, the user should click the Read from File button to open the Read from File Screen shown in Figure 3-7.

Figure 3-6 General Reactor Parameters Screen 3-7

Figure 3-7 Read from File Screen Select the Browse button to open a Windows Explorer Screen and navigate to the location of the input file in the working directory described in Section 3.1 and shown in Figure 3-1. Select the PWRGALE.in file on the Windows Explorer Screen and then select the Read from File button in Figure 3-7 to populate the input field boxes with values from the PWRGALE.in file.

These values may be changed and will be updated in the input file once it is run. Figure 3-8 shows the General Reactor Parameters Screen after the Read from File button has been selected with the values from the PWRGALE.in file. The General Reactors Parameter Screen also contains two option buttons for the execution of the GALE-PWR 3.2 code. As shown in Figure 3-8, the user has the option of either saving any changes made to the inputs on this screen by selecting the Save button or executing the GALE calculations by selecting the Run button. The user should note that selecting the Run button without first selecting the Save button will result in the code executing without saving the changes made to the inputs on this screen. Therefore, to save any changes made to the inputs, the user should select the Save button before selecting the Run button.

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Figure 3-8 General Reactor Parameters Screen with default values from the PWRGALE.in file Figure 3-9 shows the basic structure of a PWR. This figure should be used in reference to the items described in the following sections on the general reactor parameters.

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Figure 3-9 Basic structure of a PWR Name of Reactor Enter the name of the reactor for which the analysis is performed. Selecting the Range button next to this input field will open the Name of Reactor Allowable Range Screen shown in Figure 3-10. This input field is a text field with a character limit of 120. The default value from the PWRGALE.in file is Sample PWR.

Figure 3-10 Name of Reactor Allowable Range Screen 3-10

Thermal Power Level Enter the maximum thermal power level (MWt) evaluated for safety considerations for the plant from the Design Certification Document (DCD), Final Safety Analysis Report (FSAR), or other plant design documentation. Selecting the Range button next to this input field will open the Thermal Power Level Allowable Range Screen shown in Figure 3-11. The default value from the PWRGALE.in file is 3400 MWt, and the allowable range for values in this field is greater than 0 MWt.

Figure 3-11 Thermal Power Level Allowable Range Screen Mass of Coolant in Primary System Enter the mass of coolant in the primary system at operating temperature and pressure in units of either thousands of pounds (thousand lb) or thousands of kilograms (thousand kg). Selecting the Range button next to this input field will open the Mass of Coolant in Primary System Allowable Range Screen shown in Figure 3-12. The default value from the PWRGALE.in file is 550 thousand lb, and the allowable range for values in this field is greater than either 0 thousand lb or 0 thousand kg.

Figure 3-12 Mass of Coolant in Primary System Allowable Range Screen Primary System Letdown Rate The chemical and volume control system is a major support system for the PWR reactor coolant system. The functions of the system include the following:

Purify the reactor coolant system using filters and demineralizers.

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Add and remove boron as necessary.

Maintain the level of the pressurizer at the desired set point.

A small amount of water, about 75 gallons per minute (gal/min), is continuously routed through the chemical and volume control system. Called letdown, this process provides a continuous cleanup of the reactor coolant system, which maintains the purity of the coolant and helps to minimize the amount of radioactive material in the coolant.

Enter the average letdown rate from the primary system to the purification demineralizers in units of either gal/min or liters per min (L/min). Selecting the Range button next to this input field will open the Primary Letdown Rate Allowable Range Screen shown in Figure 3-13. The default value from the PWRGALE.in file is 75 gal/min, and the allowable range for values in this field is greater than either 0 gal/min or 0 L/min.

Figure 3-13 Primary Letdown Rate Allowable Range Screen Letdown Cation Demineralizer Flow Rate Enter the annual average flow rate through the cation demineralizers for the control of cesium in the primary coolant. The average flow rate is determined by multiplying the average letdown rate (value entered for the Primary System Letdown Rate) by the fraction of time the cation demineralizers are in service to obtain the average cation demineralizer flow rate. Selecting the Range button next to this input field will open the Letdown Cation Demineralizer Flowrate Allowable Range Screen shown in Figure 3-14. The default value from the PWRGALE.in file is 7.5 gal/min, and the allowable range for values in this field is greater than either 0 gal/min or 0 L/min.

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Figure 3-14 Letdown Cation Demineralizer Flowrate Allowable Range Screen Number of Steam Generators Enter the number of steam generators for the reactor. Selecting the Range button next to this input field will open the Number of Steam Generators Allowable Range Screen shown in Figure 3-15. The default value from the PWRGALE.in file is 4, and the allowable range for values in this field is 1 or greater.

Figure 3-15 Number of Steam Generators Allowable Range Screen Total Steam Flow Enter the total steam flow for all steam generators in either units of millions of pounds per hour (million lb/h) or millions of kilograms per hour (million kg/h). Selecting the Range button next to this input field will open the Total Steam Flow Allowable Range Screen shown in Figure 3-16.

The default value from the PWRGALE.in file is 15 million lb/h, and the allowable range for values in this field is greater than either 0 million lb/h or 0 million kg/h.

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Figure 3-16 Total Steam Flow Allowable Range Screen Mass of Liquid in Each Steam Generator Enter the mass of liquid in each steam generator in either units of thousand lb or thousand kg.

Selecting the Range button next to this input field will open the Mass of Liquid in Each Steam Generator Allowable Range Screen shown in Figure 3-17. The default value from the PWRGALE.in file is 112.5 thousand lb, and the allowable range for values in this field is greater than either 0 thousand lb or 0 thousand kg.

Figure 3-17 Mass of Liquid in Each Steam Generator Allowable Range Screen Blowdown Treatment Method Select the steam generator blowdown treatment method from one of the three dropdown menu options below:

0 - blowdown is recycled to the condensate system after treatment. Note to user: If the plant has once-through steam generators, select this value. This is the default value from the PWRGALE.in file.

1 - blowdown is recycled directly to condensate system demineralizers.

2 - blowdown is not recycled to the condensate system.

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Type of Steam Generator The steam generator uses heated water in the primary coolant loop to boil water in the secondary coolant loop. PWRs use two kinds of steam generators: U-tube and once-through steam generators.

In the Westinghouse and Combustion Engineering (CE) designs, the steam/water mixture passes through multiple stages of moisture separation. This type of steam generator is a U-tube steam generator. Figure 3-18 shows sample diagrams of U-tube steam generators.

Figure 3-18 Cutaway views of Westinghouse and Combustion Engineering U-tube steam generators The Babcock & Wilcox (B&W) design PWR uses a once-through steam generator. In this design, the flow of primary coolant is from the top of the steam generator to the bottom, instead of through U-shaped tubes as in the Westinghouse and CE designs. Figure 3-19 shows a sample diagram a of once-through steam generator.

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Figure 3-19 Cutaway view of Babcock and Wilcox once-through steam generator Select either the Once-Through Steam Generator or the U-tube Steam Generator from the dropdown menu options on the General Reactors Parameter Screen (Figure 3-8). The default value from the PWRGALE.in file is U-Tube.

Steam Generator Blowdown Rate Enter the steam generator blowdown rate as given in the applicants DCD, FSAR, or environmental report. This input is entered only for U-tube steam generators. For a once-through steam generator, enter 0.0. Enter total blowdown rate in either units of thousand lb/h or thousand kg/h. Selecting the Range button next to this input field will open the U-Tube Total Blowdown Rate Allowable Range Screen shown in Figure 3-20. The default value from the PWRGALE.in file is 75 thousand lb/h, and the allowable range for values in this field is greater than either 0 thousand lb/h or 0 thousand kg/h.

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Figure 3-20 U-Tube Total Blowdown Rate Allowable Range Screen Condensate Demineralizer Regeneration Time Enter the condensate demineralizer regeneration time in units of days. Selecting the Range button next to this input field will open the Condensate Demineralizer Regeneration Time Allowable Range Screen shown in Figure 3-21. The default value from the PWRGALE.in file is 8.4 days, and the allowable range for values in this field is greater than 0 days.

Figure 3-21 Condensate Demineralizer Regeneration Time Allowable Range Screen Fraction of Feedwater through Condensate Demineralizers Enter the fraction of feedwater to the steam generator processed through the condensate demineralizers. If condensate demineralizers are not used, enter 0.0. Selecting the Range button next to this input field will open the Fraction of Feedwater through Condensate Demineralizers Allowable Range Screen shown in Figure 3-22. The default value from the PWRGALE.in file is 0.65 fraction, and the allowable range for values in this field is between 0 and 1.0.

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Figure 3-22 Fraction of Feedwater through Condensate Demineralizers Allowable Range Screen 3.2.3 Liquid GUI Inputs (Liquid Radwaste Treatment System Input Screen)

The Liquid Inputs button on the General Reactor Parameters Screen (Figure 3-8) allows the user to open the Liquid Radwaste Treatment System Screen as shown in Figure 3-23. The Liquid Radwaste Treatment System Screen contains seven tabs, as shown in Figure 3-23, for the input of the liquid radwaste inlet streams that are considered in the code. Table 3-1 list seven liquid radwaste inlet stream tabs and the potential sources for each stream.

Figure 3-23 Liquid Radwaste Treatment System Screen (Shim Bleed Tab) 3-18

Table 3-1 Seven PWR liquid radwaste inlet stream tabs and their sources Liquid Radwaste Inlet Stream Tab Potential Source of Liquid Radwaste Stream Shim Bleed Controls reactivity by bleeding out borated water Equipment Drain Waste Equipment drains from: drywell reactor building turbine building radwaste building auxiliary building fuel pool building Clean Waste Deaerated or tritiated waste Dirty Waste Aerated or nontritiated waste Blowdown Waste For U-tube steam generator, the blowdown waste Regenerant Waste Regenerant solution from ion exchange columns (condensate polishers)

Detergent Waste Sites that have onsite laundry facilities Shim Bleed Tab As described in Section 2.1, in a PWR, the shim bleed controls reactivity by bleeding out borated water. Figure 3-23 shows the Liquid Radwaste Treatment System Screen selected to the Shim Bleed Tab with the default values from the PWRGALE.in file. The inputs are described below.

Liquid Stream Field Enter the liquid stream flow rate in either units of gallons per day (gal/d) or liters per day (L/d).

Selecting the Range button next to this input field will open the Shim Bleed Liquid Stream Flow Rate Allowable Range Screen shown in Figure 3-24. The default value from the PWRGALE.in file is 1440 gal/d, and the allowable range for values in this field is greater than or equal to either 0 gal/d or 0 L/d. Table 4-20 gives sample values of typical flow rates and PCAs for various reactor drains and other liquid streams.

Figure 3-24 Shim Bleed Liquid Stream Flow Rate Allowable Range Screen 3-19

Decontamination Factor Field Enter the DFs for shim bleed for the three categories of radionuclides shown in Figure 3-23 (Iodine DF, Cs and Rb [cesium and rubidium] DF, and Other DF). Selecting the Range button next to these input fields will open the appropriate allowable range screen shown in Figure 3-25.

The default values from the PWRGALE.in file are 5e3 for Iodine DF, 2e3 for Cs and Rb DF, and 1.5e5 for Other DF. The allowable range for values in these fields is greater than or equal to 1.0. Table 3-2 provides values of typical DFs for PWR liquid waste treatment systems.

Figure 3-25 Shim Bleed Decontamination Factor Allowable Range Screens 3-20

Table 3-2 Summary of typical decontamination factors for PWR liquid waste treatment streams Demineralizersa Cesium & Other Anionb Mixed-Bed Rubidiumb Radionuclidesb Primary coolant letdown 1.0E+01 2.0E+00 5.0E+01 1.0E+02 2.0E+00 1.0E+02 Radwaste (H+ OH-)

(1.0E+01) (1.0E+01) (1.0E+01)

Evaporator condensate polishing 5.0E+00 1.0E+00 1.0E+01 Boron recycle 1.0E+01 2.0E+00 1.0E+01 1.0E+02 1.0E+01 1.0E+02 Steam generator blowdown (1.0E+01) (1.0E+01) (1.0E+01) 1.0E+00 1.0E+01 1.0E+01 Cation bed (any system)

(1.0E+00) (1.0E+01) (1.0E+01) 1.0E+02 1.0E+00 1.0E+00 Anion bed (any system)

(1.0E+01) (1.0E+00) (1.0E+00) 1.0E+01 2.0E+01 1.0E+01 Powdex (any system)

(1.0E+01) (1.0E+01) (1.0E+01)

All Radionuclides Except Evaporatorsc Iodine Iodine Miscellaneous radwaste 1.0E+03 1.0E+02 Boric acid recovery 1.0E+03 1.0E+02 Reverse Osmosisd All Radionuclides Laundry wastes 3.0E+01 Other liquid wastes 1.0E+01 Filterse DF of 1.0 for all radionuclides a These values are from Table 1-4 of NUREG-0017, Revision 1, and Table 4-27.

b For an evaporator polishing demineralizer or for the second demineralizer in series, the DF is given in parentheses.

c These values are from Table 1-4 of NUREG-0017, Revision 1, and Table 4-28.

d These values are from Table 1-4 of NUREG-0017, Revision 1, and Section 4.1.1.19.

e This value is from Table 1-4 of NUREG-0017, Revision 1, and Section 4.1.1.20.

Waste Collection and Processing Field Enter the waste collection and processing parameters for shim bleed:

Enter the waste collection time prior to processing in days. Selecting the Range button next to this input field will open the appropriate allowable range screen shown in Figure 3-26. The default value from the PWRGALE.in file is 22.6 days, and the allowable range for values in this field is greater than or equal to 0.0 days.

Enter the waste processing and discharge time in days. Selecting the Range button next to this input field will open the appropriate allowable range screen shown in Figure 3-26. The default value from the PWRGALE.in file is 0.93 days, and the allowable range for values in this field is greater than or equal to 0.0 days.

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Enter the average fraction of wastes to be discharged after processing. Selecting the Range button next to this input field will open the appropriate allowable range screen shown in Figure 3-26. The default value from the PWRGALE.in file is 0.1, and the allowable range for values in this field is between 0.0 and 1.0.

Figure 3-26 Waste Collection and Processing Allowable Range Screens In addition to inputting the specific parameters, the information for waste collection and processing can be calculated using the Calculate button (Figure 3-23). Figure 3-27 shows the calculation screen for the input of information for waste collection time and processing and discharge time. Users input the specific information for the various waste streams and then can calculate the final values used by the code.

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Figure 3-27 Waste Collection Time, and Processing and Discharge Time Calculation Screen Selecting the Range button next to each of the input fields in Figure 3-27 will open the appropriate allowable range screen shown in Figure 3-28. The appropriate units for these input fields are either units of volume (i.e., gal or L) or rates (i.e., gal/d or L/d). The allowable range for values in this field is greater than either 0 gal or 0 L (volumes) and either 0 gal/d or 0 L/d (rates). The following sections describe how these calculations are performed.

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Figure 3-28 Waste Collection Time and Processing and Discharge Time Calculation Allowable Range Screens Collection Time for Liquid Wastes Collection time before processing is based on the input flow calculated above. Where redundant tanks are provided, assume the collection tank to be filled to 80-percent design capacity. If only one tank is provided, assume the tank to be filled to 40-percent design capacity. For example, for the liquid waste processing system shown in Figure 3-29, if flow from a 1.0E+03 gal/d floor drain is collected in two 2.0E+04 gallon tanks before processing, collection time would be calculated using Equation (3-1):

(8.0E-01) (A) (8.0E-01) (2.0E+04 gal)

Tc = Ra

= gal = 16 days (3-1) 1.0E+03 d where Tc = collection time in days; 8.0E-01 = the percent design capacity for redundant tanks (4.0E-01 for single tanks);

A = the capacity of initial tank in liquid waste flow stream (gal); and Ra = flow rate into the initial collection tank (gal/d).

Additionally, the following are other variables in Figure 3-28:

B = the limiting process based on equipment flow capacity (dimensionless);

C = the capacity of final tank in flow scheme before discharge (gal);

Rb = the equipment flow capacity of the limiting process (B) (gal/d);

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Rc = the flow capacity of the final tank (C) discharge pump (gal/d); and Ro = the rate of flow of additional waste inputs to the final tank (C) (gal/d).

Then, for example, 16 days would be calculated for the Waste collection time prior to processing block (Figure 3-27), and the user should click on the Use Values button. This will transfer the value of 16 days to the waste collection time prior to processing field in Figure 3-23.

Processing and Discharge Time Figure 3-29 shows decay during processing and discharge of liquid wastes. The process time (Tp), credited for decay, is calculated using Equation (3-2) for redundant tanks and Equation (3-3) for a single tank.

(8.0E-01) (A)

Tp = Rb (3-2)

(4.0E-01) (A)

Tp = Rb (3-3)

The discharge time (Td), 50 percent credited for decay, is calculated using Equation (3-4) for redundant tanks and Equation (3-5) for a single tank.

(8.0E-01) (C)

Td = Rc (3-4)

(4.0E-01) (C)

Td = Rc (3-5)

After performing the above two calculations, use Equation (3-6) for redundant tanks and Equation (3-7) for a single tank to calculate whether credit may be taken for decay during discharge.

(8.0E-01) (C) > Tp (R b + R o ) (3-6)

(4.0E-01) (C) > Tp (R b + R o ) (3-7)

If so, then use Equation (3-8) where Decay is the new processing and discharge time to be entered for each input stream for the input variables listed above.

Decay = Tp + (5.0E-01)(Td ) (3-8)

If, however, the results of Equations 3-6 and 3-7 are Tp(Rb + Ro), Tp is used for the holdup time, Tank C may be discharged before Tank A has been completely processed. In this case, the Tp value should be entered.

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Figure 3-29 Example of liquid waste processing system flow stream scheme For example, using Equations 3-2 and 3-4 and the liquid waste flow scheme in Figure 3-29, the following process time (Tp) and discharge time (Td) can be calculated.

(8.0E-01) (2.0E+04 gal)

Tp = gal min

= 0.7 days 1.5E+01 min 1.44E+03 d (8.0E-01) (4.0E+04 gal)

Td = gal min

= 2.2 days 1.0E+01 min 1.44E+03 d Then, checking for decay credit, 8.0E-01(C)/(Rb + Ro) = 1.45 days, which is greater than Tp; therefore, credit is taken for [Tp + 5.0E-01(Td)] or 2.2 days for processing and discharge. The input for the applicable parameter is 2.2 days for processing and discharge time. To use these values in their input, the user should click on the Use Values button (Figure 3-27).

Equipment Drain Waste Tab The equipment drain wastes are identified as the wastes from equipment located in the various reactor buildings. Figure 3-30 shows the Liquid Radwaste Treatment System Screen selected to the Equipment Drain Waste Tab with the default values from the PWRGALE.in file. The inputs are described below.

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Figure 3-30 Liquid Radwaste Treatment System Screen (Equipment Drain Waste Tab)

Liquid Stream Field Enter the Liquid Stream Flow Rate in units of either gal/d or L/d. Selecting the Range button next to this input field will open the Equipment Drain Waste Stream Flow Rate Allowable Range screen shown in Figure 3-31. The default value from the PWRGALE.in file is 330 gal/d, and the allowable range for values in this field is greater than or equal to either 0 gal/d or 0 L/d.

Figure 3-31 Equipment Drain Waste Stream Flow Rate Allowable Range Screen Enter the Activity of Inlet Stream in units of fraction of PCA. Selecting the Range button next to this input field will open the Equipment Drain Waste Activity of Inlet Stream Allowable Range screen shown in Figure 3-32. The default value from the PWRGALE.in file is 0.97, and the allowable range for values in this field is between 0.0 and 1.0 fraction of PCA. Table 4-20 gives sample values of typical flow rates and PCAs for various reactor drains and other liquid streams.

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Figure 3-32 Equipment Drain Waste Activity of Inlet Stream Allowable Range Screen In addition, because there are multiple sources of liquid waste for the different waste types, the user can access a separate calculation screen by selecting the Combine from various sources button, as shown in Figure 3-30. Figure 3-33 shows the Equipment Drain Waste Screen where specific flow rates in units of either gal/d or L/d and activity fractions can be input to calculate the total liquid stream for the various waste types.

After inputting the values in the top section, the user should press the Calculate button to calculate the values in the bottom section. GALE-PWR 3.2 calculates the input PCAs based on the weighted average activity of the composite stream entering the waste collection tanks using Equation (3-9):

(R1 ) (PCA1 ) +(R2 ) (PCA2 ) + (R3 ) (PCA3 )

PCA = (R1 + R2 + R3 )

(3-9) where R1 = the flow rate of the first input stream (gal/d);

PCA1= the PCA for the first input stream (unitless);

R2 = the flow rate of the second input stream (gal/d);

PCA2= the PCA for the second input stream (unitless);

R3 = the flow rate of the third input stream (gal/d); and PCA3= the PCA for the third input stream (unitless).

For example, if the inlet stream 1, 2, and 3 flow rates and PCAs are as listed below Stream 1 1.0E+03 gal/d at 1.0E-02 PCA Stream 2 2.0E+03 gal/d at 1.0E-01 PCA Stream 3 5.0E+02 gal/d at 1.0E+00 PCA then the composite 1, 2, 3 activity would be calculated as follows:

gal (1.0E-02) gal (1.0E-01) gal (1.0E+00) 1.0E+03 d +2.0E+03 d + 5.0E+02 d PCA = gal gal gal = 2.0E-01 1.0E+03 d + 2.0E+03 d + 5.0E+02 d The calculated composite PCA of the inlet stream is 2.0E-01, and the total average flow rate is 3.5E+03 gal/d. To use these values in the input, the user should click on the Use Values button.

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Figure 3-33 Equipment Drain Waste Screen Decontamination Factors (DF) Field Enter the DFs for the equipment drain waste for the three categories of radionuclides shown in Figure 3-30 (Iodine DF, Cs and Rb DF, and Other DF). Selecting the Range button next to these input fields will open the appropriate allowable range screen shown in Figure 3-34. The default values from the PWRGALE.in file are 5e3 for Iodine DF, 2e3 for Cs and Rb DF, and 1e5 for Other DF. The allowable range for values in these fields is greater than or equal to 1.0. Table 3-2 provides values of typical DFs for PWR liquid waste treatment systems.

Waste Collection and Processing Field Enter the waste collection and processing parameters for equipment drain waste:

Enter the waste collection time prior to processing in days. Selecting the Range button next to this input field will open the appropriate allowable range screen shown in Figure 3-26. The default value, as shown in Figure 3-30, from the PWRGALE.in file is 22.6 days, and the allowable range for values in this field is greater than or equal to 0.0 days.

Enter the waste processing and discharge time in days. Selecting the Range button next to this input field will open the appropriate allowable range screen shown in 3-29

Figure 3-26. The default value, as shown in Figure 3-30, from the PWRGALE.in file is 0.93 days and the allowable range for values in this field is greater than or equal to 0.0 days.

Enter the average fraction of wastes to be discharged after processing. Selecting the Range button next to this input field will open the appropriate allowable range screen shown in Figure 3-26. The default value, as shown in Figure 3-30, from the PWRGALE.in file is 0.1, and the allowable range for values in this field is between 0.0 and 1.0.

Figure 3-34 Equipment Drain Waste Decontamination Factor Allowable Range Screens In addition to inputting the specific parameters, the information for waste collection and processing can be calculated using the Calculate button (Figure 3-30). Figure 3-27 shows the calculation screen for the input of information for waste collection time and processing and discharge time. Users input the specific information for the various waste streams and then can calculate the final values used by the code. After inputting the values in the top section, the user should press the Calculate button to calculate the values in the bottom section. To use these values in the input, the user should click on the Use Values button (Figure 3-27).

Section 3.2.3.1 explains the equations used to determine these three parameters.

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Clean Waste Tab In Section 2.2.2, clean waste is defined as waste that has been deaerated or is tritiated.

Figure 3-35 shows the Liquid Radwaste Treatment System Screen selected to the Clean Waste Tab with the default values from the PWRGALE.in file. The following discussion describes the inputs.

Figure 3-35 Liquid Radwaste Treatment System Screen (Clean Waste Tab)

Liquid Stream Field Enter the liquid stream flow rate in units of either gal/d or L/d. Selecting the Range button next to this input field will open the Clean Waste Stream Flow Rate Allowable Range Screen shown in Figure 3-36. The default value from the PWRGALE.in file is 980 gal/d, and the allowable range for values in this field is greater than or equal to either 0 gal/d or 0 L/d.

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Figure 3-36 Clean Waste Stream Flow Rate Allowable Range Screen Enter the activity of inlet stream in units of fraction of PCA. Selecting the Range button next to this input field will open the Clean Waste Activity of Inlet Stream Allowable Range screen shown in Figure 3-37. The default value from the PWRGALE.in file is 0.093, and the allowable range for values in this field is between 0.0 and 1.0 fraction of PCA. Table 4-20 gives sample values of typical flow rates and PCAs for various reactor drains and other liquid streams.

Figure 3-37 Clean Waste Activity of Inlet Stream Allowable Range Screen In addition, because there are multiple sources of liquid waste for the different waste types, the user can access a separate calculation screen by selecting the Combine from various sources button, as shown in Figure 3-35. Figure 3-38 shows the clean waste (deaerated or tritiated) where specific flow rates in units of either gal/d or L/d and activity fractions can be input to calculate the total liquid stream for the various waste types. After inputting the values in the top section, the user should press the Calculate button to calculate the values in the bottom section. To use these values in the input, the user should click on the Use Values button.

Section 3.2.3.2 explains the equation used to combine multiple liquid streams to calculate an effective flow rate and PCA.

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Figure 3-38 Clean Waste (deaerated or tritiated) Screen Decontamination Factors (DF) Field Enter the DFs for the clean waste for the three categories of radionuclides shown in Figure 3-35 (Iodine DF, Cs and Rb DF, and Other DF). Selecting the Range button next to these input fields will open the appropriate allowable range screen shown in Figure 3-39. The default values from the PWRGALE.in file are 5e2 for Iodine DF, 1e3 for Cs and Rb DF, and 1e4 for Other DF. The allowable range for values in these fields is greater than or equal to 1.0.

Table 3-2 provides values of typical DFs for PWR liquid waste treatment systems.

Waste Collection and Processing Field Enter the waste collection and processing parameters for clean waste.

Enter the waste collection time prior to processing in days. Selecting the Range button next to this input field will open the appropriate allowable range screen shown in Figure 3-26. The default value, as shown in Figure 3-35, from the PWRGALE.in file is 5.7 days, and the allowable range for values in this field is greater than or equal to 0.0 days.

Enter the waste processing and discharge time in days. Selecting the Range button next to this input field will open the appropriate allowable range screen shown in 3-33

Figure 3-26. The default value, as shown in Figure 3-35, from the PWRGALE.in file is 0.13 days, and the allowable range for values in this field is greater than or equal to 0.0 days.

Enter the average fraction of wastes to be discharged after processing. Selecting the Range button next to this input field will open the appropriate allowable range screen shown in Figure 3-26. The default value, as shown in Figure 3-35, from the PWRGALE.in file is 0.1, and the allowable range for values in this field is between 0.0 and 1.0.

Figure 3-39 Clean Waste Decontamination Factor Allowable Range Screens In addition to inputting the specific parameters, the information for waste collection and processing can be calculated using the Calculate button (Figure 3-35). Figure 3-27 shows the calculation screen for the input of information for waste collection time, processing, and discharge time. Users input the specific information for the various waste streams and then can calculate the final values used by the code. After inputting the values in the top section, the user should press the Calculate button to calculate the values in the bottom section. To use these values in the input, the user should click on the Use Values button (Figure 3-27).

Section 3.2.3.1 explains the equations used to determine these three parameters.

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Dirty Waste Tab Section 2.2.2 defines dirty waste as waste that has been aerated or nontritiated. Figure 3-40 shows the Liquid Radwaste Treatment System selected to the Dirty Waste Tab with the default values from the PWRGALE.in file. The inputs are described below.

Figure 3-40 Liquid Radwaste Treatment System Screen (Dirty Waste Tab)

Liquid Stream Field Enter the liquid stream flow rate in units of either gal/d or L/d. Selecting the Range button next to this input field will open the Dirty Waste Stream Flow Rate Allowable Range Screen shown in Figure 3-41. The default value from the PWRGALE.in file is 2100 gal/d, and the allowable range for values in this field is greater than or equal to either 0 gal/d or 0 L/d.

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Figure 3-41 Dirty Waste Stream Flow Rate Allowable Range Screen Enter the activity of inlet stream in units of fraction of PCA. Selecting the Range button next to this input field will open the Dirty Waste Activity of Inlet Stream Allowable Range Screen shown in Figure 3-42. The default value from the PWRGALE.in file is 0.01, and the allowable range for values in this field is between 0.0 and 1.0 fraction of PCA. Table 4-20 gives sample values of typical flow rates and PCAs for various reactor drains and other liquid streams.

Figure 3-42 Dirty Waste Activity of Inlet Stream Allowable Range Screen In addition, because there are multiple sources of liquid waste for the different waste types, the user can access a separate calculation screen by selecting the Combine from various sources button, as shown in Figure 3-40. Figure 3-43 shows the Miscellaneous Dirty Waste (aerated or nontritiated) Screen, where specific flow rates in units of either gal/d or L/d and activity fractions can be input to calculate the total liquid stream for the various waste types. After inputting the values in the top section, the user should press the Calculate button to calculate the values in the bottom section. To use these values in the input, the user should click on the Use Values button. Section 3.2.3.2 explains the equation used to combine multiple liquid streams to calculate an effective flow rate and PCA.

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Decontamination Factors (DF) Field Enter the DFs for the dirty waste for the three categories of radionuclides shown in Figure 3-40 (Iodine DF, Cs and Rb DF, and Other DF). Selecting the Range button next to these input fields will open the appropriate allowable range screen shown in Figure 3-44. The default values from the PWRGALE.in file are 5e2 for iodine DF, 1e3 for Cs and Rb DF, and 1e4 for other DF. The allowable range for values in these fields is greater than or equal to 1.0.

Table 3-2 provides values of typical DFs for PWR liquid waste treatment systems.

Figure 3-43 Miscellaneous Dirty Waste (aerated or non-tritiated) Screen Waste Collection and Processing Field Enter the waste collection and processing parameters for dirty waste:

Enter the waste collection time prior to processing in days. Selecting the Range button next to this input field will open the appropriate allowable range screen shown in Figure 3-26. The default value, as shown in Figure 3-40, from the PWRGALE.in file is 3.8 days, and the allowable range for values in this field is greater than or equal to 0.0 days.

Enter the waste processing and discharge time in days. Selecting the Range button next to this input field will open the appropriate allowable range screen shown in 3-37

Figure 3-26. The default value, as shown in Figure 3-40, from the PWRGALE.in file is 0.19 days, and the allowable range for values in this field is greater than or equal to 0.0 days.

Enter the average fraction of wastes to be discharged after processing. Selecting the Range button next to this input field will open the appropriate allowable range screen shown in Figure 3-26. The default value, as shown in Figure 3-40, from the PWRGALE.in file is 1.0, and the allowable range for values in this field is between 0.0 and 1.0.

Figure 3-44 Dirty Waste Decontamination Factor Allowable Range Screens In addition to inputting the specific parameters, the user can calculate the information for waste collection and processing by using the Calculate button (Figure 3-40). Figure 3-27 shows the calculation screen for the input of information for waste collection time and processing and discharge time. Users input the specific information for the various waste streams and then can calculate the final values used by the code. After inputting the values in the top section, the user should press the Calculate button to calculate the values in the bottom section. To use these values in the input, the user should click on the Use Values button (Figure 3-27).

Section 3.2.3.1 explains the equations used to determine these three parameters.

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Blowdown Waste Tab As defined in Section 2.2.2, steam generator blowdown waste is material ejected from equipment during a steam generator (U-tube only) blowdown operation. Figure 3-45 shows the Liquid Radwaste Treatment System Screen selected to the Blowdown Waste Tab with the default values from the PWRGALE.in file. The inputs are described below.

Figure 3-45 Liquid Radwaste Treatment System Screen (Blowdown Waste Tab)

Fraction of Steam Processed Field Enter the fraction of steam processed in this field. The default value from the PWRGALE.in file is 1.0, and the allowable range for values in this field is between 0.0 and 1.0. Selecting the Range button next to this input field will open the Blowdown Waste Fraction of Steam Processed Allowable Range Screen shown in Figure 3-46.

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Figure 3-46 Blowdown Waste Fraction of Steam Processed Allowable Range Screen Decontamination Factors (DF) Field Enter the DFs for the dirty waste for the three categories of radionuclides shown in Figure 3-45 (Iodine DF, Cs and Rb DF, and Other DF). Selecting the Range button next to these input fields will open the appropriate allowable range screen shown in Figure 3-47. The default values from the PWRGALE.in file are 1e3 for Iodine DF, 1e2 for Cs and Rb DF, and 1e3 for Other DF. The allowable range for values in these fields is greater than or equal to 1.0.

Table 3-2 provides values of typical DFs for PWR liquid waste treatment systems.

Waste Collection and Processing Field Enter the waste collection and processing parameters for blowdown waste:

Enter the waste collection time prior to processing in days. Selecting the Range button next to this input field will open the appropriate allowable range screen shown in Figure 3-26. The default value, as shown in Figure 3-45, from the PWRGALE.in file is 0.0 days, and the allowable range for values in this field is greater than or equal to 0.0 days.

Enter the waste processing and discharge time in days. Selecting the Range button next to this input field will open the appropriate allowable range screen shown in Figure 3-26. The default value, as shown in Figure 3-45, from the PWRGALE.in file is 0.0 days, and the allowable range for values in this field is greater than or equal to 0.0 days.

Enter the average fraction of wastes to be discharged after processing. Selecting the Range button next to this input field will open the appropriate allowable range screen shown in Figure 3-26. The default value, as shown in Figure 3-45, from the PWRGALE.in file is 0.0, and the allowable range for values in this field is between 0.0 and 1.0.

In addition to inputting the specific parameters, the user can calculate the information for waste collection and processing by using the Calculate button (Figure 3-45). Figure 3-27 shows the 3-40

calculation screen for the input of information for waste collection time and processing and discharge time. Users input the specific information for the various waste streams and then can calculate the final values used by the code. After inputting the values in the top section, the user should press the Calculate button to calculate the values in the bottom section. To use these values in the input, the user should click on the Use Values button (Figure 3-27).

Section 3.2.3.1 explains the equations used to determine these three parameters.

Figure 3-47 Blowdown Waste Decontamination Factor Allowable Range Screens Regenerant Waste Tab As defined in Section 2.2.2, regenerant waste is a liquid solution from ion exchange columns or condensate polishers that are part of the demineralizer system. Figure 3-48 shows the Liquid Radwaste Treatment System Screen selected to the Regenerant Waste Tab with the default values from the PWRGALE.in file. These inputs are described below.

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Figure 3-48 Liquid Radwaste Treatment System Screen (Regenerant Waste Tab)

Liquid Stream Field Enter the liquid stream flow rate in units of either gal/d or L/d. Selecting the Range button next to this input field will open the Regenerant Waste Stream Flow Rate Allowable Range Screen shown in Figure 3-49. The default value from the PWRGALE.in file is 3400 gal/d, and the allowable range for values in this field is greater than or equal to either 0 gal/d or 0 L/d.

Table 4-20 gives sample values of typical flow rates and PCAs for various reactor drains and other liquid streams.

Figure 3-49 Regenerant Waste Stream Flow Rate Allowable Range Screen Decontamination Factors (DF) Field Enter the DFs for the regenerant waste for the three categories of radionuclides shown in Figure 3-48 (Iodine DF, Cs and Rb DF, and Other DF). Selecting the Range button next to these input fields will open the appropriate allowable range screen shown in Figure 3-50. The 3-42

default values from the PWRGALE.in file are 5e2 for Iodine DF, 1e3 for Cs and Rb DF, and 1e4 for Other DF. The allowable range for values in these fields is greater than or equal to 1.0. Table 3-2 provides values of typical DFs for PWR liquid waste treatment systems.

Waste Collection and Processing Field Enter the waste collection and processing parameters for regenerant waste.

Enter the waste collection time prior to processing in days. Selecting the Range button next to this input field will open the appropriate allowable range screen shown in Figure 3-26. The default value, as shown in Figure 3-48, from the PWRGALE.in file is 4.7 days, and the allowable range for values in this field is greater than or equal to 0.0 days.

Enter the waste processing and discharge time in days. Selecting the Range button next to this input field will open the appropriate allowable range screen shown in Figure 3-26. The default value, as shown in Figure 3-48, from the PWRGALE.in file is 0.37 days, and the allowable range for values in this field is greater than or equal to 0.0 days.

Enter the average fraction of wastes to be discharged after processing. Selecting the Range button next to this input field will open the appropriate allowable range screen shown in Figure 3-26. The default value, as shown in Figure 3-48, from the PWRGALE.in file is 0.1, and the allowable range for values in this field is between 0.0 and 1.0.

In addition to inputting the specific parameters, the user can calculate information for waste collection and processing by using the Calculate button (Figure 3-48). Figure 3-27 shows the calculation screen for the input of information for waste collection time and processing and discharge time. Users input the specific information for the various waste streams and then can calculate the final values used by the code. After inputting the values in the top section, the user should press the Calculate button to calculate the values in the bottom section. To use these values in the input, the user should click on the Use Values button (Figure 3-27).

Section 3.2.3.1 explains the equations used to determine these three parameters.

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Figure 3-50 Regenerant Waste Decontamination Factor Allowable Range Screens Detergent Waste Tab Figure 3-51 shows the Liquid Radwaste Treatment System Screen selected to the Detergent Waste Tab.

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Figure 3-51 Liquid Radwaste Treatment System Screen (Detergent Waste Tab)

Detergent Waste Partition Factor Enter the Detergent Waste Partition Factor in this field. Selecting the Range button next to this input field will open the Detergent Waste Partition Factor Allowable Range Screen shown in Figure 3-52. The default value from the PWRGALE.in file is 1.0, and the allowable range for values in this field is either greater than or equal to 0 and less than or equal to 1.0.

Figure 3-52 Detergent Waste Partition Factor Allowable Range Screen If the plant does not have an onsite laundry, enter 0.0. If the plant has an onsite laundry and detergent wastes are released without treatment, enter 1.0. If detergent wastes are treated before discharge, enter the fraction of radionuclides remaining after treatment (1/DF). The radionuclides added to the liquid release resulting from an onsite laundry and any cleanup systems in place for the laundry are calculated according to Table 4-21 in Section 4.1.1.14.

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After entering the Liquid Radwaste Treatment System Screen parameters, select the Save button to save any changes made to the liquid radwaste input parameters. Then select the OK button to close the Liquid Radwaste Treatment System Screen (Figure 3-51) and return to the General Reactor Parameters Screen (Figure 3-8). Note that selecting the OK button without first selecting the Save button will result in the code returning to the General Reactor Parameters Screen without saving any changes made to the liquid radwaste input parameters.

Therefore, to save any changes made to the inputs, the user should select the Save button before selecting the OK button.

3.2.4 Gaseous GUI Inputs (Gaseous Radwaste Treatment System Input Screen)

The Gas Inputs button on the General Reactor Parameters Screen (Figure 3-8) allows the user to open the Gaseous Radwaste Treatment System Screen as shown in Figure 3-53. The Gaseous Radwaste Treatment System Screen includes fields for the Letdown System, Holdup time for fission gases stripped from the primary coolant, Iodine, and six building subsection tabs, as shown in Figure 3-53. The six building subsection tabs identify the parameters for the following release pathways: (1) Waste Gas System Particulate Release, (2) Fuel handling building, (3) Auxiliary Building, (4) Containment building, (5) Containment high volume purge, and (6) Containment low volume purge.

Figure 3-53 Gaseous Radwaste Treatment System Screen 3-46

Letdown System Field This field allows the user to select one of three options from the dropdown menu to define a value for the noble gas radwaste flow ratio (parameter Y). Figure 3-54 shows the Gaseous Radwaste Treatment System Screen with the dropdown menu for the Letdown System field.

These options are the following:

0 - no continuous gas stripping of the full letdown flowIn this option, there is no continuous gas stripping of the full letdown flow, and parameter Y = 0.0.

1 - continuous degasification of the full letdown flowIn this option, there is continuous degasification of the full letdown flow to the gaseous radwaste system via a gas stripper, and parameter Y = 1.0. This is the default value from the PWRGALE.in file.

2 - continuous purging of the volume control tankIn this option, there is continuous purging of the volume control tank, and parameter Y = 0.25.

Figure 3-54 Gaseous Radwaste Treatment System Screen with Letdown System field options 3-47

The code calculates the total amount of fission gases routed to the gaseous radwaste system from several systems in the plant (e.g., volume control tank, shim bleed gas stripper, equipment drain tanks, and cover gas).

Holdup Time for Fission Gases Stripped from the Primary Coolant Field As shown in Figure 3-55 in the Holdup time for fission gases stripped from the primary coolant field on the Gaseous Radwaste Treatment System Screen, enter the following parameters:

Figure 3-55 Gaseous Radwaste Treatment System Screen with holdup time for fission gases stripped from the primary coolant field

Enter the holdup time for xenon (Xe) in days. Selecting the Range button next to this input field will open the Holdup Time for Xe Allowable Range Screen shown in Figure 3-56. The default value from the PWRGALE.in file is 60 days, and the allowable range for values in this field is greater than or equal to 0.0 days.

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Figure 3-56 Holdup Time for Xe Allowable Range Screen

Enter the holdup time for krypton (Kr) in days. Selecting the Range button next to this input field will open the Holdup Time for Kr Allowable Range Screen shown in Figure 3-57. The default value from the PWRGALE.in file is 3.54 days, and the allowable range for values in this field is greater than or equal to 0.0 days.

Figure 3-57 Holdup Time for Kr Allowable Range Screen

Enter the fill time of decay tanks for the gas stripper in days. Selecting the Range button next to this input field will open the Fill Time of Decay Tanks for the Gas Stripper Allowable Range screen shown in Figure 3-58. The default value from the PWRGALE.in file is 0.0 days, and the allowable range for values in this field is greater than or equal to 0.0 days.

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Figure 3-58 Fill Time of Decay Tanks for the Gas Stripper Allowable Range Screen The holdup time for gases stripped from the primary coolant is hand calculated because of the multiplicity of holdup system designs. The calculations are based on the following holdup systems:

Pressurized Storage TanksSection 4.1.1.11 discusses holdup parameters for this gaseous radwaste treatment system. For pressurized storage tank systems, enter the computed Xe and Kr holdup time in days and enter the fill time.

Charcoal Delay SystemsSection 4.1.1.12 discusses holdup parameters for this gaseous radwaste treatment system. For charcoal delay systems, enter the computed Xe and Kr holdup time in days, and enter zero for the fill time.

Cover Gas Recycle SystemThe calculations for holdup parameters for this gaseous radwaste treatment or other systems designed to hold gases indefinitely are based on a 90-day holdup time. For cover gas recycle systems, enter the computed Xe and Kr holdup time of 90 days, and enter zero for the fill time.

Iodine Field This field allows the user to enter the fraction of iodine released from the blowdown tank vent and the fraction of radioiodine removed from air ejector release as shown in Figure 3-59. The inputs for the iodine field are described below:

Fraction of iodine released from blowdown tank ventEnter 0.0 in this field (1) if the gases from the blowdown flash tank are vented through a condenser before release, (2) if the blowdown flash tank is vented to the main condenser air ejector, and (3) for a once-through steam generator system. For older plants that still use flash tanks that vent directly to the atmosphere, an iodine partition factor of 0.05 is used. Selecting the Range button next to this input field will open the Fraction of Iodine Released from Blowdown Tank Vent Allowable Range screen shown in Figure 3-60. The default value from the PWRGALE.in file is 0.0, and the allowable range for values in this field is greater than or equal to 0.0 and less than or equal to 1.0.

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Figure 3-59 Gaseous Radwaste Treatment System Screen with Iodine field Figure 3-60 Fraction of Iodine Released from Blowdown Tank Vent Allowable Range Screen

Percent of iodine removed from air ejector releaseIn the turbine condenser that condenses the steam that passes through the turbine, the air ejector removes air from 3-51

the condenser by passing steam through a series of nozzles to create a vacuum that removes air from the condenser. Figure 3-61 shows a diagram of this system.

(a) (b)

Figure 3-61 Diagram of turbine condenser (a) and steam-jet air ejector (b)

If, before release, the offgases from the condenser air ejector are processed through charcoal adsorbers that satisfy the guidelines of RG 1.140, Revision 1, Design, Testing, and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants, issued October 1979 [Ref. 14], enter the percent removal efficiency for radioiodine corresponding to the depth of charcoal. If the offgases are released from the condenser air ejector without treatment, enter 0.0 percent. Selecting the Range button next to this input field will open the Percent of Iodine Removed from Air Ejector Release Allowable Range Screen shown in Figure 3-62. The default value from the PWRGALE.in file is 0.0 percent, and the allowable range for values in this field is between 0.0 and 100.0 percent.

Figure 3-62 Percent of Iodine Removed from Air Ejector Release Allowable Range Screen Waste Gas System Particulate Release Tab Figure 3-63 shows the six building subsections of the Gaseous Radwaste Treatment System Screen with the Waste Gas System Particulate Release Tab open. If ventilation exhaust air 3-52

from the waste gas system is treated through high-efficiency particulate air (HEPA) filters that satisfy the guidelines of RG 1.140, Revision 1, then select Yes. When Yes is selected, the code uses a removal efficiency of 99 percent for particulates. Select No if there is no treatment provided to remove particulates or if the HEPA filters do not satisfy the guidelines of RG 1.140, Revision 1. When No is selected, the code uses a removal efficiency of 0.0 percent. The default value from the PWRGALE.in file is Yes.

Figure 3-63 Gaseous Radwaste Treatment System Screen with Waste Gas System Particulate Release Tab options Fuel Handling Building Tab Section 2.2.1 describes the PWR fuel handling building, and Figure 3-64 shows the six building subsections of the Gaseous Radwaste Treatment System Screen with the Fuel Handling Building tab open. The inputs for the Fuel Handling Building Tab are described below:

Charcoal adsorbersSelect if the ventilation exhaust air from the fuel handling building is treated through charcoal adsorbers that satisfy the guidelines of NUREG-0017, Revision 1, and RG 1.140, Revision 3, Design, Testing, and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants, issued August 2016 [Ref. 15]. If Yes is selected, enter the percent removal efficiency for iodine corresponding to the depth of charcoal. The 3-53

Reg. Guide 1.140 efficiency and NUREG-0017 efficiency buttons provide the user with these reference values as shown in Figures 3-65 and 3-66. If No is selected, the code uses a removal efficiency of 0.0 percent. Selecting the Range button next to this input field will open the Fuel Handling Building Charcoal Adsorber Removal Efficiency Allowable Range Screen shown in Figure 3-67. The default value from the PWRGALE.in file is Yes, and the removal efficiency is 90 percent. The allowable range for values in the removal efficiency field is greater than or equal to 0.0 and less than or equal to 100.0 percent.

Figure 3-64 Gaseous Radwaste Treatment System Screen with Fuel Handling Building Tab options 3-54

Figure 3-65 NUREG-0017, Revision 1 Charcoal Efficiency Information Screen 3-55

Figure 3-66 Regulatory Guide 1.140, Revision 3 Charcoal Efficiency Information Screen 3-56

Figure 3-67 Fuel Handling Building Charcoal Adsorber Removal Efficiency Allowable Range Screen

HEPA filtersIf ventilation exhaust air from the fuel handling building is treated through HEPA filters that satisfy the guidelines of RG 1.140, Revision 3, then select Yes.

When Yes is selected, the code uses a removal efficiency of 99 percent for particulates. Select No if there is no treatment provided to remove particulates or if the HEPA filters do not satisfy the guidelines of RG 1.140, Revision 3. If No is selected, the code uses a removal efficiency of 0.0 percent. The default value from the PWRGALE.in file is No.

Auxiliary Building Tab Section 2.2.1 describes the PWR auxiliary or reactor building, and Figure 3-68 shows the six building subsections of the Gaseous Radwaste Treatment System Screen with the Auxiliary Building Tab open. The inputs for the Auxiliary Building Tab are described below:

Charcoal adsorbersSelect if the ventilation exhaust air from the auxiliary building is treated through charcoal adsorbers that satisfy the guidelines of NUREG-0017, Revision 1, and RG 1.140, Revision 3. If Yes is selected, enter the percent removal efficiency for iodine corresponding to the depth of charcoal. The Reg. Guide 1.140 efficiency and NUREG-0017 efficiency buttons give the user these reference values as shown in Figures 3-65 and 3-66. If No is selected, the code uses a removal efficiency of 0.0 percent. Selecting the Range button next to this input field will open the Auxiliary Handling Building Charcoal Adsorber Removal Efficiency Allowable Range screen shown in Figure 3-69. The default value from the PWRGALE.in file is Yes, and the removal efficiency is 90 percent. The allowable range for values in the removal efficiency field is between 0.0 and 100.0 percent.

HEPA filtersIf ventilation exhaust air from the auxiliary building is treated through HEPA filters that satisfy the guidelines of RG 1.140, Revision 3, then select Yes. If Yes is selected, the code uses a removal efficiency of 99 percent for particulates.

Select No if there is no treatment provided to remove particulates or if the HEPA filters do not satisfy the guidelines of RG 1.140, Revision 3. If No is selected, the code uses a removal efficiency of 0.0 percent. The default value from the PWRGALE.in file is No.

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Figure 3-68 Gaseous Radwaste Treatment System Screen with Auxiliary Building Tab options Figure 3-69 Auxiliary Building Charcoal Adsorber Removal Efficiency Allowable Range Screen 3-58

Containment Building Tab Section 2.2.1 describes the PWR containment building, and Figure 3-70 shows the six building subsections of the Gaseous Radwaste Treatment System screen with the Containment Building tab open. The inputs for the Containment Building tab are described below:

Containment free volumeEnter the free volume of the containment in units of either millions of cubic feet (million ft3) or millions of cubic meters (million m3). Selecting the Range button next to this input field will open the Containment Building Free Volume Allowable Range Screen shown in Figure 3-71. The default value from the PWRGALE.in file is 2.715 million ft3, and the allowable range for values in this field is greater than either 0 million ft3 or 0 million m3.

Figure 3-70 Gaseous Radwaste Treatment System Screen with Containment Building Tab options 3-59

Figure 3-71 Containment Building Free Volume Allowable Range Screen

Containment Internal Cleanup SystemThe input parameters include charcoal adsorbers, HEPA filters, and the flow rate through the internal containment cleanup system.

Charcoal adsorbersSelect if the ventilation exhaust air from the containment building is treated through charcoal adsorbers that satisfy the guidelines of NUREG-0017, Revision 1, and RG 1.140, Revision 3. If Yes is selected, enter the percent removal efficiency for iodine corresponding to the depth of charcoal. The Reg. Guide 1.140 efficiency and NUREG-0017 efficiency buttons give the user these reference values as shown in Figures 3-65 and 3-66. If No is selected, the code uses a removal efficiency of 0.0 percent. Selecting the Range button next to this input field will open the Containment Building Charcoal Adsorber Removal Efficiency Allowable Range screen shown in Figure 3-72. The default value from the PWRGALE.in file is No, and the removal efficiency is 0 percent. The allowable range for values in the removal efficiency field is between 0.0 and 100.0 percent.

Figure 3-72 Containment Building Charcoal Adsorber Removal Efficiency Allowable Range Screen 3-60

HEPA filtersIf ventilation exhaust air from the containment building is treated through HEPA filters that satisfy the guidelines of RG 1.140, Revision 3, then select Yes. If Yes is selected, the code uses a removal efficiency of 99 percent for particulates.

Select No if there is no treatment provided to remove particulates or if the HEPA filters do not satisfy the guidelines of RG 1.140, Revision 3. If No is selected, the code uses a removal efficiency of 0.0 percent. The default value from the PWRGALE.in file is No.

Flow rate through the internal cleanup systemEnter the volumetric flow rate through the internal cleanup system in units of either thousands of cubic feet per minute (thousand ft3/min) or thousands of cubic meters per minute (thousand m3/min).

Selecting the Range button next to this input field will open the Containment Building Internal Cleanup System Flow Rate Allowable Range Screen shown in Figure 3-73. The default value from the PWRGALE.in file is 0.0 thousand ft3/min, and the allowable range for values in this field is greater than either 0 thousand ft3/min or 0 thousand m3/min.

Figure 3-73 Containment Building Internal Cleanup System Flow Rate Allowable Range Screen Containment High Volume Purge Tab Figure 3-74 shows the six buildings subsection of the Gaseous Radwaste Treatment System Screen with the Containment High Volume Purge Tab open. The inputs for the Containment High Volume Purge Tab are described below:

Charcoal adsorbersSelect if the ventilation exhaust air from the auxiliary building is treated through charcoal adsorbers that satisfy the guidelines of NUREG-0017, Revision 1, and RG 1.140, Revision 3. If Yes is selected, enter the percent removal efficiency for iodine corresponding to the depth of charcoal. The Reg. Guide 1.140 efficiency and NUREG-0017 efficiency buttons give the user these reference values as shown in Figures 3-65 and 3-66. If No is selected, the code uses a removal efficiency of 0.0 percent. Selecting the Range button next to this input field will open the Containment High Volume Purge Charcoal Adsorber Removal Efficiency Allowable Range Screen shown in Figure 3-75. The default value from the PWRGALE.in file is Yes, and the removal efficiency is 90 percent. The allowable range for values in the removal efficiency field is between 0.0 and 100.0 percent.

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Figure 3-74 Gaseous Radwaste Treatment System Screen with Containment High Volume Purge Tab options Figure 3-75 Containment High Volume Purge Charcoal Adsorber Removal Efficiency Allowable Range Screen

HEPA filtersIf ventilation exhaust air from the auxiliary building is treated through HEPA filters that satisfy the guidelines of RG 1.140, Revision 3, select Yes. When Yes is selected, the code uses a removal efficiency of 99 percent for particulates.

Select No if there is no treatment provided to remove particulates or if the HEPA filters 3-62

do not satisfy the guidelines of RG 1.140, Revision 3. If No is selected, the code uses a removal efficiency of 0.0 percent. The default value from the PWRGALE.in file is Yes.

Number of purges per year during power operationsEnter the number of purges per year, but do not include the two purges at shutdown. Selecting the Range button next to this input field will open the Containment High Volume Purge Number of Purges Allowable Range Screen shown in Figure 3-76. The default value from the PWRGALE.in file is 0, and the allowable range for values in this field is greater than or equal to 0.

Figure 3-76 Containment High Volume Purge Number of Purges Allowable Range Screen Containment Low Volume Purge Tab Figure 3-77 shows the six buildings subsection of the Gaseous Radwaste Treatment System screen with the Containment Low Volume Purge Tab open. The inputs for the Containment Low Volume Purge Tab are described below:

Charcoal adsorbersSelect if the ventilation exhaust air from the auxiliary building is treated through charcoal adsorbers that satisfy the guidelines of NUREG-0017, Revision 1, and RG 1.140, Revision 3. If Yes is selected, enter the percent removal efficiency for iodine corresponding to the depth of charcoal. The Reg. Guide 1.140 efficiency and NUREG-0017 efficiency buttons give the user these reference values as shown in Figures 3-65 and 3-66. If No is selected, the code uses a removal efficiency of 0.0 percent. Selecting the Range button next to this input field will open the Containment Low Volume Purge Charcoal Adsorber Removal Efficiency Allowable Range Screen shown in Figure 3-78. The default value from the PWRGALE.in file is Yes, and the removal efficiency is 90 percent. The allowable range for values in the removal efficiency field is between 0.0 and 100.0 percent.

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Figure 3-77 Gaseous Radwaste Treatment System Screen with Containment Low Volume Purge Tab options Figure 3-78 Containment Low Volume Purge Charcoal Adsorber Removal Efficiency Allowable Range Screen

HEPA filtersIf ventilation exhaust air from the auxiliary building is treated through HEPA filters that satisfy the guidelines of RG 1.140, Revision 3, then select Yes. If Yes is selected, the code uses a removal efficiency of 99 percent for particulates.

Select No if there is no treatment provided to remove particulates or if the HEPA filters 3-64

do not satisfy the guidelines of RG 1.140, Revision 3. If No is selected, the code uses a removal efficiency of 0.0 percent. The default value from the PWRGALE.in file is Yes.

Continuous Containment Purge RateEnter the volumetric flow rate of the continuous purge rate in units of either ft3/min or m3/min. Selecting the Range button next to this input field will open the Containment Low Volume Purge Continuous Purge Rate Allowable Range Screen shown in Figure 3-79. The default value from the PWRGALE.in file is 1000 ft3/min, and the allowable range for values in this field is greater than or equal to either 0 ft3/min or 0 m3/min.

Figure 3-79 Containment Low Volume Purge Continuous Purge Rate Allowable Range Screen After entering the Gaseous Radwaste Treatment System Screen parameters, select the Save button to save any changes made to the gaseous radwaste input parameters. Then select the OK button to close the Liquid Radwaste Treatment System Screen (Figure 3-77) and return to the General Reactor Parameters Screen (Figure 3-8). Note that selecting the OK button without first selecting the Save button will result in the code returning to the General Reactor Parameters Screen without saving any changes made to the gaseous radwaste input parameters. Therefore, to save any changes made to the inputs, the user should select the Save button before selecting the OK button.

Upon returning to the General Reactor Parameters Screen (Figure 3-8), the user can either chose to save the inputs or execute the code by selecting either the Save or Run button, respectively. If the user chose the Save button after completing the liquid and gaseous radwaste inputs, there is no need to select the Save button on the General Reactor Parameters Screen (Figure 3-8), as all input changes previously made have been saved. If user did not choose the Save button, then the user should select the Save button on the General Reactor Parameters Screen (Figure 3-8). To run the code, the user should press the Run button in the General Reactor Parameters Screen (Figure 3-8).

When the code has completed the calculations, the GALE-PWR 3.2 dialog box opens to inform the user that the run is complete and that results are available, as shown in Figure 3-80. When the user presses the Exit button, the code will open Windows Explorer in the directory containing the input and output files, as shown in Figure 3-81. Note that the code will save the PWRGALE.in and the two output files to the working directory set by the user in Section 3.1.

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The two output files generated by the GALE-PWR 3.2 code are the PWRGE.out for the gaseous effluent output and the PWRLE.out for the liquid effluent output file.

Figure 3-80 GALE code completion dialog box Figure 3-81 Windows Explorer directory with GALE-PWR 3.2 input and output files 3-66

3.3 Viewing the Code Outputs The codes two output files, PWRGE.out and PWRLE.out, are ASCII text files and may be viewed using any text editor such as Notepad, as shown in Figures 3-82 and 3-83. Both output files display the fixed model parameters that the user selected (GALE version) and the reactor coolant source term (ANS-18.1 Version) at the very top of the output files. Under these data, the output files also list any modifications that the user requested to the GALE fixed parameter file followed by the echo if the input parameters from the General Reactor Parameter, Liquid Radwaste Treatment System Input, and Gaseous Radwaste Treatment System Input Screens.

Lastly, the output files display the liquid and gaseous effluent releases.

Alternatively, the user may choose an Excel file to view the code outputs. To use this file, open the Excel file, PWR GALE Output 3.2.xls, in the working directory shown in Figure 3-81.

Figure 3-84 displays the PWR GALE Output 3.2.xls file and the first tab, specifying the output file names along with the respective locations of these files in the working directory. The user also has the option to browse for these files by selecting the Browse button. Selection of the Read GE Data and Read LE Data buttons will cause the Excel file to read the PWRGE and PWRLE output files listed in the cells above them to the spreadsheet. Once this is done, the PWRGE and PWRLE outputs and some plots are available on the other tabs of the PWR GALE Output 3.2.xls file for easier visualization or to copy into other programs or spreadsheets.

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Figure 3-82 PWRGE.out file 3-68

Figure 3-83 PWRLE.out file 3-69

Figure 3-84 GALE PWR Output 3.2.xls file 3-70

4.0 MODELING PARAMETERS This section discusses the GALE version option and the reactor coolant source term (ANS-18.1 Version) options available in GALE-PWR 3.2. In the code, the options for the ANS-18.1 version and GALE version are defaulted to ANSI/ANS-18.1-1999 source term and GALE86 reactor design parameters (NUREG-0017, Revision 1), respectively. These GALE-PWR 3.2 default values are consistent with the guidance in references DC/COL-ISG-05; RG 1.112, Revision 1; and NUREG-0800. Note that when the source term or reactor design parameters differ from those given in the GALE86 code, DC/COL-ISG-05, and RG 1.112, Revision 1, they should be described with sufficient detail, and the basis of the alternate method, model, parameters, and assumptions should be provided, to allow the NRC to conduct an independent evaluation.

Section 4.1 discusses the GALE version option, the default GALE86 parameters, which are shown in Figure 3-3. Section 4.2 discusses the reactor coolant source term, ANS-18.1 version options, and radionuclide concentrations in the primary and secondary coolant options available to the user (either the default 1999, 1984, or 2016 option, as shown in Figure 3-4). Finally, Section 4.3 discusses the user option to modify certain GALE-PWR fixed modeling parameters by means of the PWRfixed-parameters.txt file.

4.1 GALE Version (Fixed Model Parameters) 4.1.1 GALE86 Fixed Modeling Parameters Thermal Power Level Parameter This is the maximum thermal power level (in MWt) evaluated for safety considerations in the DCD or FSAR.

Bases The power level used for the source term in the code is the maximum power level evaluated for safety considerations in the DCD or FSAR. Using this value, the evaluation of the radwaste management systems need not be repeated when the applicant applies for a stretch power license at a later date. Experience shows that most utilities request approval to operate at maximum power soon after reaching commercial operation.

Plant Capacity Factor Parameter A plant capacity factor of 80 percent is used (i.e., 292 effective full-power days).

Bases The source term calculations are based on a plant capacity factor of 80 percent averaged over the 30-year operating life of the plant (i.e., the plant operates at 100-percent power 80 percent of the time). Table 4-1 lists the plant capacity factors experienced at PWRs for the period 1972 through 1977.

The average plant capacity factors shown in Table 4-1 indicate that the 80-percent factor assumed is higher than the average factors experienced. However, it is expected that the major 4-1

maintenance problems and extended refueling outages that have contributed to the lower plant capacity factors will be overcome and that the plants will achieve the 80-percent capacity factor when averaged over 30 years of operation.

Table 4-1 Plant capacity factors at operating PWRs Date of Percent Capacity Factora Facilityb Commercial Operationc 1972 1973 1974 1975 1976 1977 Connecticut Yankee 1/68 86 48d 89 84 81 82 San Onofre 1 1/68 72 60 83 85 66 62 R.E. Ginna 7/70 58 81 50d 73 52 83 Point Beach 1 12/70 69 67 76 70 78 85 H.B. Robinson 2 3/71 78 65 81 71 82 74 Palisades 12/71 61 40e d 46e 50d 78 Point Beach 2 10/72 72 77 88 86 82 Turkey Point 3 12/72 55 61 76 75 78 Surry 1 12/72 51 50d 60 67 78 Maine Yankee 12/72 17f 54 69 91 77 Surry 2 5/73 40d 76 51e 65 Oconee 1 7/73 54 71 54 54 Indian Point 2 8/73 51 68 31 73 Turkey Point 4 9/73 71 68 64 62 Fort Calhoun 9/73 61 54 57 76 Prairie Island 1 12/73 36e 83 73 83 Zion 1 12/73 49g 68 55 58 Kewaunee 6/74 75 75 77 Three Mile Island 1 9/74 79 63 79 Oconee 2 9/74 68 58 53d Arkansas 1 12/74 69 54 73 Prairie Island 2 12/74 73 69 87 Rancho Seco 4/75 28g 75 Calvert Cliffs 1 5/75 88 65 Cook 1 8/75 75 54f Millstone 2 12/75 68 63 Trojan 5/76 71 Indian Point 3 8/76 72 Beaver Valley 1 10/76 44e St. Lucie 1 12/76 78 Average 71 64 69 72 69 74 a These values are from monthly Operating Units Status Reports and Table 2-1 of NUREG-0017, Revision 1.

b Indian Point 1 and Yankee Rowe are not included because they are small reactors (<7.0E+02 MWt).

c Plant capacity factors listed are for the first full year of commercial operation. Therefore, this list does not include the following plants, which began commercial operation in 1977 and 1978: Calvert Cliffs 2, Cook 2, Crystal River 3, Davis-Besse 1, Farley 1, Salem 1, North Anna 1, and Three Mile Island 2.

d Not included because of extended outage for refueling/maintenance.

e Not included because of extended maintenance/repair to the secondary system.

f Not included because of extended operation at reduced power.

g Not included because of extended maintenance outage to repair generator.

4-2

Iodine Releases from Building Ventilation Systems Parameter The iodine releases from building ventilation systems before treatment are calculated by the code using the data in Tables 4-2 through 4-6 and 4-42 through 4-43.

Bases The iodine-131 releases from building ventilation systems are based on measurements made at some operating reactors during routine plant operation and plant shutdowns. The Electric Power Research Institute (EPRI) has identified sources of radioiodine at three operating PWRs:

R.E. Ginna NPP; Calvert Cliffs NPP, Unit 1; and Three Mile Island Nuclear Station, Unit 1 (EPRI NP-939, Sources of Radioiodine at Pressurized Water Reactors, issued November 1978 [Ref.

16]). The NRC contracted with Idaho National Engineering Laboratory to obtain operational data for liquid and gaseous waste treatment systems for the following PWRs:

Fort Calhoun Station (NUREG/CR-0140, In-Plant Source Term Measurements at Fort Calhoun StationUnit 1, issued July 1978 [Ref. 17])

Zion, Units 1 and 2 (NUREG/CR-0715, In-Plant Source Term Measurements at Zion Station, issued February 1979 [Ref. 18])

Turkey Point Nuclear Generating, Units 3 and 4 (NUREG/CR-1629, In-Plant Source Term Measurements at Turkey Point StationUnits 3 and 4, issued September 1980

[Ref. 19])

Prairie Island Nuclear Generating Plant, Units 1 and 2 (NUREG/CR-4397, In-Plant Source Term Measurements at Prairie Island Nuclear Generating Station, issued September 1985 [Ref. 20])

Rancho Seco Nuclear Generating Station (NUREG/CR-2348, In-Plant Source Term Measurements at Rancho Seco Station, issued October 1981 [Ref. 21])

4-3

Table 4-2 Radioiodine releases from building ventilation systems before treatment Containment Auxiliary Turbine Annual Normalized Radioiodine Building Buildingb Buildingc Release Ratea (Ci/yr/Ci/g) (Ci/yr/Ci/g) (Ci/yr/Ci/g)

Power Operation 8.0E-04d 7.2E-01e 3.8E+03 Refueling/Maintenance Outages 3.2E-01b 2.59E+00 4.2E+02 a The values in this table come from Table 1-1 of NUREG-0017, Revision 1. The normalized release rate during different modes of operation represents the effective leak rate for radioiodine. It is the combination of the reactor water leakage rate into the building and the partitioning of the radioiodine between the water phase in the leakage and the gas phase where it is measured. For the turbine building, the effective leak rate must consider the carryover for radioiodine from water to steam in the steam generator.

b To obtain the actual radioiodine release from these buildings in Ci/yr, multiply the normalized release by the radioiodine coolant concentration in Ci/g.

c To obtain the actual radioiodine release from the turbine building in Ci/yr, multiply the normalized release by the secondary coolant concentration in Ci/g and by the PC (NS) from Table 4-41.

d This release rate is expressed in percent per day of leakage of primary coolant inventory of radioiodine and represents the effective leak rate for radioiodine. It is the combination of the reactor water leakage rate into the buildings and the partitioning of the radioiodine between the water phase in the leakage and the gas phase where it is measured. To obtain the releases in Ci/yr during power operations from the containment building of a particular PWR, the normalized leak rates in Table 4-1 are multiplied in the GALE-PWR 3.2 PWR code by the radioiodine concentration in the reactor coolant for that particular PWR, and then this leak rate is considered along with the containment purging method for that particular PWR.

e Includes contribution from the fuel pool area.

These measurements indicate that iodine-131 building vent releases are directly related to the reactor coolant iodine-131 concentration. As a result, the releases of iodine are expressed as normalized releases; that is, the absolute measured release rate in Ci/yr is divided by the reactor coolant concentration in µCi/g to give a normalized release rate of iodine-131 in Ci/yr/µCi/g as shown in Equation (4-1):

A RN = (4-1)

RW where RN = the normalized release rate of iodine-131 (Ci/yr/µCi/g);

RA = absolute (measured) iodine-131 release rate (Ci/yr); and CRW = the measured reactor water iodine-131 concentration (µCi/g).

4-4

Table 4-3 Annual radioiodine normalized releases from containment ventilation systems Normal Operation Leak Ratea,b Data Source Normalized Release/Unit (E-03 percent/d)c Fort Calhoun 1.4E-03 Three Mile Island 1 2.5E+00 Turkey Point 3 & 4 9.0E-01 Maine Yankee 1.0E-01 R.E. Ginna 6.4E-02 Yankee Rowe 1.0E+00 Prairie Island 1 & 2 5.0E-03 Rancho Seco 2.56E+00 Average 8.0E-01 Release for Extended Outagesa,d Data Source Normalized Release/Unit (Ci/yr/µCi/g)c Three Mile Island 1 4.4E+01 Calvert Cliffs 1 1.9E-01 Average 3.2E-01 a These values are from Table 2-13 of NUREG-0017, Revision 1.

b The normalized release rate, expressed in percent/day of leakage of primary coolant inventory or radioiodine, represents the effective leak rate for radioiodine. It is the combination of the reactor water leakage rate into the buildings, and the partitioning of the radioiodine between the water phase in the leakage and the gas phase where it is measured.

c These results were obtained using iodine-131 data. The normalized release rates apply to both iodine-131 and iodine-133.

d The normalized release rate, expressed in Ci/yr/µCi/g, represents the effective leak rate for radioiodine. It is the combination of the reactor water radioiodine leakage rate into the buildings and the partitioning of the radioiodine between the water phase in the leakage and the gas phase where it is measured.

4-5

Table 4-4 Annual radioiodine normalized releases from auxiliary building ventilation systems Normal Operationa,b Data Source Normalized Release/Unit (Ci/yr/µCi/g)c Zion 1 & 2 1.0E+00 Fort Calhoun 1.2E-01 R.E. Ginna 3.2E-02 Calvert Cliffs 1 5.7E-01 Three Mile Island 1 3.4E-02 Turkey Point 3 & 4 1.85E+00 Prairie Island 1 & 2 1.3E-02 Rancho Seco 9.7E-01 Average 6.8E-01 Shutdowna,b Data Source Normalized Release/Unit (Ci/yr/µCi/g)c R.E. Ginna 8.0E-02 Calvert Cliffs 1 1.6E-02 Three Mile Island 1 1.4E-01 Turkey Point 3 & 4 6.8E+00 Rancho Seco 1.14E+00 Average 2.5E+00 a These values are from Table 2-14 of NUREG-0017, Revision 1.

b The normalized release rate, expressed in Ci/yr/µCi/g during different modes of operation, represents the effective leak rate for radioiodine. It is the combination of the reactor water radioiodine leakage rate into the buildings and the partitioning of the radioiodine between the water phase in the leakage and the gas phase where it is measured.

c These results were obtained using iodine-131 data. The normalized release rates apply to both iodine-131 and iodine-133.

4-6

Table 4-5 Annual radioiodine normalized releases from refueling area ventilation systems Normal Operationa,b Data Source Normalized Release/Unit (Ci/yr/µCi/g)c R.E. Ginna 8.0E-02 Calvert Cliffs 1 4.9E-02 Three Mile Island 1 1.2E-03 Turkey Point 3 & 4 1.6E-01 Prairie Island 1 & 2 1.9E-02 Rancho Seco 1.0E-02 Average 3.8E-02 Shutdowna,b Data Source Normalized Release/Unit (Ci/yr/µCi/g)c R.E. Ginna 1.4E-02 Calvert Cliffs 1 2.9E-02 Three Mile Island 1 6.0E-02 Turkey Point 3 & 4 5.0E-02 Rancho Seco 3.0E-01 Average 9.3E-02 a These values are from Table 2-15 of NUREG-0017, Revision 1.

b The normalized release rate, expressed in Ci/yr/µCi/g during different modes of operation, represents the effective leak rate for radioiodine. It is the combination of the reactor water radioiodine leakage rate into the building and the partitioning of the radioiodine between the water phase in the leakage and the gas phase where it is measured.

c These results were obtained using iodine-131 data. The normalized release rates apply to both iodine-131 and iodine-133.

4-7

Table 4-6 Annual radioiodine normalized releases from turbine building ventilation systems Normal Operationa,b Data Source Normalized Release/Unit (Ci/yr/µCi/g)c Monticello 3.1E+03 Oyster Creek 6.0E+03 Vermont Yankee 3.5E+02 Pilgrim 8.5E+03 Browns Ferry 1.3E+03 References 3 and 5 of NUREG-0016, Revision 1 [Ref. 22] 3.3E+03 Average 3.8E+03 Extended Shutdowna,b Data Source Normalized Release/Unit (Ci/yr/µCi/g)c Monticello 1.7E+02 Oyster Creek 3.5E+02 Vermont Yankee 6.3E+01 Browns Ferry 1.3E+02 References 3 and 5 of NUREG-0016, Revision 1 1.4E+03 Average 4.2E+02 a The data in this table are taken from Table 2-16 of NUREG-0017, Revision 1.

b The normalized release rate, expressed in Ci/yr/µCi/g during different modes of operation, represents the effective leak rate for radioiodine. It is a function of radioiodine leak rate via steam and the PC for radioiodine from reactor water to steam in the reactor vessel.

c These results were obtained using iodine-131 data. The normalized release rates apply to both iodine-131 and iodine-133.

The normalized reactor water release rate, expressed in Ci/yr/Ci/g, represents an effective leak rate for reactor water containing radioiodine. It is the combination of the water leakage rate into the building and the effect of radioiodine partitioning between the water phase in the system leakage and the vapor phase in the building atmosphere.

For the turbine building, the secondary coolant radioiodine releases are directly related to the secondary coolant iodine-131 concentration. Therefore, for the turbine building, the normalized radioiodine release, RN, is determined using Equation (4-2):

RA RN = C (4-2)

RW x PC where RN = normalized release rate of secondary coolant water containing iodine-131 (Ci/yr/Ci/g);

4-8

RA = absolute (measured) iodine-131 release rate (Ci/yr);

CRW = measured secondary coolant iodine-131 concentration (Ci/g); and PC = measured radioiodine PC from secondary coolant water to steam.

The normalized release rate is used to estimate the release from PWRs since this expression for release rate is least variable with time for a given mode of operation. For this reason, it is useful in the determination of releases from PWRs.

Tables 4-2 through 4-5 present data on normalized release rates from the three reactors used in the EPRI study (Reference 16) and the five reactors used in the NRC-sponsored study (References 17 through 21) for normal operation and shutdown periods for the containment building, auxiliary building, and refueling area, respectively. Table 4-2 also gives the normalized value of the radioiodine release data discussed in NUREG-0017, Revision 1. For Table 4-6, it was considered that since the basic design and operation of PWR and boiling-water reactor (BWR) power generation equipment housed in the turbine building are essentially identical, the turbine building leakage rates from PWRs and BWRs should be similar. Therefore, for the PWR turbine building normalized radioiodine release rate, Table 4-6 of this report reproduces the values for BWRs given in Table 2-16 of NUREG-0016, Revision 1.

The data in Tables 4-2 through 4-5 are expressed as total normalized releases during power operation of 300 days and the total normalized releases during shutdowns of 65 days. Because the reactors used in the EPRI study and the NRC study experienced several intermittent brief shutdowns during the power operation measurement period, the radioiodine releases during these short-duration outages are included under power operation.

Since the releases from the containment building depend on the method of containment purging (see Section 4.1.1.8), the releases in Table 4-3 are expressed in terms of a leak rate (in E-03 percent/day of primary coolant inventory). In addition, the release from the containment building during extended outages is expressed as a total normalized release as discussed above for other buildings.

To obtain the releases in Ci/yr from the auxiliary building and the refueling area of a particular PWR, the normalized release data in Tables 4-4 and 4-5, respectively, are multiplied in the code by the radioiodine concentrations in the reactor coolant for that particular PWR using Equation (4-3):

RPWRi = RN x CPWRi (4-3) where RPWRi = calculated annual release rate for particular PWR for radioiodine radionuclide i (Ci/yr);

RN = normalized annual release rate of radioiodine from Tables 4-4 and 4-5 (Ci/yr/Ci/g); and CPRWi = calculated reactor water concentration for particular PWR for radioiodine isotope i (Ci/g).

To obtain the release in Ci/yr from the turbine building of a particular PWR, the code multiplies the normalized release data in Table 4-6 by the radioiodine concentration in the secondary coolant water and the radioiodine PC from the water to steam in the steam generator for that particular PWR using Equation (4-4):

RPWRi = RN x SCPWRi x PCPWR (4-4) 4-9

where RPWRi = calculated annual release rate for particular PWR for radioiodine isotope i (Ci/yr);

RN = normalized annual release rate of radioiodine from Table 4-6 (Ci/yr/Ci/g);

SCPRWi = calculated secondary coolant concentration for particular PWR for radioiodine isotope i (Ci/g); and PCPWR = PC from the secondary coolant water to steam for the particular PWR (see Table 4-41).

To obtain the releases in Ci/yr from the containment building of a particular PWR, the code multiplies the normalized leak rates in Table 4-2 by the radioiodine concentration in the reactor coolant for that particular PWR. This leak rate is then considered along with the containment purging method for that particular PWR.

To obtain the releases during shutdown, the code multiplies the normalized release rates for the shutdown period by the same reactor coolant concentration as for power operations. Use of this reactor coolant concentration is acceptable since the normalization technique based the shutdown normalized release rate on the reactor coolant concentrations before shutdown.

Iodine released from PWR building ventilation systems appears in one of the following chemical forms: particulate, elemental, hypoiodous acid (HOI), and organic. Based on data in NUREG/CR-4397 and NUREG/CR-2348, Table 4-7 gives the fraction of the radioiodine appearing in each of the chemical forms for each building ventilation system.

Table 4-7 Fraction of radioiodine appearing in each chemical form from PWR building ventilation systems Iodine Typea Containment Auxiliary Turbine Fuel Handling Particulate 9.0E-02 4.0E-02 b 1.0E-02 Elemental 2.1E-01 2.1E-01 7.8E-01 1.7E-01 Hypoiodous acid (HOI) 2.1E-01 2.2E-01 b 5.7E-01 Organic 4.9E-01 5.3E-01 b 2.5E-01 a The data in this table are taken from Section 2.2.4.2 of NUREG-0017, Revision 1.

b There are no data on breakdown of other species.

4-10

Radioactive Particulates Released in Gaseous Effluents Parameter Use the radioactive particulate release rates for gaseous effluents given in Table 4-8.

Table 4-8 Particulate release rate for gaseous effluents Containment Auxiliary Building Fuel Pool Area Waste Gas System Nuclidea (Ci/yr)/unit (Ci/yr)/unit (Ci/yr)/unit (Ci/yr)/unit Cr-51 9.2E-03 3.2E-04 1.8E-04 1.4E-05 Mn-54 5.3E-03 7.8E-05 3.0E-04 2.1E-06 Co-57 8.2E-04 NA NA NA Co-58 2.5E-02 1.9E-03 2.1E-02 8.7E-06 Co-60 2.6E-03 5.1E-04 8.2E-03 1.4E-05 Fe-59 2.7E-03 5.0E-05 NA 1.8E-06 Sr-89b 1.3E-02 7.5E-04 2.1E-03 4.4E-05 Sr-90b 5.2E-03 2.9E-04 8.0E-04 1.7E-05 Zr-95 NA 1.0E-03 3.6E-06 4.8E-06 Nb-95 1.8E-03 3.0E-05 2.4E-03 3.7E-06 Ru-103 1.6E-03 2.3E-05 3.8E-05 3.2E-06 Ru-106 NA 6.0E-06 6.9E-05 2.7E-06 Sb-125 NA 3.9E-06 5.7E-05 NA Cs-134 2.5E-03 5.4E-04 1.7E-03 3.3E-05 Cs-136 3.2E-03 4.8E-05 NA 5.3E-06 Cs-137 5.5E-03 7.2E-04 2.7E-03 7.7E-05 Ba-140 NA 4.0E-04 NA 2.3E-05 Ce-141 1.3E-03 2.6E-05 4.4E-07 2.2E-06 a The data in this table are taken from Table 2-17 of NUREG-0017, Revision 1. Particulate release rates are before filtration.

b Data were not available from (References 16-19); therefore, strontium-89 and strontium-90 data were extracted from semiannual effluent release reports. Release from each area above was calculated by use of percent released from each area (References 16-19).

NA = No release observed from this source. Release assumed to be less than 1 percent of total.

4-11

Basis Tables 4-9 through 4-12 list measured particulate releases at 12 operating reactors (References 17 through 21). The average annual release rates for each radionuclide released from four sources within the plant have been calculated based on the data in Tables 4-9 through 4-12.

The measurements shown in Tables 4-9 through 4-12 were taken upstream of HEPA filters on streams on which HEPA filters are located. Based on these measurements, 63 percent of the releases came from the containment, 5 percent from the auxiliary building, 31 percent from the fuel pool area, and less than 1 percent from the waste gas processing system.

Table 4-9 Measured releases upstream of HEPA filterscontainment Three Turkey Prairie Fort Zion Calvert R.E. Rancho Mile Point Island Average Calhounb 1 & 2c Cliffs 1a Ginnaa Secof Nuclide Island 1a 3 & 4d 1 & 2e Ciyr Ciyr Ciyr Ciyr Ciyr Ciyr Ciyr Ciyr Ciyr unit unit unit unit unit unit unit unit unit Cr-51 5.5E-02 ND ND ND NA NA NA NA 9.2E-03 Mn-54 2.1E-02 1.4E-08 3.9E-06 NA NA NA NA NA 5.3E-03 Co-57 4.9E-03 ND ND ND NA NA NA NA 8.2E-04 Co-58 2.2E-01 5.6E-08 1.5E-05 3.2E-06 NA NA 6.6E-08 2.5E-03 2.5E-02 Co-60 2.3E-02 3.8E-08 1.2E-05 3.0E-05 NA NA 1.4E-07 3.3E-04 2.6E-03 Fe-59 1.6E-02 ND ND ND NA NA NA NA 2.7E-03 Zr-95 NA NA NA NA NA NA NA NA NA Nb-95 1.1E-02 ND ND ND NA NA NA NA 1.8E-03 Ru-103 9.5E-03 ND ND ND NA NA NA NA 1.6E-03 Ru-106 NA NA NA NA NA NA NA NA NA Sb-125 NA NA NA NA NA NA NA NA NA Cs-134 2.1E-02 3.2E-06 2.3E-04 7.7E-05 NA NA 3.2E-08 1.5E-03 2.5E-03 Cs-136 1.9E-02 ND ND ND NA NA NA NA 3.2E-03 Cs-137 4.4E-02 4.1E-06 3.2E-04 1.9E-04 NA NA 6.6E-08 5.0E-03 5.5E-03 Ba-140 NA NA NA NA NA NA NA NA NA Ce-141 8.0E-03 ND ND ND NA NA NA NA 1.3E-03 a

The data in this column are taken from Table 2-18 of NUREG-0017, Revision 1, and EPRI NP-939.

b The data in this column are taken from Table 2-18 of NUREG-0017, Revision 1, and NUREG/CR-0140.

c The data in this column are taken from Table 2-18 of NUREG-0017, Revision 1, and NUREG/CR-0715.

d The data in this column are taken from Table 2-18 of NUREG-0017, Revision 1, and NUREG/CR-1629.

e The data in this column are taken from Table 2-18 of NUREG-0017, Revision 1, and NUREG/CR-4397.

f The data in this column are taken from Table 2-18 of NUREG-0017, Revision 1, and NUREG/CR-2348.

ND = Not detected. For averaging purposes, a value of zero was assumed.

NA = Not analyzed (or no measurement taken); plants not included in averaging.

4-12

Table 4-10 Measured releases upstream of HEPA filtersauxiliary building Three Turkey Prairie Fort Zion Calvert R.E. Rancho Mile Point Island Average Calhounc 1 & 2d Cliffs 1b Ginnab Secog Nuclidea Island 1b 3 & 4e 1 & 2f Ciyr Ciyr Ciyr Ciyr Ciyr Ciyr Ciyr Ciyr Ciyr unit unit unit unit unit unit unit unit unit Cr-51 1.4E-03 ND NA ND NA 1.9E-04 NA NA 3.2E-04 Mn-54 1.1E-04 NA NA 6.3E-05 3.0E-04 6.7E-05 2.7E-06 NA 7.8E-05 Co-57 NA NA NA NA NA NA NA NA NA Co-58 1.1E-03 2.0E-03 NA 1.1E-03 4.8E-04 6.3E-04 4.0E-05 1.2E-02 1.9E-03 Co-60 2.0E-04 2.7E-04 NA 6.0E-04 2.0E-03 7.7E-04 4.5E-05 7.3E-04 5.1E-04 Fe-59 2.3E-04 ND ND ND NA 1.9E-05 NA NA 5.0E-05 Zr-95 2.7E-04 ND ND ND 7.9E-03 4.1E-05 5.7E-06 NA 1.0E-03 Nb-95 1.4E-04 ND ND ND NA 6.0E-05 1.0E-05 NA 3.0E-05 Ru-103 9.1E-05 ND NA ND NA 6.9E-05 2.7E-06 NA 2.3E-05 Ru-106 NA ND NA ND NA 2.4E-05 NA NA 6.0E-06 Sb-125 NA NA NA NA NA NA 7.7E-06 NA 3.9E-06 Cs-134 8.0E-05 1.6E-03 NA 7.9E-04 2.0E-03 3.4E-04 1.5E-06 5.2E-05 5.4E-04 Cs-136 NA ND NA ND NA 1.9E-04 NA NA 4.8E-05 Cs-137 2.0E-04 1.8E-03 NA 1.4E-03 1.9E-03 1.1E-03 9.4E-06 8.0E-05 7.2E-04 Ba-140 NA ND ND ND NA 1.6E-03 NA NA 4.0E-04 Ce-141 1.5E-04 ND NA ND NA 2.8E-05 1.5E-06 NA 2.6E-05 a

Measurements were made downstream of the auxiliary building HEPA filter. Because of the uncertainty in the DFs of the HEPA filter, the data are not considered.

b The data in this column are taken from Table 2-19 of NUREG-0017, Revision 1, and EPRI NP-939.

c The data in this column are taken from Table 2-19 of NUREG-0017, Revision 1, and NUREG/CR-0140.

d The data in this column are taken from Table 2-19 of NUREG-0017, Revision 1, and NUREG/CR-0715.

e The data in this column are taken from Table 2-19 of NUREG-0017, Revision 1, and NUREG/CR-1629.

f The data in this column are taken from Table 2-19 of NUREG-0017, Revision 1, and NUREG/CR-4397.

g The data in this column are taken from Table 2-19 of NUREG-0017, Revision 1, and NUREG/CR-2348.

ND = Not detected. For averaging purposes, a value of zero was assumed.

NA = Not analyzed (or no measurement taken); plants not included in averaging.

4-13

Table 4-11 Measured releases upstream of HEPA filtersfuel pool area Three Turkey Prairie Fort Zion Calvert R.E. Rancho Mile Point Island Average Calhounb 1 & 2c Cliffs 1a Ginnaa Secof Nuclide Island 1a 3 & 4d 1 & 2e Ciyr Ciyr Ciyr Ciyr Ciyr Ciyr Ciyr Ciyr Ciyr unit unit unit unit unit unit unit unit unit Cr-51 1.8E-04 NA NA NA NA NA NA NA 1.8E-04 Mn-54 1.0E-05 NA NA NA 1.2E-03 NA 2.6E-06 NA 2.4E-04 Co-57 NA NA NA NA NA NA NA NA NA Co-58 8.5E-05 NA NA NA 1.1E-02 NA 8.8E-06 6.7E-05 1.8E-03 Co-60 4.4E-05 NA NA NA 5.0E-03 NA 6.9E-06 7.6E-06 8.4E-03 Fe-59 NA NA NA NA NA NA NA NA NA Zr-95 NA NA NA NA NA NA 7.2E-06 NA 3.6E-06 Nb-95 3.0E-05 NA NA NA 9.5E-03 NA 1.7E-05 NA 1.9E-03 Ru-103 9.8E-05 NA NA NA NA NA 1.7E-05 NA 3.8E-05 Ru-106 6.9E-05 NA NA NA NA NA NA NA 6.9E-05 Sb-125 1.7E-04 NA NA NA NA NA ND NA 5.7E-05 Cs-134 9.0E-06 NA NA NA 2.2E-03 NA 9.8E-07 9.6E-07 3.7E-04 Cs-136 NA NA NA NA NA NA NA NA NA Cs-137 2.4E-05 NA NA NA 5.6E-03 NA 4.1E-06 7.4E-07 9.4E-04 Ba-140 NA NA NA NA NA NA NA NA NA Ce-141 NA NA NA NA NA NA 8.8E-07 NA 4.4E-07 a

The data in this column are taken from Table 2-20 of NUREG-0017, Revision 1, and EPRI NP-939.

b The data in this column are taken from Table 2-20 of NUREG-0017, Revision 1, and NUREG/CR-0140.

c The data in this column are taken from Table 2-20 of NUREG-0017, Revision 1, and NUREG/CR-0715.

d The data in this column are taken from Table 2-20 of NUREG-0017, Revision 1, and NUREG/CR-1629.

e The data in this column are taken from Table 2-20 of NUREG-0017, Revision 1, and NUREG/CR-4397.

f The data in this table are taken from Table 2-20 of NUREG-0017, Revision 1, and NUREG/CR-2348.

ND = Not detected. For averaging purposes, a value of zero was assumed.

NA = Not analyzed (or no measurement taken); plants not included in averaging.

4-14

Table 4-12 Measured releases upstream of HEPA filterswaste gas system Three Turkey Prairie Fort Zion Calvert R.E. Rancho Mile Point Island Average Calhounb 1 & 2c Cliffs 1a Ginnaa Secof Nuclide Island 1a 3 & 4d 1 & 2e Ciyr Ciyr Ciyr Ciyr Ciyr Ciyr Ciyr Ciyr Ciyr unit unit unit unit unit unit unit unit unit Cr-51 8.4E-05 ND ND ND NA NA NA NA 1.4E-05 Mn-54 1.1E-05 ND 4.0E-06 ND NA NA NA 8.4E-09 2.1E-06 Co-57 NA NA NA NA NA NA NA NA NA Co-58 4.5E-05 3.8E-06 1.1E-05 8.8E-07 NA NA NA 5.1E-08 8.7E-06 Co-60 8.0E-05 NA 3.2E-06 2.9E-07 NA NA NA 5.9E-08 1.4E-05 Fe-59 7.2E-06 1.8E-06 1.9E-06 ND NA NA NA NA 1.8E-06 Zr-95 1.9E-05 ND ND NA NA NA NA NA 4.8E-06 Nb-95 2.2E-05 ND ND ND NA NA NA NA 3.7E-06 Ru-103 1.9E-05 ND ND ND NA NA NA NA 3.2E-06 Ru-106 1.6E-05 ND ND ND NA NA NA NA 2.7E-06 Sb-125 NA NA NA NA NA NA NA NA NA Cs-134 1.2E-04 1.2E-06 1.1E-04 3.8E-08 NA NA NA 1.1E-08 3.3E-05 Cs-136 3.2E-05 ND ND ND NA NA NA NA 5.3E-06 Cs-137 3.5E-04 NA 1.1E-04 8.8E-08 NA NA NA 2.6E-08 7.7E-05 Ba-140 1.4E-04 ND ND ND NA NA NA NA 2.3E-05 Ce-141 1.3E-05 ND ND ND NA NA NA NA 2.2E-06 a

The data in this column are taken from Table 2-21 of NUREG-0017, Revision 1, and EPRI NP-939.

b The data in this column are taken from Table 2-21 of NUREG-0017, Revision 1, and NUREG/CR-0140.

c The data in this column are taken from Table 2-21 of NUREG-0017, Revision 1, and NUREG/CR-0715.

d The data in this column are taken from Table 2-21 of NUREG-0017, Revision 1, and NUREG/CR-1629.

e The data in this column are taken from Table 2-21 of NUREG-0017, Revision 1, and NUREG/CR-4397.

f The data in this column are taken from Table 2-21 of NUREG-0017, Revision 1, and NUREG/CR-2348.

ND = Not detected. For averaging purposes, a value of zero was assumed.

NA = Not analyzed (or no measurement taken); plants not included in averaging.

4-15

Noble Gas Releases from Building Ventilation Systems Parameter The noble gas releases from building ventilation systems are based on a daily leak rate of 3 percent of the noble gas inventory in the primary coolant released to the containment atmosphere; on a 1.6E+02 pounds per day (lb/d) primary coolant leak to the auxiliary building; and on a 1.7E+03 lb/h steam leak rate in the turbine building.

Basis The containment building leakage rate is derived from xenon-133 measurements in the containment atmosphere at R.E. Ginna and the Maine Yankee NPP (Reference 3). The xenon-133 concentrations in the containment atmospheres at steady state were approximately 5.0E-03 microcuries per cubic centimeter (Ci/cm3) for Maine Yankee and 7.0E-03 Ci/cm3 for R.E. Ginna. The containment volumes at these facilities are approximately 1.8E+06 ft3 for Maine Yankee and 1.0E+06 ft3 for R.E. Ginna. Based on these values, the total Ci of xenon-133 in the containment building atmosphere are as follows:

Maine Yankee (5.0E-03 Ci/cm3) (1.8E+06 ft3) (2.83E+04 cm3/ft3) = 2.5E+08 Ci xenon-133 R.E. Ginna (7.0E-03 Ci/cm3) (1.0E+06 ft3) (2.83E+04 cm3/ft3) = 2.0E+08 Ci xenon-133 Based on the half-life of xenon-133 (5.3 days) and the assumption of a constant leakage rate to containment, the daily leakage rate of xenon-133 to the containment for the two plants is as follows:

Maine Yankee 2.5E+08 Ci Ci

= 3.3E+07 d xenon-133 leakage 5.3E+00 d6.93E-01 R.E. Ginna 2.0E+08 Ci Ci

= 2.6E+07 d xenon-133 leakage 5.3E+00 d6.93E-01 Based on the xenon-133 concentration during power operation and the masses of primary coolant of the two plants, the fraction of the xenon-133 inventory in the containment released per day is as follows:

Maine Yankee Ci 3.3E+07 d = 1.0E-02d = 1 Percentd 1.0E+01 Ci cm 3 3 x 2.83E+04 3 x 1.1E+04 ft cm3 ft 4-16

R.E. Ginna Ci 2.6E+07 d = 5.0E-03d = 0.5 Percentd 3.0E+01 Ci cm3 x 6.234E+03 ft 3 x 2.83E+04 cm3 ft 3 NUREG-0017, Revision 1, also contains data for the xenon-133 concentration in the containment atmosphere and the primary coolant at Yankee Rowe for the periods August-October 1971, December 1971-January 1972, and August-November 1973. These periods encompass several shutdowns and a wide variety of operating conditions, and during these periods, the xenon concentration in the containment and in the primary coolant varied by 2 orders of magnitude. The percent of xenon-133 inventory in the coolant released to the containment atmosphere varied from approximately 0.05 percent/day to 0.5 percent/day. Also, for Rancho Seco, this percent was determined to be 10.4 (Reference 21).

These data indicate that 3 percent per day of the noble gas inventory in the primary coolant is released to the containment atmosphere.

In the auxiliary building, the source term calculation is based on an assumed primary coolant leakage rate of 1.6E+02 lb/d (2.0E+01 gal/d). In the absence of available data, this value is based on engineering judgment and is consistent with values proposed in environmental reports.

In the turbine building, it is assumed that steam leaks to the turbine building atmosphere at a rate of 1.7E+03 lb/h. The leakage is considered to be from many sources, each too small to be detected individually, but when taken collectively, they total 1.7E+03 lb/h. The most significant leakage pathway is considered to be leakage through valve stem packings.

Steam Generator Blowdown Flash Tank Vent Parameter The PWRs with U-tube steam generators that are currently under design either direct their blowdown through a heat exchanger to cool the blowdown or, if a flash tank is used, vent the flash tank to a flash tank vent condenser or the main condenser. For these plants, radioiodine releases by this path are negligible, and a PF of zero is used for the steam generator blowdown flash tank vent.

For older plants that still use flash tanks that vent directly to the atmosphere, a radioiodine PF of 5.0E-02 is used.

Basis Approximately one-third of the blowdown stream flashes to steam in the flash tank, provided that there is a heat balance between steam generator operating conditions (550 degrees Fahrenheit (°F), 1,000 pounds per square inch atmospheric), and the blowdown flash tank conditions (240 degrees F, sat.). Although the radioiodine species in the blowdown stream will be predominantly nonvolatile (volatile species are degassed in the steam generator),

significant radioiodine removal will occur because of entrainment by the flashing steam. The evaluation considers a steam quality of 85 percent. For currently designed PWRs that have provisions to prevent flashing (cooling blowdown below 212 degrees F) or to condense the steam leaving the flash tank, the entrainment losses will be negligible (i.e., a PF of zero).

4-17

Iodine Released from Main Condenser Air Ejector Exhaust Parameter The code calculates radioiodine releases from the main condenser air ejector exhaust before treatment by using the data in Tables 4-13, 4-37, and 4-38.

Basis The radioiodine releases from the main condenser air ejector exhaust are based on secondary side measurements at Point Beach Nuclear Plant, Unit 2 (Reference 16); Turkey Point, Units 3 and 4 (Reference 19); Point Beach, Unit 1 (Reference 3); and Connecticut Yankee NPP (Reference 3).

Table 4-13 Annual radioiodine normalized releases from main condenser air ejector exhaust Normalized Release Data Sourcea (Ci/yr/µCi/g)b Turkey Point 3 & 4 3.5E+03 Point Beach 1 & 2 6.1E+02 Connecticut Yankee 3.0E+01 Average 1.7E+03 a The data in this table are taken from Table 2-22 of NUREG-0017, Revision 1.

b The normalized release rate represents the effective release rate for radioiodine. It is the combination of the stream flow to the main condenser, the partitioning of the radioiodine between the main condenser and the air ejector exhaust where it is measured, and the PC for radioiodine from water to steam in the steam generator.

In a manner similar to that discussed for normalized releases for building ventilation releases in Section 4.1.1.3, the main condenser air ejector exhaust radioiodine releases are directly related to the secondary coolant iodine-131 concentration. Therefore, for the air ejector exhaust, the normalized radioiodine release, RN, is determined using Equation (4-5):

RA RN = C (4-5)

RW x PC where RN = normalized effective release rate of iodine-131 (Ci/yr/Ci/g);

RA = absolute (measured) iodine-131 release rate (Ci/yr);

CRW = measured secondary coolant iodine-131 concentration (Ci/g); and PC = measured radioiodine PC from secondary coolant water to steam in the steam generator.

Table 4-13 presents data on normalized release rates from the main condenser air ejector exhaust. To obtain the release in Ci/yr from the air ejector exhaust of a particular PWR, the code multiplies the normalized release data in Table 4-13 by the radioiodine concentration in the secondary coolant water and the radioiodine PC from the water to steam for that particular PWR using Equation (4-6):

4-18

RPWRi = RN x CPWRi x PCPWR (4-6) where RPWRi = calculated annual release for particular PWR for radioiodine isotope i (Ci/yr);

RN = normalized annual release rate of radioiodine from Table 4-13 (Ci/yr/Ci/g);

CPWRi = calculated secondary coolant concentration for particular PWR for radioiodine isotope i (Ci/g); and PCPWR = radioiodine PC from water to steam in the steam generator for the particular PWR (see Table 4-41).

As discussed in EPRI NP-939 and NUREG/CR-1629, most of the radioiodine in the secondary system is not available for release to the main condenser air ejector exhaust because the radioiodine bypasses the condenser hotwell in the moisture separator/reheater drains and extraction steam, and possibly because of radioiodine plating out in the moisture separator/reheater, turbine, and main condenser. As a result, the radioiodine release from the main condenser air ejector exhaust is small compared to the building ventilation releases.

Containment Purge Frequency Parameter For those plants equipped with small-diameter purge lines (diameter of about 8 inches or less),

releases are based on continuous ventilation during power operation and on two purges per year at cold shutdown with the large containment purge lines. The continuous ventilation rate used in the evaluation is based on the applicants design.

For older plants (those under review for operating licenses or those for which the construction permit safety evaluation report was issued before July 1, 1975) not equipped with small-diameter purge lines, releases are based on 2 purges per year at cold shutdown and 22 purges per year during power operation. The 22 purges consider the effect of using large containment purge lines and separate vent lines, if any. If, for a specific plant, there is filtration on the large purge lines but not on the vent lines, an additional GALE code run will be made to account for the effect of the vent.

Operating experience and special design features (for example, little or no air-operated equipment or the containment) to reduce the frequency of containment purging will be considered on a case-by-case basis.

Basis It is assumed that the containment building is purged twice a year for refueling and maintenance. The two purges are considered for cold shutdowns for annual fuel loading and planned maintenance. In addition, experience at operating reactors (Table 4-14) has shown a need to purge or vent the containment frequently during full-power operation and hot standby to control the containment pressure, temperature, humidity, and airborne activity levels (Reference 3). For the above reasons, NUREG-75/087, Revision 1, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, issued September 1975 (ADAMS Accession No. ML081510817) [Ref. 23], Section 6.2.4, Containment Isolation System, indicates that new plant designs should include the capability to purge the containment continuously through small-diameter purge lines (about 8-inch diameter) and use the large containment purge lines only at cold shutdowns and refueling outages. On this basis, source term calculations for new plants should consider a continuous ventilation rate based on the applicants containment design, along with the two cold shutdown purges per year with the large 4-19

containment purge lines, unless special provisions are made to eliminate or reduce the need for continuous ventilation flow.

For older plants (those under review for operating licenses or those for which the construction permit safety evaluation report was issued before July 1, 1975) not equipped with small-diameter purge lines, frequent periodic purges or vents will be used to control the above parameters (Reference 19). A frequency of 22 purges per year during power operation is considered representative of plant operating experience for the combined effects of purging and venting.

Table 4-14 PWR containment purging and venting experience Yankee Rowea Purge and vent only after cooldown following shutdown.

Routinely pressurize containment for leak detection system checks and bring Reasons:

activity down.

Duration: 2 to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Maine Yankeea Purge once per quarter.

Reasons: Bring activity down.

Duration: 2 to 3 days each quarter.

Indian Point 2a Vent two times each day.

Reasons: Pressure balance control.

Duration: Approximately 1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Purge once every 2 weeks (duration not stated).

Three Mile Island 1a Purge approximately once per week during operation; always purge before shutdown.

Reasons: Improve temperature and humidity conditions.

Duration: Approximately 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Connecticut Yankeea PurgeCannot purge during operation, only during shutdown.

Reasons: Primarily to remove activity.

Duration: 1 to 2 days.

a The data in this table are taken from Table 2-23 of NUREG-0017, Revision 1.

b Generally, long purges occur during plant outages while at cold shutdown conditions.

c Data for these plants were obtained from the semiannual release reports for the plants for the period indicated.

4-20

Table 4-14 PWR containment purging and venting experience (cont.)

San Onofre 1a Purge each cooldown approximately four times per year; no purging during power operation.

Purge for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; ventilate during entire shutdown period.

Oconee 1a Continuous purge from startup through July 1, 1974.

Purged twice since July 1, 1974: once on July 8, 1974, for several days and again on August 22, 1974, for 1 to 2 days.

Reasons: Reduction of gaseous activity for maintenance, etc.

Oconee 2a Continuous purge from startup, lowest purge rate approximately 2.0E+04 ft3/min.

Reasons: Reduction of gaseous activity for maintenance, etc.

H.B. Robinson 2a Purge approximately 20 times per year for 2 minutes each purge to test purge valves. In addition, purge approximately 10 times per year for an average of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months each purge for personnel comfort reasons.

Vent about 75 times per year for about 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months each. Venting occurs to control containment pressure and to bring containment pressure to zero gauge before purging as noted above.

Turkey Point 3a For period January 1 to July 1, 1974.

Total purges 14 Total time 502 hour0.00581 days 0.139 hours 8.300265e-4 weeks 1.91011e-4 months sb Maximum duration (one purge) 253 hours0.00293 days 0.0703 hours 4.183201e-4 weeks 9.62665e-5 months Minimum duration (one purge) 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months Infrequent purges or vents of 10 minutes for pressure control.

Turkey Point 4a For period January 1 to July 1, 1974.

Total purges 5 Total time 984 hour0.0114 days 0.273 hours 0.00163 weeks 3.74412e-4 months sb Maximum duration (one purge) 742 hours0.00859 days 0.206 hours 0.00123 weeks 2.82331e-4 months Minimum duration (one purge) 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months a The data in this table are taken from Table 2-23 of NUREG-0017, Revision 1.

b Generally, long purges occur during plant outages while at cold shutdown conditions.

c Data for these plants were obtained from the semiannual release reports for the plants for the period indicated.

4-21

Table 4-14 PWR containment purging and venting experience (cont.)

Surry 1 and 2a Containment operates at negative pressure. Discharge from vacuum pumps through filters to stack. During cold shutdown, there is continuous purging of containment.

Prairie Islanda Frequency: Once per week for about 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

To relieve pressure buildup resulting from instrument air leakage to Reasons:

containment.

Kewauneea Frequency: Five times in 60 days usually for less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , longer if for personnel entry.

Pressure control. During the 60-day period, purging occurred for personnel Reasons:

entry.

Point Beacha Continuous venting through a monitoring line at about 10 ft3/min flow. Gas filtered on way to stack.

Palisadesa Once per week for a duration of about 10 minutes (planned upon resumption Frequency:

of power operation).

Reasons: To control pressure buildups.

Ziona Venting for pressure buildup about twice per week depending on outside temperature. Ranges from twice per day to once every 2 weeks.

Purges to control environment range from once per day to once every 2 weeks.

Duration: 3/4 hour on venting; 3 to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months on purging.

Fort Calhouna,c For periods from July 1 to June 30, 1976, and May 5 to December 31, 1977.

Average 65 purges per year with an average duration of about 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

Millstone 2a,c For period July 1, 1975, through December 31, 1977.

About 45 purges per year with an average duration of about 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months .

a The data in this table are taken from Table 2-23 of NUREG-0017, Revision 1.

b Generally, long purges occur during plant outages while at cold shutdown conditions.

c Data for these plants were obtained from the semiannual release reports for the plants for the period indicated.

4-22

Containment Internal Cleanup System Parameter Assume the internal cleanup system will operate for 1.6E+01 hours before purging, that it provides a DF for radioiodine removal on charcoal adsorbers corresponding to the values in Table 4-15 and a DF of 1.0E+02 for particulate removal on HEPA filters, and that there is a containment air mixing efficiency of 70 percent.

Table 4-15 Removal efficiencies for charcoal adsorbers for iodine removal Iodine Removal Efficiencies for Charcoal Adsorbersa Percent Removal Activated Carbon Bed Depthb Efficienciesc (Percent)

Air filtration system designed to operate inside reactor 2 inches 90 containment.

Air filtration system designed to operate outside the 2 inches reactor containment and relative humidity is controlled 70 at 70 percent.

Air filtration system designed to operate outside the 4 inches reactor containment and relative humidity is controlled 90 at 70 percent.

Air filtration system designed to operate outside the 6 inches reactor containment and relative humidity is controlled 99 to 70 percent.

a These values are from Table 1-5 and Section 2.2.11.1 of NUREG-0017, Revision 1, and Table 2 of RG 1.140, Revision 1.

b Multiple beds (e.g., two 2-inch beds in series) should be treated as a single bed of aggregate depth of 4 inches.

c The removal efficiencies assigned to HEPA filters for particulate removal and charcoal adsorbers for radioiodine removal are based on the design, testing, and maintenance criteria recommended in RG 1.140, Revision 1.

Basis Internal cleanup systems may be used to reduce airborne radioiodine concentrations in the containment air before purging. Such systems normally recirculate containment air through HEPA filters and charcoal adsorbers to effect radioiodine and particulate removal. For source term calculations, it is assumed that the cleanup systems are operated for 1.6E+01 hours before purging. It is considered that charcoal adsorbers provide a DF for radioiodine corresponding to the values in Table 4-15 that HEPA filters provide a DF of 1.0E+02 for particulates, and that the containment air mixing efficiency is 70 percent. The system operation time of 1.6E+01 hours considers that two shifts will elapse following a decision to enter the containment. The time period of two shifts is a reasonable amount of time for preentry preparations.

4-23

A 70-percent mixing efficiency, based on data from the R.E. Ginna containment building atmosphere test conducted in 1971 (Reference 3), is used in evaluations. Table 4-16 shows these data, which are discussed below.

Table 4-16 Data from the R.E. Ginna containment building atmosphere test Parametera Symbol Value Length of test run T 6.0E+00 hours Initial radioiodine activity Ao 1.2E-08 Ci/cm3 Final radioiodine activity A 1.2E-09 Ci/cm3 Containment volume V 1.0E+06 ft3 Internal recirculation system flow rate F 6.1E+05 ft3/h a The data in this table are taken from Section 2.2.10.2 of NUREG-0017, Revision 1.

The efficiency of radioiodine removal, E, can be estimated from Equation (4-7):

FET Ao A

= V (4-7)

Substituting the R.E. Ginna data into Equation (4-7) results in the following:

(6.1E+05) (E) (6.0E+00) 1.0E-08

= 1.0E+06 1.0E-09 1.0E+01 = (3.7E+00) (E) ; therefore, E = 6.3E-01 The radioiodine removal efficiency, E, is a function of filter efficiency, Ea, and mixing efficiency, Em, as shown in Equation (4-8):

E = (Ea )(Em ) = 6.3E-01 (4-8)

In calculating Em, use the assumed DF of 10 for charcoal derived from Table 4-15 (90-percent removal). Using Ea equal to 9.0E-01, Em is calculated as follows:

(E)

Em = (E ) = 6.3E-019.0E-01 = 7.0E-01 or 70 percent a

Radioiodine Removal Efficiencies for Charcoal Adsorbers and Particulate Removal Efficiencies for HEPA Filters Parameter Use a removal efficiency of 99 percent for particulate removal by HEPA filtration. For charcoal adsorbers, Table 4-15 shows removal efficiencies for all forms of radioiodine (from Table 1-5 of NUREG-0017, Revision 1, and Table 1 of RG 1.140, Revision 1).

4-24

Basis The removal efficiencies assigned to HEPA filters for particulate removal and charcoal adsorbers for radioiodine removal are based on the design, testing, and maintenance criteria recommended in RG 1.140, Revision 3.

Waste Gas System Input Flow to Pressurized Storage Tanks Parameter In this gaseous radwaste treatment system, one storage tank is held in reserve for back-to-back shutdowns, one tank is in the process of filling, and the others are used for storage. The GALE-PWR 3.2 code will calculate the effective holdup time for filling and add it to the holdup time for storage. The input flow rate to the pressurized storage tanks varies depending on the system design as shown in Tables 4-16 and 4-17. Therefore, each applicant should supply the value of the waste gas system input flow (F) to the pressurized storage tanks.

Basis As Tables 4-16 and 4-17 show, the waste gas system input flow varies among PWR system designs. Calculations are based on a waste gas input flow rate of 1.4E+02 cubic feet per day (ft3/d) (5.0E+04 cubic feet per year (ft3/yr)) and a storage tank pressure that is 70 percent of the design value. If the calculated holdup time exceeds 9.0E+01 days, assume the remaining gases are released after 9.0E+01 days. The holdup time (Th) and fill time (Tf) are calculated as follows in Equations (4-9) and (4-10):

P V (n-2)

Th = F (4-9)

PV Tf = (4-10)

F where n = the number of tanks; (n - 2) = the correction to subtract the tank being filled and the tank held in reserve; P = the storage pressure, in atmospheres; Tf = the time required to fill one tank, in days; Th = the holdup time, in days; V = the volume of each tank, in ft3 at standard temperature; and F = the waste gas flow rate to pressurized storage tanks in ft3/d.

A review of the waste gas processing systems proposed for a number of PWRs as given in the respective FSARs yielded the design flow rates shown in Tables 4-17 and 4-18. Table 4-17 indicates that for PWRs designed without recombiners to treat the gas before holdup in pressurized storage tanks, the average expected flow is approximately 1.7E+02 ft3/d (standard temperature and pressure (STP)) per reactor. Table 4-18 indicates that for PWRs designed with recombiners to remove hydrogen before holdup in pressurized storage tanks, the average expected flow is approximately 3.0E+01 ft3/d (STP) per reactor.

4-25

Table 4-17 Waste gas system input flow to pressurized storage tanks for PWRs without recombiners Net Flow per Reactor Reactora ft3/d (STP)b San Onofre 2 & 3 5.7E+01 Waterford 3 1.71E+02 Pilgrim 2 6.9E+01 St. Lucie 1 & 2 1.39E+02 Millstone 2 4.9E+01 Arkansas 1 & 2 6.8E+01 Byron 1 & 2 1.73E+02 Sequoyah 1 & 2 1.73E+02 Marble Hill 1 & 2 1.73E+02 Diablo Canyon 1 & 2 3.43E+02 Trojan 2.25E+02 Oconee 1, 2, & 3 1.8E+02 Davis-Besse 1 1.44E+02 Bellefonte 1 & 2 1.63E+02 a The data in this table are taken from Table 2-24 of NUREG-0017, Revision 1.

b Average net flow for PWRs without recombiners = 1.7E+02 ft3/d (STP) per reactor.

Table 4-18 Waste gas system input flow to pressurized storage tanks for PWRs with recombiners Net Flow per Reactor Reactora ft3/d (STP)b Washington Nuclear Project 1 9.6E+01 Farley 1 & 2 3E+00 McGuire 1 & 2 1.8E+01 a The data in this table are taken from Table 2-25 of NUREG-0017, Revision 1.

b Average net flow for PWRs without recombiners = 3.0E+01 ft3/day (STP) per reactor. Net flow rate is determined downstream of any recombiner (which is assumed 100-percent effective in removing hydrogen).

4-26

Holdup Times for Charcoal Delay System Parameter Holdup times for charcoal delay systems are calculated using Equation (4-11):

(M)(K)

T = (1.1E-02) F (4-11) where T = the holdup time, in days; F = the system flow rate, in cubic feet per min (ft3/m) (see guideline above for pressurized storage tanks);

K = the dynamic adsorption coefficient, in cubic centimeters per gram (cm3/g);

M = the mass of charcoal adsorber, in 1.0E+03 lb; and 1.1E-02 = the factor to convert from (1.0E+03 lb cm3/g)/(ft3/min) to days, Equation (4-12).

3 g 3 (1.0E+03 lb) cm g 4.54E+02 lb 3.53E-03 ft 3 (M)(K) cm

() = 3 = 1.1E-02 (4-12) ft min 1.44E+03 mind F Table 4-19 shows the dynamic adsorption coefficients, K, (in cm3/g).

Table 4-19 Dynamic adsorption coefficient for xenon and krypton Dynamic Adsorption Coefficient (cm3/g)

Operating 77 °F Operating 77 °F Operating 77 °F Operating 0 °F Nuclidea Dew Point 45 °F Dew Point 0 °F Dew Point -40 °F Dew Point -20 °F Kr 1.85E+01 2.5E+01 7.0E+01 1.05E+02 Xe 3.3E+02 4.4E+02 1.16E+03 2.41E+03 a The data in this table are taken from Section 2.2.13.1 of NUREG-0017, Revision 1.

Basis Charcoal delay systems are evaluated using Equation (4-11) and dynamic adsorption coefficients. Equation (4-11) is a standard equation for the calculation of delay times in charcoal adsorption systems (Reference 3). The dynamic adsorption coefficients (K) for xenon and krypton are dependent on operating temperature and moisture content (Reference 3) in the charcoal, as indicated by the values in the above parameter. The K values represent a composite of data from operating reactor charcoal delay systems and reports concerning charcoal adsorption systems (Reference 3).

The factors influencing the selection of K values are:

operational data from the Gundremmingen, Lingen, and Vermont Yankee NPPs (Reference 3) 4-27

the effect of temperature on the dynamic adsorption coefficients, indicated in Figure 4-1 (Reference 3)

the effect of moisture on the dynamic adsorption coefficients, shown in Figure 4-2

the affinity of charcoal for moisture, shown in Figure 4-3

the variation in K values between researchers and between the types of charcoal used in these systems (Reference 3). Because of the variation in K values in different types of charcoal and in the data reported, average values taken from the Gundremmingen and Lingen NPPs (Reference 3) shown in Figure 4-1 are used.

Figure 4-1 Krypton and xenon K values as a function of reciprocal temperature 4-28

Figure 4-2 Effect of moisture content on the dynamic adsorption coefficient Figure 4-3 Charcoal moisture as a function of relative humidity 4-29

Liquid Waste Inputs Parameter The PWR liquid waste flow rates listed in Table 4-20 are used as inputs to the liquid radwaste treatment system. Flows that cannot be standardized are added to those listed in Table 4-20 to fit an individual application (e.g., shim bleed and equipment leaks to the reactor coolant drain tank). Disposition of liquid streams to the appropriate collection tanks is based on the applicants proposed method of processing.

Basis The flow rates used represent average values for a plant operating at steady-state conditions.

The values are derived from values proposed by ANSI/ANS-55.6-1979, Liquid Radioactive Waste Processing System for Light Water Reactor Plants [Ref. 24], from operating and design data, and from information furnished by applicants in response to source term questions. Data from Zion (Reference 18) indicate that the values for fraction of primary coolant activity given in Table 4-20 are reasonable estimates of plant operating experience.

Detergent Waste Parameter For plants with an onsite laundry, use the radionuclide distribution given in Table 4-21 for untreated detergent wastes. The quantities shown in Table 4-21 should be added to the adjusted liquid source term. If treatment is provided, detergent waste releases should be reduced, using appropriate DFs from this report.

Basis In the evaluation of liquid radwaste treatment systems, it is assumed that detergent wastes (laundry and personnel drains) will have the radionuclide distribution given in Table 4-21. The radionuclide distribution is based on measurements at four NPPs, shown in Table 4-22.

4-30

Table 4-20 PWR liquid wastes Expected Daily Average Input Flow Rate (gal/d)

Type of Treatment of Blowdown Recycled to Secondary System (U-tube steam generator plants) or Type of Treatment of Condensate (once-through steam Sourcea generator plants)

Deep-Bed Cond. Deep-Bed Cond.

Plant with Demineralizers Demineralizers Filter Fraction of Blowdown with Ultrasonic without Ultrasonic Demineralizer PCA Treatmentb Resin Cleaner Resin Cleaner

1. Reactor Containment Primary coolant pump seal leakage 2.0E+01 2.0E+01 2.0E+01 2.0E+01 1.0E-01 Primary coolant leakage, 1.0E+01 1.0E+01 1.0E+01 1.0E+01 1.67E+00c miscellaneous sources Primary coolant equipment 5.0E+02 5.0E+02 5.0E+02 5.0E+02 1.0E-03

2. Primary Coolant Systems (outside of containment)

Primary coolant system equipment 8.0E+01 8.0E+01 8.0E+01 8.0E+01 1.0E+00 drains Spent fuel pit liner drains 7.0E+02 7.0E+02 7.0E+02 7.0E+02 1.0E-03 Primary coolant sampling system 2.0E+02 2.0E+02 2.0E+02 2.0E+02 5.0E-02 drains Auxiliary building floor drains 2.0E+02 2.0E+02 2.0E+02 2.0E+02 1.0E-01

3. Secondary Coolant Systems Secondary coolant sampling 1.4E+03 1.4E+03 1.4E+03 1.4E+03 1.0E-04 system drains Condensate demineralizer rinse 3.0E+03 1.2E+04 -- -- 1.0E-08 and transfer solutions Condensate demineralizer Calculated in 8.5E+02 3.4E+03 -- --

regenerant solutions GALE Code Ultrasonic resin cleaner solutions 1.5E+04 -- -- -- 1.0E-06 Condensate filter-demineralizer

-- -- 8.1E+03 -- 2.0E-06 backwash Plant Plant Steam generator blowdown -- -- --

dependentd dependentd Calculated in Turbine building floor drains 7.2E+03 7.2E+03 7.2E+03 7.2E+03 GALE Code

4. Detergent and Decontamination Systems Onsite laundry facility 3.0E+02 3.0E+02 3.0E+02 3.0E+02 e Hot showers Negligible Negligible Negligible Negligible --

Hand wash sink drains 2.0E+02 2.0E+02 2.0E+02 2.0E+02 e Equipment and area 4.0E+01 4.0E+01 4.0E+01 4.0E+01 e decontamination Totals 2.97E+04 2.63E+04 1.9E+04 1.0E+04 a

The data in this table are taken from Table 2-26 of NUREG-0017, Revision 1.

b The product not recycled to condenser or secondary coolant system.

c About 40 percent of the leakage flashes, resulting in a PCA fraction of the leakage greater than 1.0.

d Input parameter.

e GALE code uses release data given in Table 4-21 to calculate releases from this source.

4-31

Table 4-21 Calculated annual release of radioactive materials in untreated detergent waste Radionuclidea Ci/yr per reactor P-32 1.8E-04 Cr-51 4.7E-03 Mn-54 3.8E-03 Fe-55 7.2E-03 Fe-59 2.2E-03 Co-58 7.9E-03 Co-60 1.4E-02 Ni-63 1.7E-03 Sr-89 8.8E-05 Sr-90 1.3E-05 Y-91 8.4E-05 Zr-95 1.1E-03 Nb-95 1.9E-03 Mo-99 6.0E-05 Ru-103 2.9E-04 Ru-106 8.9E-03 Ag-110m 1.2E-03 Sb-124 4.3E-04 I-131 1.6E-03 Cs-134 1.1E-02 Cs-136 3.7E-04 Cs-137 1.6E-02 Ba-140 9.1E-04 Ce-141 2.3E-04 Ce-144 3.9E-03 Total 9.0E-02 Ci a

The data in this table are taken from Table 2-27 of NUREG-0017, Revision 1.

4-32

Table 4-22 Radionuclide distribution of detergent waste Oyster Creek R.E. Ginna Zionb Fort Calhoun Radionuclidea (1971-1973) (1972-1973) (1977) (1977)

(Ci/month) (Ci/month) (Ci/month) (Ci/month)

P-32 1.5E-02 NA NA NA Cr-51 2.3E-01 NA 9.4E-01 NA Mn-54 1.3E+00 1.2E-01 1.6E-01 1.9E-02 Fe-55 3.5E-01 NA 1.9E+00 1.6E-01 Fe-59 2.9E-01 NA 2.6E-01 NA Co-58 3.5E-01 4.1E-01 2.4E+00 1.5E-01 Co-60 3.8E+00 9.0E-01 9.8E-01 3.0E-02 Ni-63 NA NA 3.5E-01 7.1E-02 Sr-89 2.1E-02 NA 7.0E-03 1.4E-03 Sr-90 2.5E-03 NA 7.6E-04 NA Y-91 NA NA 1.4E-02 NA Zr-95 8.3E-02 1.6E-01 1.4E-01 NA Nb-95 1.6E-01 2.0E-01 2.7E-01 NA Mo-99 NA 5.0E-03 NA NA Ru-103 1.3E-02 3.2E-02 5.2E-02 NA Ru-106 NA 7.4E-01 NA NA Ag-110m NA 1.0E-01 NA NA Sb-124 6.1E-02 NA 4.7E-02 NA I-131 4.3E-01 5.5E-02 1.7E-01 1.7E-02 Cs-134 1.7E-01 1.4 1.5E+00 1.4 Cs-136 NA NA 6.2E-02 NA Cs-137 2.9E-01 2.5E+00 2.1E+00 1.7E+00 Ba-140 7.6E-02 NA NA NA Ce-141 3.3E-02 5.0E-03 NA NA Ce-144 7.3E-02 5.8E-01 NA NA Total 7.7E+00 7.2E+00 1.14E+01 3.5E+00 a

The data in this table are taken from Table 2-28 of NUREG-0017, Revision 1.

b For two units.

NA = radionuclide was not analyzed.

Chemical Wastes from Regeneration of Condensate Demineralizers Parameter

Liquid flows to demineralizer at main steam activity.

All radionuclides removed from the secondary coolant by the demineralizers are removed from the resins during regeneration.

4-33

Use a regeneration cycle of 1.2 days times the number of demineralizers for a deep-bed condensate system without ultrasonic resin cleaner (URC). For systems with URC, use a regeneration cycle of 8 days times the number of demineralizers.

Basis Operating data (Reference 3) from Arkansas Nuclear One, Unit 1, indicate that one condensate demineralizer (without URC) is chemically regenerated every 1.2 days. The 8-day period for systems using URC is taken from ANSI/ANS-55.6-1979.

All material exchanged or filtered out by the resins between regenerations is contained in the regenerant waste streams. Therefore, each regeneration will have approximately the same effectiveness (i.e., each regeneration removes all material collected since the previous regeneration, leaving a constant quantity of material on the resins after regeneration).

Regeneration cycles are normally controlled by particulate buildup on resin beds, resulting in high pressure drops across the bed.

Tritium Releases Parameter The tritium releases through the combined liquid and vapor pathways are 4.0E-01 Ci/yr per MWt. The quantity of tritium released through the liquid pathway is based on the calculated volume of liquid released, excluding secondary system wastes, with a primary coolant tritium concentration of 1.0E+00 Ci/ml up to a maximum of 90 percent of the total quantity of tritium calculated to be available for release. It is assumed that the remainder of the tritium produced is released as a gas from building ventilation exhaust systems. About 80 percent of the tritium in the gaseous effluents is released from the auxiliary building ventilation system, including the refueling area, and the remaining 20 percent is released from the containment building ventilation system.

Basis The release rate of 4.0E-01 Ci/yr per MWt is based on a review of the tritium release rates at a number of PWRs and on data from specific measurements of tritium inventory and tritium releases at the R.E. Ginna (Reference 3). The measurements at R.E. Ginna were made during the first two core cycles, during which the reactor operated 605 effective full-power days. The observed tritium buildup during this period was 1.41E+03 Ci. For the same period, 9.1E+05 MWd of thermal power were generated. Using these data and considering an 80-percent plant capacity factor and tritium decay, Equation (4-13) shows the annual average tritium release:

Ci 1.41E+03 Ci9.1E+05 (8.0E-01)3.65E+02 dyr 6.93E-01(1)1.23E+01 = 4.3E-01 yr MWt (4-13)

Table 4-23 gives the reported liquid and gaseous tritium releases for 1972-1978 for 35 operating PWRs that use Zircaloy-clad fuel and started commercial operation before 1978.

Table 4-23 shows these data expressed as the average release rate from the plants as a function of the number of years of operation of each plant. The tritium release rate from a PWR should reach a steady-state value after a few years as a result of leakages from the plant.

Table 4-24 illustrates that the tritium release rate is approaching a steady-state value of approximately 4.0E-01 Ci/yr per MWt, which is the value obtained from the R.E. Ginna 4-34

measurements. At steady state, the release rate from a plant is approximately equal to the amount entering the primary coolant since only about 5 percent per year of the plant tritium inventory will decay. Based on the data from R.E. Ginna and the data in Table 4-24, use a release rate of 4.0E-01 Ci/yr per MWt, which considers both liquid and vapor pathways.

The amount of tritium released via the liquid pathway is calculated from the volume of primary coolant released in the nonrecyclable waste streams for the boron recovery, clean waste, and dirty waste systems. The concentration of tritium in wastes originating from primary coolant is assumed to be 1.0E+00 µCi/ml, consistent with ANSI/ANS-18.1-1984. Tritium in liquid that leaks into, or is used as makeup to, the secondary system is considered to be released in liquid effluents through the turbine building floor drain discharge. The parameters for primary coolant activity before processing are used to calculate the tritium concentration in the waste streams.

Data in Table 4-25 indicate that tritium released in liquid effluents can make up a large fraction of the total tritium produced. Therefore, the tritium calculated to be released in liquid effluents is up to a maximum of 90 percent of the total quantity of tritium calculated to be available for release.

The difference between the tritium calculated to be available for release from the primary coolant and the tritium calculated to be released in liquid effluents is considered to be released as a vapor through building ventilation exhaust systems. Table 4-26 provides the distribution of tritium released from various sources within the plant based on measurements made at the following plants:

R.E. Ginna, Calvert Cliffs, and Three Mile Island Nuclear Station, 1975 through 1977 (Reference 16)

Zion, Units 1 and 2, 1976 and 1977 (Reference 18)

Turkey Point, 1977 (Reference 19)

Prairie Island, Units 1 and 2, 1980 and 1981 (Reference 20)

Rancho Seco, 1978 and 1979 (Reference 21)

Based on data in Table 4-26, approximately 32 percent of tritium in gaseous effluents is released from the auxiliary building, 50 percent from the refueling area, and 18 percent from the containment. Since the refueling area in a PWR generally vents to the same release point as the auxiliary building, these two releases are included together in the parameter.

4-35

Table 4-23 Tritium release data from PWRs with Zircaloy-clad fuels Power Nuclear Thermal Output/Unit Startup per (1.0E+06 MWdt)b Reactora Date Unit (year) 1972 1973 1974 1975 1976 1977 (MWt)

R.E. Ginna 1520 1969 0.32 0.45 0.28 0.40 0.29 0.46 H.B. Robinson 2 2200 1970 0.62 0.51 0.39 0.57 0.66 0.59 Point Beach 1 & 2 1518 1970/72 0.42 0.77 0.43 0.81 0.91 0.93 Palisades 2530 1971 0.24 0.27 0.02 0.37 0.40 0.72 Maine Yankee 2440 1972 0.48 0.61 0.81 0.69 Indian Point 2 & 3 2758 1973/76 0.48 0.69 0.56 1.46 Surry 1 & 2 2441 1972/73 0.80 1.21 1.05 1.27 Turkey Point 3 & 4 2200 1972/73 1.08 1.16 1.12 1.13 Oconee 1, 2, & 3 2568 1973/74/72 0.51 1.95 1.65 1.67 Zion 1 & 2 3250 1973/73 1.37 1.29 1.53 Fort Calhoun 1420 1973 0.28 0.30 0.39 Prairie Island 1 & 2 1650 1973/74 0.94 0.86 1.03 Kewaunee 1650 1974 0.45 0.45 0.46 Three Mile Island 1 2535 1974 0.73 0.58 0.73 Rancho Seco 2772 1974 0.17 0.27 0.75 Arkansas 1 2568 1974 0.64 0.50 0.68 Calvert Cliffs 1 & 2 2700 1974/76 0.58 0.84 1.24 Cook 1 3250 1975 0.90 0.64 Millstone 2 2530 1975 0.63 0.59 Trojan 3411 1975 0.31 0.88 St. Lucie 1 2560 1975 0.73 Beaver Valley 1 2352 1976 0.42 Salem 1 338 1976 0.28 a These values are from Table 2-29 of NUREG-0017, Revision 1.

b Data from semiannual reports of reactors listed.

c No reported data.

d No radioactive liquid wastes were discharged from Unit 2 during the entire year. In 1975, no radioactive liquid wastes were discharged from Unit 1 during the last 6 months.

e Rancho Seco is designed to be a zero or very low liquid release plant.

4-36

Table 4-23 Tritium release data from PWRs with Zircaloy-clad fuels (cont.)

Power Gaseous Tritium Released/Site Startup per (Ci/yr)b Reactora Date Unit (year) 1972 1973 1974 1975 1976 1977 (MWt)

R.E. Ginna 1520 1969 0.01 1.1 0.36 5.8 23.6 50 H.B. Robinson 2 2200 1970 1.0 4.0 52.0 193 158 61 Point Beach 1 & 2 1518 1970/72 8.0 25.0 43.0 177 395 194 Palisades 2530 1971 5.0 0.3 c c c 2.2 Maine Yankee 2440 1972 7.5 4.7 3.7 2.1 Indian Point 2 & 3 2758 1973/76 20.0 24.5 23.7 12.4 Surry 1 & 2 2441 1972/73 60.0 32 372 879 Turkey Point 3 & 4 2200 1972/73 9.2 3.5 5.2 3.9 Oconee 1, 2, & 3 2568 1973/74/72 0.75 1600 502 62.6 Zion 1 & 2 3250 1973/73 c c c Fort Calhoun 1420 1973 0.24 2.5 3.0 Prairie Island 1 & 2 1650 1973/74 10.1 33.1 88 Kewaunee 1650 1974 37.3 0.70 3.75 Three Mile Island 1 2535 1974 40.3 717 129 Rancho Seco 2772 1974 7.73 9.1 20.7 Arkansas 1 2568 1974 0.52 6.7 190 Calvert Cliffs 1 & 2 2700 1974/76 1.23 41 117 Cook 1 3250 1975 0.11 0.20 Millstone 2 2530 1975 21.3 47 Trojan 3411 1975 1.5 2.9 St. Lucie 1 2560 1975 320 Beaver Valley 1 2352 1976 213 Salem 1 338 1976 51 a These values are from Table 2-29 of NUREG-0017, Revision 1.

b Data from semiannual reports of reactors listed.

c No reported data.

d No radioactive liquid wastes were discharged from Unit 2 during the entire year. In 1975, no radioactive liquid wastes were discharged from Unit 1 during the last 6 months.

e Rancho Seco is designed to be a zero or very low liquid release plant.

4-37

Table 4-23 Tritium release data from PWRs with Zircaloy-clad fuels (cont.)

Power Liquid Tritium Released/Site Startup per (Ci/yr)b Reactora Date Unit (year) 1972 1973 1974 1975 1976 1977 (MWt)

R.E. Ginna 1520 1969 120 286 195 261 242 119 H.B. Robinson 2 2200 1970 410 431 475 624 980 685 Point Beach 1 & 2 1518 1970/72 560 556 832 886 694 1000 Palisades 2530 1971 210 185 8.3 41.3 9.6 56 Maine Yankee 2440 1972 219 177 368 153 Indian Point 2 & 3 2758 1973/76 48 366 332 371 Surry 1 & 2 2441 1972/73 246 442 782 408 Turkey Point 3 & 4 2200 1972/73 580 793 771 924 Oconee 1, 2, & 3 2568 1973/74/72 124 3550 2192 1918 Zion 1 & 2 3250 1973/73 39.4d 1.1d 727d Fort Calhoun 1420 1973 111 122 157 Prairie Island 1 & 2 1650 1973/74 763 1925 1349 Kewaunee 1650 1974 277 213 295 Three Mile Island 1 2535 1974 463 189 192 Rancho Seco 2772 1974 132 0.0e 0.90e Arkansas 1 2568 1974 460 212 245 Calvert Cliffs 1 & 2 2700 1974/76 263 274 575 Cook 1 3250 1975 192 285 Millstone 2 2530 1975 277 211 Trojan 3411 1975 36 311 St. Lucie 1 2560 1975 242 Beaver Valley 1 2352 1976 108 Salem 1 338 1976 296 a These values are from Table 2-29 of NUREG-0017, Revision 1.

b Data from semiannual reports of reactors listed.

c No reported data.

d No radioactive liquid wastes were discharged from Unit 2 during the entire year. In 1975, no radioactive liquid wastes were discharged from Unit 1 during the last 6 months.

e Rancho Seco is designed to be a zero or very low liquid release plant.

4-38

Table 4-23 Tritium release data from PWRs with Zircaloy-clad fuels (cont.)

Power Total Tritium Released/Unit Startup per (Ci/yr-MWt at 80-percent capacity)b Reactora Date Unit (year) 1972 1973 1974 1975 1976 1977 (MWt)

R.E. Ginna 1520 1969 0.11 0.19 0.20 0.19 0.27 0.11 H.B. Robinson 2 2200 1970 0.19 0.25 0.39 0.42 0.50 0.37 Point Beach 1 & 2 1518 1970/72 0.39 0.22 0.59 0.36 0.35 0.37 Palisades 2530 1971 0.26 0.20 0.02 Maine Yankee 2440 1972 0.14 0.09 0.13 0.07 Indian Point 2 & 3 2758 1973/76 0.04 0.17 0.19 0.08 Surry 1 & 2 2441 1972/73 0.11 0.11 0.32 0.30 Turkey Point 3 & 4 2200 1972/73 0.16 0.20 0.20 0.24 Oconee 1, 2, & 3 2568 1973/74/72 0.07 0.79 0.48 0.35 Zion 1 & 2 3250 1973/73 Fort Calhoun 1420 1973 0.12 0.12 0.12 Prairie Island 1 & 2 1650 1973/74 0.24 0.66 0.41 Kewaunee 1650 1974 0.20 0.14 0.19 Three Mile Island 1 2535 1974 0.20 0.46 0.13 Rancho Seco 2772 1974 0.24 0.01 0.01 Arkansas 1 2568 1974 0.21 0.13 0.19 Calvert Cliffs 1 & 2 2700 1974/76 0.13 0.11 0.16 Cook 1 3250 1975 0.06 0.13 Millstone 2 2530 1975 0.14 0.13 Trojan 3411 1975 0.04 0.10 St. Lucie 1 2560 1975 0.22 Beaver Valley 1 2352 1976 0.22 Salem 1 338 1976 0.36 a These values are from Table 2-29 of NUREG-0017, Revision 1.

b Data from semiannual reports of reactors listed.

c No reported data.

d No radioactive liquid wastes were discharged from Unit 2 during the entire year. In 1975, no radioactive liquid wastes were discharged from Unit 1 during the last 6 months.

e Rancho Seco is designed to be a zero or very low liquid release plant.

4-39

Table 4-24 Tritium release rate from operating PWRs Function of the Number of Years of Operation Reactora (Ci/yr-MWt at 80-percent capacity)b 1 2 3 4 5 6 7 R.E. Ginna 0.11 0.19 0.20 0.19 0.27 0.11 0.17 H.B. Robinson 2 0.19 0.25 0.39 0.42 0.50 0.37 Point Beach 1 & 2 0.39 0.22 0.59 0.36 0.35 0.37 0.51 Maine Yankee 0.14 0.09 0.13 0.07 0.18 Indian Point 2 & 3 0.04 0.17 0.19 0.08 Surry 1 & 2 0.11 0.11 0.32 0.30 Turkey Point 3 & 4 0.16 0.20 0.20 0.24 0.20 Oconee 1, 2, & 3 0.07 0.77 0.48 0.35 0.19 Fort Calhoun 0.12 0.12 0.12 0.13 Prairie Island 1 & 2 0.24 0.66 0.41 0.25 Kewaunee 0.20 0.14 0.19 0.20 Three Mile Island 1 0.20 0.46 0.13 0.17 Arkansas 1 0.21 0.13 0.19 Calvert Cliffs 1 & 2 0.13 0.11 0.16 Cook 0.06 0.13 0.31 Millstone 0.14 0.13 Trojan 0.04 0.10 St. Lucie 0.22 Beaver Valley 0.22 0.51 Salem 0.36 0.41 Average 0.16 0.29 0.30 0.25 0.26 0.31 0.40 a These values are from Table 2-30 of NUREG-0017, Revision 1.

b Data from semiannual reports of reactors listed.

4-40

Table 4-25 Total tritium release data from operating PWRs (percent)

Percent of Total Tritium Released in Liquid Effluents Reactora 1972 1973 1974 1975 1976 1977 R.E. Ginna 100.0 99.6 99.8 97.8 91.9 70.4 H.B. Robinson 2 99.8 99.1 90.1 76.4 86.1 91.8 Point Beach 1 & 2 98.6 95.7 95.1 83.3 63.7 93.8 Palisades 97.7 99.8 b b b 96.2 Maine Yankee 96.7 97.4 99.0 98.6 Indian Point 2 & 3 70.6 93.7 93.3 96.8 Surry 1 & 2 80.4 93.2 67.8 31.7 Turkey Point 3 & 4 98.4 99.6 99.3 96.8 Oconee 1, 2, & 3 99.4 68.9 81.4 96.8 Zion 1 & 2 b b b Fort Calhoun 97.9 98.0 98.1 Prairie Island 1 & 2 98.7 98.3 93.9 Kewaunee 88.1 99.7 98.7 Three Mile Island 1 92.0 20.9 59.8 Rancho Seco 94.5 0.0c 0.43c Arkansas 1 99.9 96.9 56.3 Calvert Cliffs 1 & 2 99.5 87.0 83.1 Cook 1 100.0 100.0 Millstone 2 92.9 81.8 Trojan 96.0 99.1 St. Lucie 1 43.1 Beaver Valley 1 33.6 Salem 1 85.3 Weighted Averaged 99.2 98.0 91.1 89.5 83.5 78.5 a These values are from Table 2-31 of NUREG-0017, Revision 1.

b No reported data.

c Rancho Seco is designed to be a zero or very low liquid release plant.

d Average weighted by nuclear thermal output per unit.

4-41

Table 4-26 Distribution of tritium release gaseous effluents Source of Gaseous Tritium Release (Percent of Total)b Planta Auxiliary Refueling Containment Building Area Building R.E. Ginna 31 69 NM Calvert Cliffs 1 & 2 38 46 16 Three Mile Island 1 5 43 52 Zion 1 & 2 79 WA 21 Turkey Point 3 & 4 75 17 8 Rancho Seco 92 WA 8 Prairie Island 1 & 2 7.2 91.8 1.0 Average 32 50 18 a These values are from Table 2-32 of NUREG-0017, Revision 1.

b The following method is used to determine the tritium release in this table:

- Containment building operation average percent total release (16 + 52 + 21 + 8 + 8 +1)/(6) = 17.7 percent = 18 percent.

- Refueling area operation average percent total release The refueling area for R.E. Ginna is reduced by 18 percent >> (69 - 18) = 51 percent. Therefore, the operation average percent total release is (51 + 46 + 43 + 17 + 91.8)/(5) = 49.8 percent = 50 percent.

- Auxiliary building operation average percent total release The auxiliary area for Zion is reduced by 50 percent >> (79 - 50) = 29 percent and so is Rancho Seco (92 - 50) = 42 percent. Therefore, the operation average percent total release is (31 + 38 + 5 + 29 + 75 +

42 + 7.2)/(7) = 32.4 percent = 32 percent.

NM = Not measured.

WA = Releases from refueling are combined with the auxiliary building release.

Decontamination Factors for Demineralizers Parameter Table 4-27 lists the DFs used by the code for the various demineralizers when the user selects the default (GALE86) fixed modeling parameters. The following operating conditions were considered for the evaluation of demineralizer performance:

The DF is dependent on the inlet radioactivity, ion concentrations, and bed volume ion exchange capacity. For demineralizer performance within the same range of controlled operating conditions, the DF increases with inlet radioactivity concentration and decreases with bed volume throughout.

4-42

When two demineralizers are used in series, the first demineralizer will have a higher DF than the second. However, the data in NUREG/CR-0143, The Use of Ion Exchange to Treat Radioactive Liquids in Light-Water-Cooled Nuclear Power Plants, issued August 1978 [Ref. 25], indicate that cesium and rubidium will be more strongly exchanged in the second demineralizer in series than in the first, as the concentration of preferentially exchanged competing nuclides is reduced.

As indicated in NUREG/CR-0143, compounds of yttrium, molybdenum, and technetium form colloidal particles that tend to plate out on solid surfaces. Mechanisms such as plateout on the relatively large surface areas provided by demineralizer resin beds result in removal of these nuclides to the degree stated above. An analysis of effluent release data indicates that these nuclides, although present in the primary coolant, are not found in the effluent streams.

Table 4-27 Decontamination factors for demineralizers cesium & Other Demineralizera Anionb Rubidiumb Radionuclidesb Mixed bed purification system (LiBO3) 1.0E+01 2.0E+00 5.0E+01 Boron recycle system 1.0E+01 2.0E+00 1.0E+01 Evaporator condensate (H+, OH-) 5.0E+00 1.0E+00 1.0E+01 1.0E+02 2.0E+00 1.0E+02 Radwaste (H+, OH-)

(1.0E+01) (1.0E+01) (1.0E+01) 1.0E+02 1.0E+01 1.0E+02 Steam generator blowdown (1.0E+01) (1.0E+01) (1.0E+01) 1.0E+00 1.0E+01 1.0E+01 Cation bed (H+) (any system)

(1.0E+00) (1.0E+01) (1.0E+01) 1.0E+02 1.0E+00 1.0E+00 Anion bed (OH-) (any system)

(1.0E+01) (1.0E+00) (1.0E+00) 1.0E+01 2.0E+01 1.0E+01 Powdex (any system)

(1.0E+01) (1.0E+01) (1.0E+01) a These values are from Section 2.2.18.1 of NUREG-0017, Revision 1.

b For two demineralizers in series, the DF for the second demineralizer is given in parentheses.

Basis The DFs for purification, radwaste, and evaporator condensate demineralizers are based on (1) source term measurements made at Fort Calhoun, Zion, Turkey Point, Prairie Island, and Rancho Seco stations by the in-plant source term measurement program (References 16 through 21), (2) the findings of a generic review in the nuclear industry by Oak Ridge National Laboratory (ORNL) (Reference 25), and (3) measurements taken at Three Mile Island 1 (Reference 3). The DFs for the remaining demineralizers are based on ORNLs findings.

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The ORNL generic review contains operating and theoretical data that provide a basis for the numerical values assigned. The ORNL data were projected to obtain a performance value expected over an extended period of operation. It is assumed that attempts to extend the service life of the resin will reduce the DFs below those expected under controlled operating conditions.

Average DFs for the Fort Calhoun, Zion, Turkey Point, Rancho Seco, and Prairie Island stations were obtained by dividing the average inlet radionuclide concentration of samples by that of the average outlet concentration for each nuclide.

The DF used for the parameter was considered to be representative of the data in References 16 through 21.

Decontamination Factors for Evaporators Parameter Table 4-28 lists the DFs for evaporators used by the code when the user selects the default (GALE86) fixed modeling parameters.

Table 4-28 Decontamination factors for evaporators All Nuclides Evaporatora Iodine Except Iodine Miscellaneous radwaste evaporators 1.0E+03 1.0E+02 Boric acid evaporators 1.0E+03 1.0E+02 Separate evaporator for detergent wastes 1.0E+02 1.0E+02 a These values are from Section 2.2.19.1 of NUREG-0017, Revision 1.

Basis The DFs for evaporators are based on (1) source term measurements made at the Fort Calhoun, Zion, Turkey Point, Prairie Island, and Rancho Seco stations by the in-plant source term measurement program (References 16 through 21) and (2) the findings of a generic review in the nuclear industry by ORNL (NUREG/CR-0142, The Use of Evaporation to Treat Radioactive Liquids in Light-Water-Cooled Nuclear Power Plants, issued September 1978 [Ref.

26]).

Average DFs for Zion, Fort Calhoun, Turkey Point, Rancho Seco, and Prairie Island were obtained by dividing the average inlet radioactivity of samples by the average outlet radioactivity of samples for each radionuclide.

The DF used for the parameter was that considered to be the most representative of the data given in References 16 through 21.

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Decontamination Factors for Liquid Radwaste Filters Parameter A DF of 1.0E+00 for liquid radwaste filters is assigned for all radionuclides.

Basis NUREG/CR-0141, The Use of Filtration to Treat Radioactive Liquids in Light-Water-Cooled Nuclear Power Plants, issued September 1978 [Ref. 27], presents the findings of a generic review by ORNL of liquid radwaste filters used in the nuclear industry. Because of the various filter types and filter media employed, reported values of DFs vary widely, with no discernible trend. The principal conclusion reached in the ORNL report is that no credit should be assigned to liquid radwaste filters (DF of 1) until a larger database is obtained.

Additional data from Fort Calhoun, Zion, Turkey Point, Rancho Seco, and Prairie Island stations (References 16 through 21) indicate that DFs in liquid radwaste filters vary widely from less than 1 to greater than 50 (with a mean value of 1.3). Therefore, a DF of 1 for liquid radwaste filters is used.

Decontamination Factors for Reverse Osmosis Parameter An overall DF of 3.0E+01 for laundry wastes and a DF of 1.0E+01 for other liquid radwastes are used.

Basis Reverse osmosis processes are generally run as semibatch processes. The concentrated stream rejected by the membrane is recycled until a desired fraction of the batch is processed through the membrane. The ratio of the volume processed through the membrane to the inlet batch volume is the percent recovery. The DF normally specified for the process is the ratio of nuclide concentrations in the concentrated liquid stream to the concentrations in the effluent stream. This ratio is termed the membrane DF (DFm). For source term calculations, the system DF (DFs) should be used. The DFs is the ratio of the nuclide concentrations in the feed stream to those in the effluent stream. The relationship between the DFs and the DFm is nonlinear and is a function of the percent recovery. This relationship can be expressed as follows in Equation (4-14):

F DFs = 1 (4-14) 1 (1 F) DFm where DFm = the membrane DF; DFs = the system DF; and F = the ratio of effluent volume to inlet volume (fractional recovery).

Tables 4-29 through 4-32 give membrane DFs derived from operating data at Point Beach, R.E.

Ginna, and the H.B. Robinson Steam Electric Plant (NUREG/CR-0724, A Study of Reverse Osmosis Applicability to Light Water Reactor Radwaste Processing, issued November 1978

[Ref. 28) and laboratory data on simulated radwaste liquids (Reference 3). These data indicate that the overall membrane DF is approximately 1.0E+02. The percent recovery for liquid 4-45

radwaste processes using reverse osmosis is expected to be approximately 95 percent (i.e., 5-percent concentrated liquid). Using these values in Equation (4-14), the system DF is approximately 3.0E+01, as shown in Equation (4-15):

0.95 DFs = 1 = 3.0E+01 (4-15) 1 (1 0.95) 100 The data used were derived mainly from tests on laundry wastes. The DF for other plant wastes (e.g., floor drain wastes) is expected to be lower because of the higher concentrations of iodine and cesium isotopes. As indicated by the data in Tables 4-29 through 4-32, the membrane DF for these isotopes is lower than the average membrane DF used in the evaluation for laundry waste.

Table 4-29 Reverse osmosis decontamination factors for R.E. Ginna Concentrate Product Membrane Nuclidea Activity Activity DF (Ci/cm3) (Ci/cm3)

Ce-144 2.68E-04 <2.2E-07 1.2E+03 Co-58 8.55E-05 <3.4E-08 2.5E+03 Ru-103 5.83E-05 <5.5E-08 1.1E+03 Cs-137 4.09E-04 6.6E-06 6.0E+01 Cs-134 2.02E-04 3.2E-06 6.0E+01 Nb-95 5.35E-05 <5.3E-08 1.0E+03 Zr-95 2.36E-05 <3.7E-08 6.4E+02 Mn-54 8.82E-05 <3.4E-08 2.6E+03 Co-60 9.63E-04 <8.1E-08 1.2E+04 Total Isotopic 2.15E-03 9.8E-06 2.19E+03 Gross Beta 1.63E-03 1.86E-05 8.8E+01 Total 3.78E-03 2.84E-05 Average 1.33E+02 a These values are from Table 2-33 of NUREG-0017, Revision 1.

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Table 4-30 Reverse osmosis decontamination factors for Point Beach Concentrate Product Membrane Datea Time Activity Activity DF (Ci/cm3) (Ci/cm3) 0840 1.1E-05 6.8E-07 1.6E+01 June 14, 1971 1225 6.3E-05 4.2E-07 1.5E+02 1350 6.8E-05 3.2E-07 2.1E+02 1030 2.7E-04 3.1E-06 8.7E+01 1315 1.0E-04 1.7E-06 5.9E+01 June 15, 1971 1440 1.3E-04 1.1E-07 1.2E+03 1510 1.6E-04 1.1E-07 1.5E+03 1530 1.8E-04 5.7E-07 3.16E+02 Total 9.8E-04 7.0E-06 Average 1.4E+02 a These values are from Table 2-34 of NUREG-0017, Revision 1.

Table 4-31 Reverse osmosis decontamination factors for H.B. Robinson 2 Co-60a Co-58a I-131a 2.6E+02 2.9E+01 5.0E+01 3.82E+02 --- 2.0E+01 4.36E+02 --- 3.9E+01 1.07E+02 2.29E+02 2.6E+01 7.6E+01 4.9E+02 9.6E+01 9.4E+01 1.31E+02 1.1E+01 Average 2.27E+02 2.2E+02 3.4E+01 a These values are from Table 2-35 of NUREG-0017, Revision 1.

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Table 4-32 Expected reverse osmosis decontamination factors for specific nuclides Concentrate Product Membrane Nuclidea Activity Activity DF (Ci/cm3) (Ci/cm3)

Co-60 2.5E-04 5.0E-07 5.0E+02 Mo-99 3.8E-02 1.0E-03 4.0E+01 I-131, I-132, I-133, I-134, & I-135 1.2E-01 4.0E-03 3.0E+01 Cs-134 & Cs-137 4.3E-02 2.0E-04 2.0E+02 Total 2.0E-01 5.2E-03 ---

Average 4.0+01 a These values are from Table 2-36 of NUREG-0017, Revision 1.

Guideline for Calculating Liquid Waste Holdup Times The holdup times to permit radioactive decay applied to the input waste streams are calculated using the following parameters:

(1) The collection time should be calculated for an 80-percent volume change in the tank, based on the liquid waste flow rates from the inlet sources.

(2) The process time is the total time liquid remains in the system for processing, based on the flow rate through the limiting process step.

(3) The discharge time is one-half the time required to empty the final liquid waste sample (test) tank to the environment. This value is based on the maximum rate of the discharge pumps and the nominal tank volume.

The calculated values in step 1 above and the total of steps 2 and 3 are used as inputs to the computer code.

Adjustment to Liquid Radwaste Source Terms for Anticipated Operational Occurrences Parameter

Increase the calculated source term by 1.6E-01 Ci/yr per reactor, using the same isotopic distribution as for the calculated source term to account for AOOs, such as operator errors that result in unplanned releases.

Assume evaporators to be unavailable for 2 consecutive days per week for maintenance. If a 2-day holdup capacity exists in the system (including surge tanks) or an alternative evaporator is available, no adjustment is needed. If less than a 2-day capacity is available, assume the waste excess is handled as follows:

- High-purity or low-purity wasteProcessed through an alternative system (if available) using a discharge fraction consistent with the lower purity system.

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- Chemical wasteDischarged to the environment to the extent holdup capacity or an alternative evaporator is available.

The following methods should be used for calculating holdup times and effective system DF:

- Holdup capacityIf two or more holdup tanks are available, assume one tank is full (80-percent capacity) with the remaining tanks empty at the start of the 2-day outage. If there is only one holdup tank, assume that it is 40 percent full at the start of the 2-day outage with a usable capacity of 80 percent.

- Effective system DFIf the reserve storage capacity is inadequate for waste holdup over a 2-day evaporator outage, and if an alternate evaporator is unavailable to process the wastes from the out-of-service evaporator, the subsystem DF should be adjusted to show the effect of the evaporator outage.

For example, a DF of 1.0E+05 was calculated for a radwaste demineralizer (1.0E+02 from Table 4-27) and radwaste evaporator (1.0E+03 from Table 4-28) in series. If an adjustment were required for the evaporator being out of service 2 days per week, with only a 1-day holdup tank capacity, then the effective system DF can be calculated as follows:

For 6 days (7 - 2 + 1) out of 7, the system DF would be 1.0E+05.

For the remaining 1 day, the system DF would be 1.0E+02 (only the demineralizer DF is considered). The effective DF is calculated as shown in Equation (4-16):

6 1 1 DF = (1.0E-05) + (1.0E-02) = 7.0+02 (4-16) 7 7 Basis Reactor operating data over an 8-year period (January 1970 through December 1977),

representing 154 reactor-years of operation, were evaluated to determine the frequency and extent of unplanned liquid releases. During the period evaluated, 62 unplanned liquid releases occurred: 23 were caused by operator errors, 26 by component failures, 5 by inadequate procedures or failure to follow procedures, and the remaining 8 resulted from miscellaneous causes such as design errors. Table 4-33 summarizes the findings of this evaluation. Based on the data in Table 4-33, it is estimated that 1.6E-01 Ci/reactor-year will be discharged in unplanned releases in liquid effluents.

The availability of evaporators in waste treatment systems is expected to be in the range of 60 to 80 percent. Unavailability is attributed to scaling, fouling of surfaces, instrumentation failures, corrosion, and occasional upsets resulting in high carryovers requiring system cleaning.

A value of 2 consecutive days of unavailability per week was chosen as being representative of operating experience. For systems having sufficient tank capacity to collect and hold wastes during the assumed 2-day per week outage, no adjustments are required for the source term. If less capacity is available, the difference between the waste expected during 2 days of normal operation and the available holdup capacity is assumed to follow an alternative route for processing. Because processing through an alternative route implies the mixing of wastes having different purities and different dispositions after treatment, it is assumed that the fraction 4-49

of waste discharged following processing will be that normally assumed for the less pure of the two waste streams combined.

Since chemical and regenerant wastes are not amenable to processes other than evaporation, it is assumed that unless an alternative evaporation route is available, chemical and regenerant wastes in excess of the storage capacity are discharged without treatment.

Table 4-33 Frequency and extent of unplanned liquid radwaste releases from operating plants Unplanned Liquid Releasesa Total number (unplanned releases) 6.2E+01 Fraction caused by personnel error 3.7E-01 Fraction caused by component failure 4.2E-01 Fraction caused by inadequate procedures or failure to follow procedures 8.0E-02 Fraction attributable to other causes 1.3E-01 Approximate activity (Ci) 2.4E+1 Fraction of cumulative occurrence per reactor-year (plants reporting releases >5 gal of 1.6E-01 liquid waste/reactor-year)

Fraction of cumulative occurrences per reactor-year (plants reporting activity released 2.5E-01

>0.01 Ci/reactor-year)

Activity per release (Ci/release) 3.9E-01 Activity released per reactor-year (Ci/reactor-year) 1.6E-01 Volume of release per reactor-year (gal/reactor-year) 6.33E+02 a The data in this table are taken from Table 2-37 of NUREG-0017, Revision 1. Values in this table are based on reported values in 1970-1977 licensee event reports representing 1.54E+02 reactor-years of operation.

Atmospheric Steam Dump Parameter Noble gases and radioiodines released to the atmosphere from the steam dumps because of turbine trips and low-power physics tests will have a negligible effect on the calculated gaseous source term.

Basis The evaluation considered the quantity of noble gases and radioiodine released to the atmosphere from steam dumps because of low-power physics testing and turbine trips from full power. The evaluation indicates that the iodine-131 and noble gas releases will be less than 1 percent of the turbine building gaseous source term.

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The evaluation of releases following a turbine trip from full power is based on the following parameters:

an average of two turbine trips annually

40-percent turbine bypass capacity to the main condenser

2-second rod insertion time required to scram the reactor following a turbine trip

12-second cycle time to recirculate one primary coolant volume through the reactor and steam generator The above parameters are based on a 3,400-MWt Reference Safety Analysis Report-3 (RESAR-3) reactor plant. Using these parameters, it is postulated that steam will continue to be produced at a full-power rate during the time the control rods are inserted and during the time required to recirculate one primary coolant volume. After this time, the turbine bypass will be adequate to handle steam generated from decay heat. The quantity of steam released is calculated as shown in Equation (4-17):

= (1.5E+07 lbh)(6.0E-01)(1.4E+01 s) (2 tripsyr)(4.54E+02 gallb) h3.6E+03 s (4-17)

= 3.2E+07 gal-steam/yr The iodine-131 concentration in the main steam for a U-tube steam generator is approximately 1.8E-08 Ci/g-steam from Table 4-37.

Based on the steam release calculated above, the associated iodine-131 release is approximately 6.0E-07 Ci/yr as calculated in Equation (4-18):

gal steam Ci I 131yr = 3.2E+07 yr 1.8E-08 gal steam 1.0E+06 CiCi (4-18) iodine-131/yr = 5.8E-07 Ci/yr Releases resulting from low-power physics testing are calculated based on one 10-hour release of steam each year following a refueling. For a RESAR-3 reactor, low-power physics testing is conducted at 5-percent power. The conditions given above for power level and steady-state main steam iodine-131 activity are used. In addition, it is assumed that the reactor will be shut down for 30 days for refueling before low-power physics testing. The iodine-131 releases are calculated to be approximately 4.6E-06 Ci/yr using Equation (4-19):

Ci

= (1.5E+07 lbh)(5.0E-02)(4.54E+02 gallb)(1.0E + 02 hyr) 1.8E-08 gal-steam (6.93E-01)(3.0E+01 day) Ci exp (8.05E+00 day) 1.0E+06 Ci (4-19) iodine-131/yr = 4.6E-06 Ci/yr 4-51

Carbon-14 Releases Parameter The annual quantity of carbon-14 released from a PWR is 7.3E+00 Ci/yr. It is assumed that most of the carbon-14 will form volatile compounds that will be released from the waste gas processing system and from the containment and auxiliary building atmospheres to the environment.

Basis The annual release of 7.3E+00 Ci of carbon-14 is based on measurements at 10 operating PWRs presented in Table 4-34. NUREG-0017, Revision 1, states that the carbon-14 reacts to form volatile compounds (principally CH, C2H6, and CO2) that are collected in the waste gas processing system through degassing of the primary coolant and released to the environment via the plant vent. Data also indicate that carbon-14 is released from the containment and auxiliary building vent as a result of leakage of primary coolant into the containment and auxiliary building atmospheres (References 16 through 21).

As shown in Table 4-35, an average of measurements, made at Turkey Point 3 and 4, Zion 1 and 2, Fort Calhoun, Prairie Island 1 and 2, and Rancho Seco, indicates that the release of carbon-14 breaks down to 22.6 percent from the containment building, 61.0 percent from the auxiliary building vents, and 16.4 percent from the waste gas processing system. Therefore, on this basis, it is assumed that 1.6E+00 Ci/yr of carbon-14 is released from the containment building, 4.5E+00 Ci/yr of carbon-14 is released from the auxiliary building vents, and 1.2E+00 Ci/yr of carbon-14 is released from the waste gas processing system.

Argon-41 Releases Parameter The annual quantity of argon-41 released from a PWR is 3.4E+01 Ci/yr. The argon-41 is released to the environment via the containment vent when the containment is vented or purged.

Basis Argon-41 is formed by neutron activation of stable naturally occurring argon-40 in the containment air surrounding the reactor vessel. The argon-41 is released to the environment when the containment is vented or purged. Table 2-40 of NUREG-0017, Revision 1, summarizes the available data and gaseous argon-41 releases from operating PWRs. The information reported by the licensees is not sufficiently detailed to correlate reported argon-41 releases with plant size and plant operating parameters. However, the average argon-41 release is estimated to be 3.4E+01 Ci/yr.

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Table 4-34 Carbon-14 release data from operating PWRs 1975 1976 1977 1978 Annual Average Planta (Ci/yr) (Ci/yr) (Ci/yr) (Ci/yr) Ci/yr - unit Connecticut Yankee 4.4E+01 4.0E+01 3.0E+01 7.0E+01 4.6E+01 Yankee Rowe 1.6E+00 1.3E-01 2.4E-01 3.3E-01 5.8E-01 Annual Average Plantb Area Ci/yr - unit Auxiliary Building 2.4E+00 Containment Building 7.5E-02 Turkey Point 3 & 4 Waste Gas Processing System (WGPS) 8.2E-02 Spent Fuel Area 3.8E-01 Total 3.7E+00 Fuel Pool and Auxiliary Building 3.0E-01 WGPS 8.1E-01 Fort Calhoun Containment 7.8E-01 Total 1.9E+00 Containment 1.8E+00 Fuel Pool and Auxiliary Building 1.4E+00 Zion 1 & 2 WGPS 6.2E-02 Total 3.3E+00 Containment Building 1.6E-02 Fuel Pool and Auxiliary Building 3.3E+00 Prairie Island 1 & 2 WGPS 2.5E-01 Total 3.6E+00 Containment Building 9.0E-01 Fuel Pool and Auxiliary Building 1.85E+00 Rancho Seco WGPS 8.5E-01 Total 3.6E+00 Average 7.3E+00 a These values are from Table 2-38 of NUREG-0017, Revision 1, and are based on semiannual release reports.

b These values are from Table 2-38 of NUREG-0017, Revision 1, and are based on in-plant source term measurements.

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Table 4-35 Distribution of carbon-14 released in gaseous effluents Plant Area Planta Auxiliary Building Containment WGPS and Fuel Handling (Percent) (Percent)

(Percent)

Turkey Point 3 & 4 2 75 23 Fort Calhoun 41 16 43 Zion 1 & 2 55 43 2 Rancho Seco 25 51 24 Prairie Island 1 & 2 0.5 92.5 7 Average: 22.6 61.0 16.4 a These values are from Table 2-39 of NUREG-0017, Revision 1.

Table 4-36 Summary of argon-41 releases from operating PWRs Release Reactor Namea Year (Ci/yr - reactor) 1974 8.5E-01 1975 9.3E-01 Yankee Rowe 1976 3.0E-01 1977 4.9E-01 1978 (1/2 year) 4.7E-01 1973 4.4E-02 Connecticut Yankee 1977 8.0E-02 1978 (1/2 year) 4.1E-02 1975 5.8E+00 R.E. Ginna 1976 1.9E-01 1973 1.76E+01 1974 1.6E+01 1975 2.08E+02 Point Beach 1 & 2 1976 3.1E+01 1977 9.2E+00 1978 (1/2 year) 1.33E+01 a These values are from Table 2-40 of NUREG-0017, Revision 1, and all data provided by the semiannual effluent release reports and the annual operating reports for each PWR listed.

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Table 4-36 Summary of argon-41 releases from operating PWRs (cont.)

Release Reactor Namea Year (Ci/yr - reactor) 1975 (1/2 year) 1.62E+01 1976 1.54E+01 H.B. Robinson 2 1977 2.31E+01 1978 (1/2 year) 4.62E+01 1974 (1/2 year) 1.5E+01 1975 3.2E-01 Surry 1976 9.15E+00 1977 (1/2 year) 1.65E+01 D.C. Cook 1978 (1/2 year) 1.97E+01 1974 2.6E+01 1975 5.13E+01 Turkey Point 3 & 4 1976 3.94E+01 1977 4.5E+01 1974 (1/2 year) 5.95E+01 1975 4.2E+01 Oconee 1, 2, & 3 1976 1.18E+02 1977 8.1E+00 1978 (1/2 year) 1.99E+01 1975 8.2E+00 1976 2.2E+00 Fort Calhoun 1977 2.3E+00 1978 (1/2 year) 2.7E-01 Palisades 1978 (1/2 year) 1.0E-02 Zion 1 & 2 1978 (1/2 year) 2.48E+01 1975 1.3E+00 1976 2.1E+01 Prairie Island 1 & 2 1977 3.18E+01 1978 1.35E+01 a These values are from Table 2-40 of NUREG-0017, Revision 1, and all data provided by the semiannual effluent release reports and the annual operating reports for each PWR listed.

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Table 4-36 Summary of argon-41 releases from operating PWRs (cont.)

Release Reactor Namea Year (Ci/yr - reactor) 1975 1.3E+00 1976 2.1E+01 Prairie Island 1 & 2 1977 3.18E+01 1978 1.35E+01 1976 (1/2 year) 3.0E+01 Kewaunee 1978 (1/2 year) 5.9E+00 1975 (1/2 year) 5.0E+01 1976 1.2E+01 Three Mile Island 1 1977 6.6E+01 1978 (1/2 year) 4.65E+01 1976 (1/2 year) 2.0E+00 Calvert Cliffs 1977 (1/2 year) 3.1E+00 1977 9.8E+00 Rancho Seco 1978 (1/2 year) 1.8E+00 a These values are from Table 2-40 of NUREG-0017, Revision 1, and all data provided by the semiannual effluent release reports and the annual operating reports for each PWR listed.

Guidelines for Rounding off Numerical Values The estimated annual release of radioactive materials in liquid and gaseous wastes is deemed at best to be no more accurate than two significant figures. The revised output listings file shows all estimated release rates in exponential form with two significant figures.

Supplementary listings provide additional precision for use in special applications such as sensitivity studies.

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4.2 ANS-18.1 Version (Source Term Options)

Regulatory source terms are deeply embedded in the regulatory policy and practices of the NRC where the current licensing process has evolved over the past 50 years. It is based upon the concept of defense-in-depth in which power plant design, operation, siting, and emergency planning are the cornerstones of safety. This approach encourages nuclear plant designers to incorporate several lines of defense in order to maintain the effectiveness of physical barriers between radiation sources and materials from workers, members of the public and environment in operational states and, for some barriers, in accident conditions. The various regulatory source terms are used, in part, to establish and confirm the design basis of the nuclear facility and items important to safety; ensuring that the plant design meets the safety and numerical radiological criteria set forth in regulation and subsequent guidance.

Specific examples of the various regulatory source terms can be found in NUREG-0800 which provides information on the NRC staffs regulatory guides. In the past, the normal operating source terms have been calculated using the techniques described in the various versions of the ANSI/ANS-18.1 standard. The purpose of this standard is to provide a set of typical radionuclide concentrations for estimating the radioactivity in the principal fluid streams of an LWR. The referenced coolant concentrations are based on collected data and calculations reported by various sources which include the NRC, EPRI, and utilities. The values in this standard are considered as representative coolant concentrations in an LWR over its lifetime based upon the currently available data. It is important for current operating reactor licensees and new reactor applicants to apply a best estimate reactor coolant radiological source term for their design of plant-specific purification and radioactive waste systems and to estimate the expected release of radioactivity via various effluent streams.

Each subsequent update to the ANSI/ANS-18.1 standard reflects improvements in industry fuel performance, burnup, power uprates, treatment technologies and practices, and reduction of occurrences and severity of fuel defects. Therefore, the selection of an ANSI/ANS-18.1 version (1984, 1999, or 2016) should reflect the current design and operating conditions of the given facility/design. This is so because it has been demonstrated that previous versions of ANSI/ANS-18.1 could be non-conservative for some dosimetrically important radionuclides in facilities which have undergone significant power uprates (125 percent plus rated thermal power) and received approval to burn their cores much hotter (e.g., from 35 GWd/MTU up to 55-62 GWd/MTU) than when originally licensed. For example, in reviewing the offsite dose analyses for the various power uprate license amendment requests, a historical assumption has been that radionuclide inventories increase linearly with reactor power. The two factors mentioned above, among others, challenge this traditional assumption. The ANSI/ANS-18.1-2016 source term is, in general, lower than the ANSI/ANS-18.1-1999 source term, but there are 7 radionuclides with higher reactor water concentrations in ANSI/ANS-18.1-2016. For instance, the Xe-133 concentration, which is a major external dose contributor, has increased by 28 percent. This increase is due, in part, to higher fuel higher burnup. Likewise, the Cs-134 concentration, which is a typical long-term internal dose contributor, has increased by 1,386 percent. Therefore, use of the ANSI/ANS-18.1-2016 standard may not be appropriate for a SMR designed to run at low reactor fuel burnup levels. In fact, it may be best for these reactor designs to apply the previous standards which were developed at a time before many of the LWRs were approved for power uprates and increased reactor fuel burnup levels.

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In GALE-PWR 3.2, the user can select from three options for the reactor coolant source term, radionuclide concentrations in the primary and secondary coolant, as shown in Figure 3-4.

Section 4.2.1 describes the reactor coolant source term values for the default option, ANS-18.1 Version-1999, which corresponds to the values in ANSI/ANS-18.1-1999 and is consistent with the guidance in RG 1.112, Revision 1 and NUREG-0800. Section 4.2.2 describes the reactor coolant source term values for the ANS-18.1 Version-1984 option, which corresponds to the values in ANSI/ANS-18.1-1984 and is consistent with the guidance in DC/COL-ISG-05. Section 4.2.3 describes the reactor coolant source term values for the ANS 18.1 Version-2016 option, which corresponds to the values in ANSI/ANS-18.1-2016.

4.2.1 ANSI/ANS-18.1-1999 Source Term Parameters Parameter As used in the GALE-PWR 3.2 code, the default values for the reactor coolant source term, ANS-18.1 version, are from ANSI/ANS-18.1-1999. Tables 4-37 and 4-38 list the expected radionuclide concentrations in the reactor coolant and steam for PWRs with design parameters within the ranges listed in Tables 4-39 and 4-40. If any design parameters are outside the range in Tables 4-39 and 4-40, the code adjusts the concentrations in Tables 4-37 and 4-38, using the factors in Tables 4-41, 4-42, and 4-43. Figures 4-4 and 4-5 show the relationship of the design parameters.

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Table 4-37 Radionuclide concentrations in PWRs with U-tube steam generators -

ANSI/ANS-18.1-1999 Secondary Coolantc Reactor Coolantb Radionuclidea Waterd Steame (Ci/g)

(Ci/g) (Ci/g)

Noble Gases Kr-85m 1.6E-02 3.4E-09 Kr-85 4.3E-01 8.9E-08 Kr-87 1.7E-02 3.0E-08 Kr-88 1.8E-02 3.8E-09 Xe-131m 7.3E-01 1.5E-07 Xe-133m 7.0E-02 1.5E-08 Xe-133 2.9E-02 6.0E-09 Xe-135m 1.3E-01 2.7E-08 Xe-135 6.7E-02 1.4E-08 Xe-137 3.4E-02 7.1E-09 Xe-138 6.1E-02 1.3E-08 Halogens Br-84 1.6E-02 7.5E-08 7.5E-10 I-131 2.0E-03 8.1E-08 8.1E-10 I-132 6.0E-02 8.9E-07 8.9E-09 I-133 2.6E-02 9.0E-07 9.0E-09 I-134 1.0E-01 7.2E-07 7.2E-09 I-135 5.5E-02 1.4E-06 1.4E-08 Cesium & Rubidium Rb-88 1.9E-01 5.3E-07 2.6E-09 Cs-134 3.7E-05 1.7E-09 9.0E-12 Cs-136 8.7E-04 4.0E-08 2.0E-10 Cs-137, Ba-137mf 5.3E-05 2.5E-09 1.2E-11 Water Activation Products N-16 4.0E+01 1.0E-06 1.0E-07 Tritium H-3 1.0E+00 1.0E-03 1.0E-03 Other Radionuclides Na-24 4.7E-02 1.5E-06 7.5E-09 Cr-51 3.1E-03 1.3E-07 6.3E-10 a These concentrations are from Table 6 of ANSI/ANS-18.1-1999.

b The concentrations given are for reactor coolant entering the letdown line. These concentrations are obtained from ANSI/ANS-18.1-1999.

c The values are based on a primary-to-secondary leak of 7.5E+01 lb/d.

d The concentrations given are for water in a steam generator.

e The concentrations given are for steam leaving a steam generator.

f These nuclides are in secular equilibrium; other radionuclide concentrations are those of the parent.

4-59

Table 4-37 Radionuclide concentrations in PWRs with U-tube steam generators -

ANSI/ANS-18.1-1999 (cont.)

Secondary Coolantc Reactor Coolantb Radionuclidea Waterd Steame (Ci/g)

(Ci/g) (Ci/g)

Mn-54 1.6E-03 6.5E-08 3.3E-10 Fe-55 1.2E-03 4.9E-08 2.5E-10 Fe-59 3.0E-04 1.2E-08 6.1E-11 Co-58 4.6E-03 1.9E-07 9.4E-10 Co-60 5.3E-04 2.2E-08 1.1E-10 Zn-65 5.1E-04 2.1E-08 1.0E-10 Sr-89 1.4E-04 5.7E-09 2.9E-11 Sr-90 1.2E-05 4.9E-10 2.4E-12 Sr-91 9.6E-04 2.8E-08 1.4E-10 Y-91m 4.6E-04 3.2E-09 1.6E-11 Y-91 5.2E-06 2.1E-10 1.1E-12 Y-93 4.2E-03 1.2E-07 6.1E-10 Zr-95 3.9E-04 1.6E-08 7.9E-11 Nb-95 2.8E-04 1.1E-08 5.7E-11 Mo-99 6.4E-03 2.5E-07 1.2E-09 Tc-99m 4.7E-03 1.1E-07 5.7E-10 Ru-103 7.5E-03 3.1E-07 1.6E-09 Ru-106 9.0E-02 3.7E-06 1.8E-08 Ag-110m 1.3E-03 5.3E-08 2.7E-10 Te-129m 1.9E-04 7.8E-09 3.9E-11 Te-129 2.4E-02 2.2E-07 1.1E-09 Te-131m 1.5E-03 5.4E-08 2.7E-10 Te-131 7.7E-03 2.9E-08 1.5E-10 Te-132 1.7E-03 6.6E-08 3.3E-10 Ba-140 1.3E-02 5.2E-07 2.6E-09 La-140 2.5E-02 9.3E-07 4.6E-09 Ce-141 1.5E-04 6.1E-09 3.1E-11 Ce-143 2.8E-03 1.0E-07 5.1E-10 Ce-144 4.0E-03 1.6E-07 8.2E-10 W-187 2.5E-03 8.7E-08 4.4E-10 Np-239 2.2E-03 8.4E-08 4.2E-10 a These concentrations are from Table 6 of ANSI/ANS-18.1-1999.

b The concentrations given are for reactor coolant entering the letdown line. These concentrations are obtained from ANSI/ANS-18.1-1999.

c The values are based on a primary-to-secondary leak of 7.5E+01 lb/d.

d The concentrations given are for water in a steam generator.

e The concentrations given are for steam leaving a steam generator.

f These nuclides are in secular equilibrium; other radionuclide concentrations are those of the parent.

4-60

Table 4-38 Radionuclide concentrations in PWRs with once-through steam generators

- ANSI/ANS-18.1-1999 Reactor Coolantb Secondary Coolantc Radionuclidea (Ci/g) (Ci/g)

Noble Gases Kr-85m 1.6E-01 3.4E-08 Kr-85 4.3E-01 8.9E-08 Kr-87 1.7E-02 3.4E-09 Kr-88 1.8E-02 3.8E-09 Xe-131m 7.3E-01 1.5E-07 Xe-133m 7.0E-02 1.5E-08 Xe-133 2.9E-02 6.0E-09 Xe-135m 1.3E-01 2.7E-08 Xe-135 8.5E-01 1.8E-07 Xe-137 3.4E-02 7.1E-09 Xe-138 6.1E-02 1.3E-08 Halogens Br-84 1.6E-02 1.8E-08 I-131 2.0E-03 2.3E-09 I-132 6.0E-02 6.9E-08 I-133 2.6E-02 3.0E-08 I-134 1.0E-01 1.1E-07 I-135 5.5E-02 6.4E-08 Cesium & Rubidium Rb-88 1.9E-01 6.0E-07 Cs-134 3.7E-05 1.6E-10 Cs-136 8.7E-04 3.6E-09 Cs-137, Ba-137md 5.3E-05 2.2E-10 Water Activation Products N-16 4.0E+01 1.0E-06 Tritium H-3 1.0E+00 1.0E-03 Other Radionuclides Na-24 4.7E-02 1.1E-07 Cr-51 3.1E-03 6.9E-09 a These concentrations are from Table 7 of ANSI/ANS-18.1-1999.

b The concentrations given are for reactor coolant entering the letdown line. These concentrations are obtained from ANSI/ANS-18-1-1999.

c The values are based on a primary-to-secondary leak of 7.5E+01 lb/d. The concentrations given are for steam leaving a steam generator.

d These nuclides are in secular equilibrium; other radionuclide concentrations are those of the parent.

4-61

Table 4-38 Radionuclide concentrations in PWRs with once-through steam generators

- ANSI/ANS-18.1-1999 (cont.)

Reactor Coolantb Secondary Coolantc Radionuclidea (Ci/g) (Ci/g)

Mn-54 1.6E-03 3.5E-09 Fe-55 1.2E-03 2.7E-09 Fe-59 3.0E-04 6.7E-10 Co-58 4.6E-03 1.0E-08 Co-60 5.3E-04 1.2E-09 Zn-65 5.1E-04 1.1E-09 Sr-89 1.4E-04 3.1E-10 Sr-90 1.2E-05 2.7E-11 Sr-91 9.6E-04 2.1E-09 Y-91m 4.6E-04 9.7E-10 Y-91 5.2E-06 1.2E-11 Y-93 4.2E-03 9.3E-09 Zr-95 3.9E-04 8.7E-10 Nb-95 2.8E-04 6.2E-10 Mo-99 6.4E-03 1.4E-08 Tc-99m 4.7E-03 1.0E-08 Ru-103 7.5E-03 1.7E-08 Ru-106 9.0E-02 2.0E-07 Ag-110m 1.3E-03 2.9E-09 Te-129m 1.9E-04 4.2E-10 Te-129 2.4E-02 5.1E-08 Te-131m 1.5E-03 3.3E-09 Te-131 7.7E-03 1.5E-08 Te-132 1.7E-03 3.8E-09 Ba-140 1.3E-02 2.9E-08 La-140 2.5E-02 5.6E-08 Ce-141 1.5E-04 3.3E-10 Ce-143 2.8E-03 6.2E-09 Ce-144 3.9E-03 8.7E-09 W-187 2.5E-03 5.6E-09 Np-239 2.2E-03 4.9E-09 a These concentrations are from Table 7 of ANSI/ANS-18.1-1999.

b The concentrations given are for reactor coolant entering the letdown line. These concentrations are obtained from ANSI/ANS-18-1-1999.

c The values are based on a primary-to-secondary leak of 7.5E+01 lb/d. The concentrations given are for steam leaving a steam generator.

d These nuclides are in secular equilibrium; other radionuclide concentrations are those of the parent.

4-62

Table 4-39 Parameters used to describe the reference PWR with U-tube steam generators Nominal Range Parametera Symbol Units Value Maximum Minimum Thermal power P MWt 3.4E+03 3.8E+03 3.0E+03 Steam flow rate FS lb/h 1.5E+07 1.7E+07 1.3E+07 Weight of water in reactor coolant WP lb 5.5E+05 6.0E+05 5.0E+05 system Weight of water in all steam generators WS lb 4.5E+05 5.0E+05 4.0E+05 Reactor coolant letdown flow FD lb/h 3.7E+04 4.2E+04 3.2E+04 (purification)

Reactor coolant letdown flow (yearly FB lb/h 5.0E+02 1.0E+03 2.5E+02 average for boron control)

Steam generator blowdown flow (total) FBD lb/h 7.5E+04 1.0E+05 5.0E+04 Fraction of radioactivity in blowdown stream that is not returned to the NBD -- 1.0E+00b 1.0E+00 9.0E-01 secondary coolant system Flow through the purification system FA lb/h 3.7E+03 7.5E+03 0.0E+00 cation demineralizer Ratio of condensate demineralizer flow NC -- 0.0E+00c 0.0E+00 0.0E+00 rate to the total steam flow rate Ratio of the total amount of noble gases routed to gaseous radwaste from the purification system to the total amount of noble gases routed from the Y -- 0.0E+00 0.0E+00 0.0E+00 primary coolant system to the purification system (not including the boron recovery system) a These values are from Table 2-4 of NUREG-0017, Revision 1, and Table 2 of ANSI/ANS-18.1-1999.

b This value is based on a nominal case of blowdown through blowdown demineralizers back to the main condenser (no condensate demineralizers). The value is taken from blowdown demineralizer DFs in Section 4.1.1.17. Value for cesium and rubidium is 9.0E-01.

c This value is based on a nominal case of no condensate demineralizers. For a U-tube steam generator PWR with full-flow condensate demineralizers, the GALE-PWR 3.2 code uses a value of NC = 1.0E+00. For a U-tube steam generator PWR with condensate demineralizers and pumped-forward feedwater heater drains, the GALE-PWR 3.2 code uses a value for NC of 2.0E-01 for radioiodine and 1.0E-01 for cesium, rubidium, and other radionuclides as discussed in this section.

4-63

Table 4-40 Parameters used to describe the reference PWR with once-through steam generators Nominal Range Parametera Symbol Units Value Maximum Minimum Thermal power P MWt 3.4E+03 3.8E+03 3.0E+03 Steam flow rate FS lb/h 1.5E+07 1.7E+07 1.3E+07 Weight of water in reactor coolant WP lb 5.5E+05 6.0E+05 5.0E+05 system Weight of water in all steam generators WS lb 1.0E+05 b b Reactor coolant letdown flow FD lb/h 3.7E+04 4.2E+04 3.2E+04 (purification)

Reactor coolant letdown flow (yearly FB lb/h 5.0E+02 1.0E+03 2.5E+02 average for boron control)

Flow through the purification system FA lb/h 3.7E+03 7.5E+03 0.0E+00 cation demineralizer Ratio of condensate demineralizer flow NC -- 6.5E-01c 7.5E-01 5.5E-01 rate to the total steam flow rate Ratio of the total amount of noble gases routed to gaseous radwaste from the purification system to the total amount routed from the primary Y -- 0.0E+00 0.0E+00 0.0E+00 coolant system to the purification system (not including the boron recovery system) a These values are from Table 2-5 of NUREG-0017, Revision 1, and Table 3 of ANSI/ANS-18.1-1999.

b The secondary coolant inventory is not important in a once-through steam generator plant because decay is not an important removal mechanism for most of the radionuclides.

c For a PWR that is within the range indicated above (i.e., a PWR with pumped-forward feedwater heater drains), the GALE-PWR 3.2 code uses a value for NC of 2.0E-01 for radioiodine and 1.0E-01 for cesium, rubidium, and other radionuclides, as discussed in this section. For a PWR that has a full-flow condensate demineralizer, the GALE-PWR 3.2 code uses a value of NC = 1.0E+00.

4-64

Table 4-41 Values used in determining adjustment factors for PWRs Element Classa Symbol Description Cesium Water Noble Other Halogens & Activation H-3 Gases Radionuclides Rubidium Products Fraction of material removed in passing NA 0.0E+00 0.0E+00 9.0E-01 0.0E+00 0.0E+00 9.0E-01b through the cation demineralizer Fraction of material removed in passing NB through the 0.0E+00 9.9E-01 5.0E-01 0.0E+00 0.0E+00 9.8E-01 purification demineralizer Removal rate Rn 9.0E-04 6.7E-02 3.7E-02 0.0E+00 d 6.6E-02 reactor coolant (h-1)c Ratio of concentration in steam to that in water in the steam generator U-tube steam e 1.0E-02 5.0E-03 f 1.0E+00 5.0E-03 NS generator Once-through steam e 1.0E+00 1.0E+00 1.0E+00 1.0E+00 1.0E+00 generator Fraction of activity removed in passing NX through the 0.0E+00 9.0E-01 5.0E-01 0.0E+00 0.0E+00 9.0E-01 condensate demineralizers Removal ratesecondary coolant (h-1)g U-tube steam e 1.7E-01 1.5E-01 f d 1.7E-01 r generator Once-through steam e 2.7E+01 7.5E+00 f d 1.4E+01 generator Primary-to-FL secondary leakage 7.5E+01 7.5E+01 7.5E+01 7.5E+01 7.5E+01 7.5E+01 (lb/d) a These values are from Table 2-6 of NUREG-0017, Revision 1, and Table 9 of ANSI/ANS-18.1-1999.

b These represent effective removal terms and include mechanisms such as plateout. Plateout would apply to radionuclides such as molybdenum and corrosion products.

4-65

Table 4-41 Values used in determining adjustment factors for PWRs (cont.)

c These values of Rn apply to the reference PWRs whose parameters are given in Tables 4-39 and 4-40 and have been used in developing Tables 4-42 and 4-43. For PWRs not included in Tables 4-39 and 4-40 the appropriate value for Rn may be determined by the following equations:

+ ( )

=

()() + ( )( + )()

= , , ,

d The concentration of tritium is a function of (1) the inventory of tritiated liquids in the plant, (2) the rate of production of tritium caused by activation in the reactor coolant as well as releases from the fuel, and (3) the extent to which tritiated water is recycled or discharged from the plant. The tritium concentrations given in Tables 4-42 and 4-43 are representative of PWRs with a moderate amount of tritium recycle and can be used to calculate source terms in accordance with RG 1.112, Revision 1.

e Noble gases are rapidly transported out of the water in the steam generator and swept out of the vessel in the steam; therefore, the concentration in the water is negligible, and the concentration in the steam is approximately equal to the ratio of the release rate to the steam generator and the steam flow rate. These noble gases are removed from the system at the main condenser.

f Water activation products in reactor coolants exhibit varying chemical and physical properties that are not well defined. Most are not effectively removed by the demineralizers, but their concentrations are controlled by decay.

g These values of r apply to the reference PWRs whose parameters are given in Tables 4-39 and 4-40 and have been used in developing Tables 4-42 and 4-43. For PWRs not included in Tables 4-39 and 4-40, the appropriate value for r may be determined by the following equation:

()() + ()()()()

= , , ,

4-66

Table 4-42 Adjustment factors for PWRs with U-tube steam generators - ANSI/ANS-18.1-1999 Adjustment Factorsa Element Class Reactor Waterb Secondary Coolant (f) Water Steam 1.62E+02 (P) 9.0E-04 + 1.5E+07 Noble gases -- (f)

WP Rn + FS 1.62E+02 (P) 6.7E-02 + 4.5E+05 1.7E-01 + 4.5E+05 1.7E-01 +

Halogens (f) (f)

WP Rn + WS r + WS r +

Cesium & 1.62E+02 (P) 3.7E-02 + 4.5E+05 1.5E-01 + 4.5E+05 1.5E-01 +

Rubidium (f) (f)

WP Rn + WS r + WS r +

Water 4.5E+05 4.5E+05 activation 1.0E+00 products WS WS Tritium d d d Other 1.62E+02 (P) 6.6E-02 + 4.5E+05 1.7E-01 + 4.5E+05 1.7E-01 +

radionuclides (f) (f)

WP Rn + WS r + WS r +

Zn-65e 1.0E+01 1.0E+01 1.0E+01 Co-58f 1.0E+01 1.0E+01 1.0E+01 a These values are from Table 11 of ANSI/ANS-18.1-1999.

b f is the reactor water adjustment factor and is used in the secondary coolant adjustment factors.

c is the isotopic decay constant (h-1).

d The concentration of tritium is a function of (1) the inventory of tritiated liquids in the plant, (2) the rate of production of tritium caused by activation in the reactor coolant as well as releases from the fuel, and (3) the extent to which tritiated water is recycled or discharged from the plant. The tritium concentrations given in Tables 4-37 and 4-38 are representative of PWRs with a moderate amount of tritium recycle and can be used to calculate source terms in accordance with RG 1.112, Revision 1.

e Adjustment factors are for zinc addition plants using natural zinc. Use of depleted zinc would result in a lower adjustment factor, and the decrease is a function of the reduction of zinc-64.

f Adjustment factors are for zinc addition plants using natural or depleted zinc.

4-67

Table 4-43 Adjustment factors for PWRs with once-through steam generators -

ANSI/ANS-18.1-1999 Adjustment Factorsa Element Class Reactor Waterb Secondary Coolant (f) 1.62E+02 (P) 9.0E-04 + 1.5E+07 Noble gases (f)

WP Rn + FS 1.62E+02 (P) 6.7E-02 + 1.0E+05 2.7E+01 +

Halogens (f)

WP Rn + WS r +

1.62E+02 (P) 3.7E-02 + 1.0E+05 7.5E+00 +

Cesium & Rubidium (f)

WP Rn + WS r +

1.5E+05 Water activation products 1.0E+00 WS Tritium c c 1.62E+02 (P) 6.6E-02 + 1.0E+05 1.4E+01 +

Other radionuclides (f)

WP Rn + WS r +

a These values are from Table 2-8 of NUREG-0017, Revision 1, and Table 12 of ANSI/ANS-18.1-1999.

b f is the reactor water adjustment factor and is used in the secondary coolant adjustment factors.

c The concentration of tritium is a function of (1) the inventory of tritiated liquids in the plant, (2) the rate of production of tritium caused by activation in the reactor coolant as well as releases from the fuel, and (3) the extent to which tritiated water is recycled or discharged from the plant. The tritium concentrations given in Tables 4-37 and 4-38 are representative of PWRs with a moderate amount of tritium recycle and can be used to calculate source terms in accordance with RG 1.112, Revision 1.

4-68

Figure 4-4 Removal paths for the reference PWR with U-tube generators 4-69

Figure 4-5 Removal paths for the reference PWR with once-through steam generators Bases The radionuclide concentrations, adjustment factors, and procedure for effecting adjustments are based on the values and methods in ANSI/ANS-18.1-1999.

The values in Tables 4-37 and 4-38 provide a set of typical radionuclide concentrations in the primary and secondary systems for reactor designs within the parameters specified in Tables 4-39 and 4-40. The values in Tables 4-37 and 4-38 are those determined to be representative of radionuclide concentrations in a PWR, over its lifetime based on the ANSI/ANS-18.1-1999 data and models. The secondary coolant concentrations given in Tables 4-37 and 4-38 are calculated by using the reference parameters given in Table 4-41 and the equations given in Tables 4-42, and 4-43. Some systems will have design parameters that are outside the ranges specified in Tables 4-39 and 4-40. For that reason, Tables 4-37 through 4-43 provide a means of adjusting the concentrations to the actual design parameters. The adjustment factors in Tables 4-37 through 4-43 are based on Equation (4-20):

s C= (4-20) w (+R) K where C = the specific activity in µCi/g; K = a conversion factor, 4.54E+02 gallons per pound; 4-70

R = the removal rate of the radionuclide from the system because of demineralization; leakage, etc. (h-1);

s = the rate of release to and/or production of the radionuclide in the system in microcuries per hour; w = the fluid weight in lb; and

= the decay constant (h-1).

The following sample calculations illustrate the method by which the GALE-PWR 3.2 code will adjust the radionuclide concentrations in Tables 4-37 and 4-38. As indicated in Tables 4-42 and 4-43, adjustment factors will be calculated for noble gases, halogens, cesium, rubidium, and other radionuclides.

As an example, Table 4-44 presents sample case parameters compared with the range of values shown in Table 4-39.

Table 4-44 Sample case parameters versus the range of values shown in Table 4-39 Parameter (U-tube steam generator PWR) Value Range Thermal power level, MWt 3.8E+03 3.0E+03-3.8E+03 Steam flow rate, lb/h 1.7E+07 1.3E+07-1.7E+07 Mass of reactor coolant, lb 5.5E+05 5.0E+05-6.0E+05 Water weight in all steam generators, lb 4.4E+05 4.0E+05-5.0E+05 Reactor coolant letdown, lb/h 4.9E+04 3.2E+04-4.2E+04 Cation demineralizer flow, lb/h 4.9E+03 0.0E+00-7.5E+03 Shim bleed rateyearly average, lb/h 6.5E+02 2.5E+02-1.0E+03 Steam generator blowdown flow, lb/h 6.0E+04 5.0E+04-1.0E+05 Fraction of blowdown activity not returned to secondary 9.9E-01 9.0E-01-1.0E+00 system Cation demineralizer flow, lb/h 4.9E+03 0.0E+00-7.5E+03 Condensate demineralizer flow fraction 0.0E+00 0.0E+00-1.0E-02 Y (see definition in Table 4-39 and Section 3.2.4.1)

Because the parameter for reactor coolant letdown rate (4.9E+04 lb/h) in this example is outside the range specified in Table 4-39 (3.2-4.2E+04 lb/h), and the sample case employs continuous purging of the volume control tank, the PCA is recalculated using the actual design value for all parameters using the methods described below.

Noble Gases Coolant Activity (xenon-133 is used as an example.)

Using Equation (4-21) from Table 4-42 for noble gases, the adjustment factor, f, is calculated as follows:

4-71

1.62E+02 (P) 9.0E-04 +

f= WP (4-21)

Rn +

where the terms of the equation are as defined in Tables 4-39 and 4-41.

In calculating f, the variable Rn is calculated first by using Equation (4-22) given in Table 4-41 for noble gases:

FB+(FDFB) (Y)

Rn = (4-22)

WP Use the sample case parameters given above in Table 4-44 and the noble gas parameters given in Table 4-41 and substitute into Equation (4-22) above:

6.5E+02 + (4.9E+04 6.5E+02) (2.5E01)

Rn = = 2.3E-02 5.5E+05 Apply Equation (4-21) above using the computed value of Rn:

1.62E+02 (3.8E+03) 9.0E-04 + 5.5E03 f= 5.5E+05

= 2.5E-01 2.3E-02 + 5.5E03 The adjusted xenon-133 primary coolant concentration is computed as follows:

= (adjustment factor) x (standard xenon-133 concentration)

= 2.5E-01 x 2.9E+00 Ci/g = 7.3E-01 µCi/g

Halogens Coolant Activity (iodine-131 is used as an example.)

Using Equation (4-23) from Table 4-42 for halogens, the adjustment factor, f, is calculated as follows:

1.62E+02 (P) 6.7E-02 +

f= WP (4-23)

Rn +

where the terms in the equation are defined in Tables 4-39 and 4-41.

In calculating f, the variable Rn is calculated first by using Equation (4-24) given in Table 4-41:

(FD)(NB) + (1NB)(FB+FA)(NA)

Rn = WP (4-24) where the terms in the equation are as defined in Tables 4-39 and 4-41.

Use the sample case parameters given above in Table 4-44 and the halogen parameters given in Table 4-41, and substitute into Equation (4-24) above:

(4.9E+04)(9.9E01) + (1 9.9E01)(6.5E+02 + 4.9E+03)(0.0E+00)

Rn = 5.5E+05

= 8.8E-02 Use the value of Rn in Equation (4-23) above:

1.62E+02 (3.8E+03) 6.7E-02 + 3.6E03 f= 5.5E+05

= 8.6E-01 8.8E02 + 3.6E03 4-72

The adjusted iodine-131 concentration is computed as:

(adjustment factor) x (standard iodine-131 concentration)

= 8.6E-01 x 2.0E-03 µCi/g = 1.7E-03 µCi/g

Cesium and rubidium coolant activity (cesium-137 is used as an example.)

Using Equation (4-25) from Table 4-42 for cesium and rubidium, the adjustment factor, f, is calculated as follows:

1.62E+02 (P) 3.7E-02 +

f= WP (4-25)

Rn +

where the terms in Equation (4-25) are as defined in Tables 4-39 and 4-41.

In calculating f, the variable Rn is calculated first by using Equation (4-24) above. The cesium and rubidium parameters given in Table 4-41 and the sample case parameters given in Table 4-44 are used in Equation (4-24):

(4.9E+04)(5.0E01) + (1 5.0E01)(6.5E+02 + 4.9E+03)(9.0E01)

Rn = 5.5E+05

= 5.0E-02 Use the value of Rn in Equation (4-25) above:

1.62E+02 (3.8E+03) 3.7E-02 + 2.6E06 f= 5.5E+05

= 8.3E-01 5.0E02 + 2.606 The adjusted cesium-137 concentration is computed as follows:

= (adjusted factor) x (standard cesium-137 concentration)

= 8.3E-01 x 5.3E-03 µCi/g = 4.4E-03 µCi/g

Coolant Activity of Other Radionuclides (tellurium-132 is used as an example.)

Using Equation (4-26) from Table 4-42 for other radionuclides, the adjustment factor, f, is calculated as follows:

1.62E+02 (P) 6.6E-02 +

f= WP (4-26)

Rn +

where the terms in Equation (4-26) are as defined in Tables 4-39 and 4-41.

In calculating f, the variable Rn is calculated first by using Equation (4-24) above. The parameters for other radionuclides given in Table 4-41 and the sample case parameters given in Table 4-44 are used in Equation (4-24):

(4.9E+04)(9.8E01) + (1 9.8E01)(6.5E+02 + 4.9E+03)(9.0E01)

Rn = 5.5E+05

= 8.7E-02 Use the value of Rn in Equation (4-26) above:

1.62E+02 (3.8E+03) 6.6E-02 + 8.9E03 f= 5.5E+05

= 8.7E-01 8.7E02 + 8.903 4-73

The adjusted concentration of tellurium-132 is computed as:

= (adjustment factor) x (standard tellurium-132 concentration)

= 8.7E-01 x 1.7E-03 µCi/g = 1.5E-03 µCi/g The GALE-PWR 3.2 code uses a similar method to adjust secondary coolant concentrations for reactors with parameters outside the ranges specified in Tables 4-37 and 4-38.

The secondary coolant concentrations are based on the primary coolant concentrations as obtained above; on 7.5E+01 lb/d primary-to-secondary leakage in the steam generators; on appropriate steam generator carryover factors; on the appropriate main steam flow, steam generator blowdown flow, and fraction of blowdown flow returned to the secondary coolant, as defined in the plant design; and on the fraction of the radionuclides in the main steam that returns to the steam generators.

The secondary coolant concentrations are based on 7.5E+01 lb/d primary-to-secondary leakage. Table 4-45 (Table 2-11 of NUREG-0017, Revision 1), gives the primary-to-secondary leakage rate experience for 7-9 years of experience at operating PWRs. The average primary-to-secondary leakage rate in Table 4-45 is 7.5E+01 lb/d. Westinghouse estimates that the data in Table 4-45 are accurate within +/- 25 percent (Reference 3).

For plants using recirculating U-tube steam generators, carryover caused by mechanical entrainment is based on 0.5-percent moisture in the steam. Table 4-46 provides measured values for moisture carryover at five operating PWRs that use recirculating U-tube steam generators. Based on data from Turkey Point 3 and 4 (Reference 19), a value of 1-percent radioiodine carryover with the steam is used in these evaluations. For once-through steam generators, it is assumed that 100 percent of both nonvolatile and volatile species is carried over with the steam since this type of steam generator has no liquid reservoir and 100 percent of the feed is converted to steam.

For PWRs that use condensate demineralizers in the secondary system, the nominal value of the ratio of the condensate demineralizer flow rate to the total steam flow rate is 0.65. This indicates that the nominal case is a design that uses a pumped-forward modelthat is, one in which the reactor steam flow is split with 65 percent flowing to the low-pressure turbines and the main condenser and 35 percent pumped forward to the feedwater. The fraction pumped forward to the feedwater does not undergo any treatment in the condensate demineralizers.

The radioiodine, cesium, rubidium, and Other Radionuclides of Tables 4-37 and 4-38 preferentially go with the pumped-forward fraction. The reason for this is that these radionuclides show a tendency to go with the condensed steam in the moisture separator-reheater drains and with the extraction steamlines from the high-pressure turbines to the feedwater system. Based on data for Turkey Point, Point Beach, and Brunswick Steam Electric Plant (References 2, 3, and 19), the percentages used in the GALE-PWR 3.2 code for the amount of activity that is pumped forward and that bypasses the condensate demineralizers is 80 percent for radioiodine and 90 percent for cesium, rubidium, and the Other Radionuclides of Tables 4-37 and 4-38. Since the remaining radionuclides listed in Tables 4-37 and 4-38 are not removed in the condensate demineralizers these radionuclides are not considered the magnitude of bypass for those radionuclides.

The category Other Radionuclides includes molybdenum, yttrium, and technetium, which are generally present in colloidal suspensions or as crud. Although the actual removal mechanism 4-74

for these radionuclides is expected to be plateout or filtration, the quantitative effect of removal is expected to be commensurate with the removal of ionic impurities by ion exchange (within the accuracy of the calculations). Consequently, the parameters for ion exchange include plateout of these radionuclides.

Table 4-45 Monthly average primary/secondary leakage 1970b Planta (gal/d)

J F M A M J J A S O N D San Onofre 4 4 4 4 3 9 11 8 14 Sc S 0 Connecticut Yankee 0 10 0 S 0 0 20 10 20 0 0 0 R.E. Ginna 0 0 0 0 0 0 Point Beach 1 0 1971b Planta (gal/d)

J F M A M J J A S O N D San Onofre 0 0 0 0 0 0 0 0 0 0 0 0 Connecticut Yankee 0 30 15 0 0 10 20 20 15 40 40 40 R.E. Ginna 0 0 S S 0 0 0 0 0 0 0 0 H.B. Robinson 2 S S S S S S 0 50 55 20 Point Beach 1 0 0 0 10 90 100 53 30 20 20 20 20 1972b Planta (gal/d)

J F M A M J J A S O N D San Onofre S 0 0 0 0 0 22 0 10 30 4 31 Connecticut Yankee 40 40 40 40 40 S 0 0 0 0 0 0 R.E. Ginna 0 0 0 S S 0 0 0 0 S 0 0 H.B. Robinson 2 60 60 60 60 3 0 0 0 0 0 0 0 Point Beach 1 40 50 55 55 55 55 55 55 55 S S S Point Beach 2 0 0 0 Surry 1 0 Turkey Point 3 0 a These values are from Table 2-11 of NUREG-0017, Revision 1.

b Leakage values listed begin with the first year of commercial operation (at 70 ºF; density = 8.3E+00 lb/gal).

c Shutdown not included in average.

d Average daily value for each reactor is obtained by the sum of the total monthly leakage rates divided by the total number of days in operation.

S = Shutdown NA = Not available 4-75

Table 4-45 Monthly average primary/secondary leakage (cont.)

1973b Planta (gal/d)

J F M A M J J A S O N D San Onofre 3 3 0 0 0 0 0 0 0 0 S S Connecticut Yankee 0 0 0 0 10 S 0 S S S S 0 R.E. Ginna 0 0 0 0 0 0 0 0 0 0 0 0 H.B. Robinson 2 6 6 6 S 0 0 1 1 1 1 7 5 Point Beach 1 S S 0 0 0 0 0 0 0 0 0 0 Point Beach 2 0 0 0 0 0 0 0 0 0 0 0 0 Surry 1 0 0 0 0 0 0 0 0 0 0 0 0 Turkey Point 3 0 0 0 0 0 0 0 0 0 0 0 0 Surry 2 0 0 0 0 0 0 0 0 Turkey Point 4 0 0 0 0 1974b Planta (gal/d)

J F M A M J J A S O N D San Onofre 0 44 60 60 0 0 0 0 0 2 2 2 Connecticut Yankee 0 0 0 S 0 0 0 0 0 0 0 0 R.E. Ginna S S S 0 0 0 0 0 0 0 0 0 H.B. Robinson 2 2 10 112 98 NA 19 2 1 1 1 1 1 Point Beach 1 0 0 0 S 0 0 0 0 0 0 0 0 Point Beach 2 0 0 0 0 0 0 0 0 0 0 S S Surry 1 S S 0 0 0 115 55 115 115 4 S S Turkey Point 3 0 0 0 0 0 0 0 NA NA S S S Surry 2 0 0 0 0 S 38 0 0 0 S S S Turkey Point 4 S 0 0 0 0 0 0 22 0 0 0 0 Zion 1 S S S 0 0 0 S S 0 0 0 0 Zion 2 0 0 0 0 Indian Point 2 0 0 0 0 0 Prairie Island 1 0 0 0 0 0 0 a These values are from Table 2-11 of NUREG-0017, Revision 1.

b Leakage values listed begin with the first year of commercial operation (at 70 ºF; density = 8.3E+00 lb/gal).

c Shutdown not included in average.

d Average daily value for each reactor is obtained by the sum of the total monthly leakage rates divided by the total number of days in operation.

S = Shutdown NA = Not available 4-76

Table 4-45 Monthly average primary/secondary leakage (cont.)

1975b Planta (gal/d)

J F M A M J J A S O N D San Onofre 2 2 2 2 3 5 0 0 0 0 0 0 Connecticut Yankee 0 0 0 0 0 S 0 0 0 0 0 0 R.E. Ginna 0 0 3 S 0 0 0 0 0 0 0 0 H.B. Robinson 2 1 1 1 3 1 5 3 2 0 0 S 7 Point Beach 1 0 61 S 0 1 2 2 2 1 2 S S Point Beach 2 0 0 0 0 0 0 0 1 0 0 0 0 Surry 1 S 0 0 0 0 0 0 0 125 S S 26 Turkey Point 3 0 0 0 0 0 0 0 0 0 0 S S Surry 2 0 0 0 0 S 0 0 0 0 0 0 0 Turkey Point 4 0 0 0 S S S 7 20 79 0 0 50 Zion 1 0 0 S 0 0 S 0 0 S 0 0 0 Zion 2 0 S 0 0 0 S 0 0 S 0 0 0 Indian Point 2 0 102 S 0 0 0 0 0 0 S 0 0 Prairie Island 1 0 0 0 0 0 0 0 0 0 0 0 0 Prairie Island 2 0 0 0 0 0 0 0 0 0 0 0 0 Cook 1 0 0 0 0 a These values are from Table 2-11 of NUREG-0017, Revision 1.

b Leakage values listed begin with the first year of commercial operation (at 70 ºF; density = 8.3E+00 lb/gal).

c Shutdown not included in average.

d Average daily value for each reactor is obtained by the sum of the total monthly leakage rates divided by the total number of days in operation.

S = Shutdown NA = Not available 4-77

Table 4-45 Monthly average primary/secondary leakage (cont.)

1976b Planta (gal/d)

J F M A M J J A S O N D San Onofre 0 0 0 0 0 0 46 0 0 S S S Connecticut Yankee 0 0 0 0 S S 0 0 0 S 0 0 R.E. Ginna 0 S S 14 0 0 0 S 0 S 0 0 H.B. Robinson 2 2 1 1 1 2 1 2 2 2 6 S S Point Beach 2 32 200 5 29 10 12 13 21 23 25 25 25 Surry 1 0 0 28 86 NA 19 39 14 33 1 S S Turkey Point 3 12 6 14 0 11 19 0 12 1 S S S Surry 2 95 31 10 0 S 0 0 0 6 S S 200 Turkey Point 4 62 0 0 S S S 0 0 80 42 S 0 Zion 1 0 0 S S S S 0 0 0 S 0 0 Zion 2 S S 0 S S 0 0 0 0 S 0 0 Indian Point 2 0 0 0 S S S S S S 139 S S Prairie Island 1 0 0 S S 0 0 0 0 0 0 0 0 Prairie Island 2 S 0 0 0 0 0 0 0 0 S S S Cook 1 0 0 0 S S 0 0 0 0 0 0 0 Trojan 0 S S S 0 S S 0 Indian Point 3 0 S 0 0 0 Point Beach 1 0 0 3 3 3 2 3 3 3 S S 0 a These values are from Table 2-11 of NUREG-0017, Revision 1.

b Leakage values listed begin with the first year of commercial operation (at 70 ºF; density = 8.3E+00 lb/gal).

c Shutdown not included in average.

d Average daily value for each reactor is obtained by the sum of the total monthly leakage rates divided by the total number of days in operation.

S = Shutdown NA = Not available 4-78

Table 4-45 Monthly average primary/secondary leakage (cont.)

1977b Planta (gal/d)

J F M A M J J A S O N D San Onofre S S S 0 0 1 2 2 S 0 2 1 Connecticut Yankee 0 0 0 0 0 0 0 0 0 S S 0 R.E. Ginna 0 0 0 S S 0 0 0 0 0 0 0 H.B. Robinson 2 1 1 0 1 1 0 0 1 0 6 41 52 Point Beach 1 4 5 3 6 3 5 5 5 4 S 8 7 Point Beach 2 25 35 33 S 0 0 0 0 0 0 0 0 Surry 1 S 77 144 53 0 0 0 26 58 58 21 0 Turkey Point 3 0 0 0 0 0 0 0 28 72 72 56 S Surry 2 548 360 S 0 NA 18 10 8 4 0 14 0 Turkey Point 4 23 29 71 96 7 S S 0 0 4 0 0 Zion 1 0 0 0 0 0 0 0 0 S S S 0 Zion 2 S S S 0 0 0 0 0 0 0 0 0 Indian Point 2 0 0 0 S 0 0 S 0 0 0 0 0 Prairie Island 1 0 0 0 S 0 0 0 0 0 0 0 0 Prairie Island 2 0 0 0 0 0 0 0 0 0 1 S S Cook 1 S S 0 0 0 0 0 0 0 0 0 0 Trojan 0 0 0 0 S S 0 0 0 0 0 0 Indian Point 3 0 0 0 0 0 0 0 0 0 S S S Beaver Valley 1 0 0 S 0 0 S S 0 0 Salem 1 0 0 0 S S 0 Farley 1 0 a These values are from Table 2-11 of NUREG-0017, Revision 1.

b Leakage values listed begin with the first year of commercial operation (at 70 ºF; density = 8.3E+00 lb/gal).

c Shutdown not included in average.

d Average daily value for each reactor is obtained by the sum of the total monthly leakage rates divided by the total number of days in operation.

S = Shutdown NA = Not available 4-79

Table 4-45 Monthly average primary/secondary leakage (cont.)

1978b Planta (gal/d) Averaged J F M A M J gal/d San Onofre 1 1 1 S 1 1 4.6 Connecticut Yankee 0 0 0 0 0 0 5.7 R.E. Ginna 4 0 0 S S 0 0.27 H.B. Robinson 2 441 S S 18 88 190 21 Point Beach 1 20 7 7 7 120 7 15 Point Beach 2 0 0 0 S 0 0 7.9 Surry 1 0 0 0 0 S S 22 Turkey Point 3 S 0 0 0 0 0 5.5 Surry 2 0 46 278 0 0 0 32 Turkey Point 4 36 193 0 0 0 0 17 Zion 1 0 0 0 0 0 0 0 Zion 2 0 S S S S 0 0 Indian Point 2 0 S S S S 0 7.8 Prairie Island 1 0 0 0 S 0 0 0 Prairie Island 2 0 0 0 0 0 0 0.03 Cook 1 0 0 0 S S S 0 Trojan 2 2 2 S S S 0.38 Indian Point 3 0 0 0 0 0 S 0 Beaver Valley 1 0 0 0 0 S S 0 Salem 1 0 0 0 S S S 0 Farley 1 0 0 0 0 0 0 0 Operation Weighted Average 9 a These values are from Table 2-11 of NUREG-0017, Revision 1.

b Leakage values listed begin with the first year of commercial operation (at 70 ºF; density = 8.3E+00 lb/gal).

c Shutdown not included in average.

d Average daily value for each reactor is obtained by the sum of the total monthly leakage rates divided by the total number of days in operation.

S = Shutdown NA = Not available 4-80

Table 4-46 Moisture carryover in recirculating U-tube steam generators Facilitya Percent Carryoversb Palisades 0.08 Kansai 0.05 Point Beach 0.2 Turkey Point 3c 0.6 Turkey Point 4c 1.6 Average 0.5 a These values are from Table 2-12 of NUREG-0017, Revision 1.

b Measurement based on sodium concentration.

c These values are from NUREG/CR-1629.

4-81

4.2.2 ANSI/ANS-18.1-1984 Source Term Parameters Parameter As mentioned in Section 4.2, in GALE-PWR 3.2, the user can also select the ANS-18.1 Version-1984 for the reactor coolant source term, radionuclide concentrations in the primary and secondary coolant, as shown in Figure 3-4. The reactor coolant source term values for the ANS-18.1 Version-1984 option correspond to the values in ANSI/ANS-18.1-1984 and are also consistent with the guidance in DC/COL-ISG-05; RG 1.112, Revision 1; and NUREG-0800.

Tables 4-47 and 4-48 list the expected radionuclide concentrations in the reactor coolant and steam for PWRs from ANSI/ANS-18.1-1984. Figures 4-4 and 4-5 depict the relationship of the design parameters.

Bases The values in Tables 4-47 and 4-48 provide a set of typical radionuclide concentrations in the primary and secondary systems for reactor designs within the parameters specified in Tables 4-39 and 4-40. The reactor designs within the parameters specified in Tables 4-39 and 4-40 are the same for ANSI/ANS-18.1-1999 and ANSI/ANS-18.1-1984. The values in Tables 4-47 and 4-48 are those determined to be representative of radionuclide concentrations in a PWR over its lifetime based on the ANSI/ANS-18.1-1984 data and models. The secondary coolant concentrations given in Tables 4-47 and 4-48 are calculated by using the reference parameters in Table 4-41 and the equations in Tables 4-49 and 4-50. Some systems will have design parameters that are outside the ranges specified in Tables 4-39 and 4-40. For that reason, the concentrations are adjusted to the actual design parameters by a method similar to that used for the adjustments discussed in Section 4.2.1.2, with the exception of using the ANSI/ANS-18.1-1984 values in Tables 4-47 through 4-50.

4-82

Table 4-47 Radionuclide concentrations in PWRs with U-tube steam generators -

ANSI/ANS-18.1-1984 Secondary Coolantc Reactor Coolantb Radionuclidea Waterd Steame (Ci/g)

(Ci/g) (Ci/g)

Noble Gases Kr-85m 1.6E-01 3.4E-08 Kr-85 4.3E-01 8.9E-08 Kr-87 1.5E-01 3.0E-08 Kr-88 2.8E-01 5.9E-08 Xe-131m 7.3E-01 1.5E-07 Xe-133m 7.0E-02 1.5E-08 Xe-133 2.6E+00 5.4E-07 Xe-135m 1.3E-01 2.7E-08 Xe-135 8.5E-01 1.8E-07 Xe-137 3.4E-02 7.1E-09 Xe-138 1.2E-01 2.5E-08 Halogens Br-84 1.6E-02 7.5E-08 7.5E-10 I-131 4.5E-02 1.8E-06 1.8E-08 I-132 2.1E-01 3.1E-06 3.1E-08 I-133 1.4E-01 4.8E-06 4.8E-08 I-134 3.4E-01 2.4E-06 2.4E-08 I-135 2.6E-01 6.6E-06 6.6E-08 Cesium & Rubidium Rb-88 1.9E-01 5.3E-07 2.6E-09 Cs-134 7.1E-03 3.3E-07 1.7E-09 Cs-136 8.7E-04 4.0E-08 2.0E-10 Cs-137 9.4E-03 4.4E-07 2.2E-09 Water Activation Products N-16 4.0E+01 1.0E-06 1.0E-07 Tritium H-3 1.0E+00 1.0E-03 1.0E-03 Other Radionuclides Na-24 4.7E-02 1.5E-06 7.5E-09 Cr-51 3.1E-03 1.3E-07 6.3E-10 a These concentrations are from Table 6 of ANSI/ANS-18.1-1984.

b The concentrations given are for reactor coolant entering the letdown line. These concentrations are obtained from ANSI/ANS-18.1-1984.

c The concentrations are based on a primary-to-secondary leak of 7.5E+01 lb/d.

d The concentrations given are for water in a steam generator.

e The concentrations given are for steam leaving a steam generator.

4-83

Table 4-47 Radionuclide concentrations in PWRs with U-tube steam generators -

ANSI/ANS-18.1-1984 (cont.)

Secondary Coolantc Reactor Coolantb Radionuclidea Waterd Steame (Ci/g)

(Ci/g) (Ci/g)

Mn-54 1.6E-03 6.5E-08 3.3E-10 Fe-55 1.2E-03 4.9E-08 2.5E-10 Fe-59 3.0E-04 1.2E-08 6.1E-11 Co-58 4.6E-03 1.9E-07 9.4E-10 Co-60 5.3E-04 2.2E-08 1.1E-10 Zn-65 5.1E-04 2.1E-08 1.0E-10 Sr-89 1.4E-04 5.7E-09 2.9E-11 Sr-90 1.2E-05 4.9E-10 2.4E-12 Sr-91 9.6E-04 2.8E-08 1.4E-10 Y-91m 4.6E-04 3.2E-09 1.6E-11 Y-91 5.2E-06 2.1E-10 1.1E-12 Y-93 4.2E-03 1.2E-07 6.1E-10 Zr-95 3.9E-04 1.6E-08 7.9E-11 Nb-95 2.8E-04 1.1E-08 5.7E-11 Mo-99 6.4E-03 2.5E-07 1.2E-09 Tc-99m 4.7E-03 1.1E-07 5.7E-10 Ru-103 7.5E-03 3.1E-07 1.6E-09 Ru-106 9.0E-02 3.7E-06 1.8E-08 Ag-110m 1.3E-03 5.3E-08 2.7E-10 Te-129m 1.9E-04 7.8E-09 3.9E-11 Te-129 2.4E-02 2.2E-07 1.1E-09 Te-131m 1.5E-03 5.4E-08 2.7E-10 Te-131 7.7E-03 2.9E-08 1.5E-10 Te-132 1.7E-03 6.6E-08 3.3E-10 Ba-140 1.3E-02 5.2E-07 2.6E-09 La-140 2.5E-02 9.3E-07 4.6E-09 Ce-141 1.5E-04 6.1E-09 3.1E-11 Ce-143 2.8E-03 1.0E-07 5.1E-10 Ce-144 4.0E-03 1.6E-07 8.2E-10 W-187 2.5E-03 8.7E-08 4.4E-10 Np-239 2.2E-03 8.4E-08 4.2E-10 a These concentrations are from Table 6 of ANSI/ANS-18.1-1984.

b The concentrations given are for reactor coolant entering the letdown line. These concentrations are obtained from ANSI/ANS-18.1-1984.

c The concentrations are based on a primary-to-secondary leak of 7.5E+01 lb/d.

d The concentrations given are for water in a steam generator.

e The concentrations given are for steam leaving a steam generator.

4-84

Table 4-48 Radionuclide concentrations in PWRs with once-through steam generators

- ANSI/ANS-18.1-1984 Reactor Coolantb Secondary Coolantc Radionuclidea (Ci/g) (Ci/g)

Noble Gases Kr-85m 1.6E-01 3.4E-08 Kr-85 4.3E-01 8.9E-08 Kr-87 1.5E-01 3.0E-08 Kr-88 2.8E-01 5.9E-08 Xe-131m 7.3E-01 1.5E-07 Xe-133m 7.0E-02 1.5E-08 Xe-133 2.6E+00 5.4E-07 Xe-135m 1.3E-01 2.7E-08 Xe-135 8.5E-01 1.8E-07 Xe-137 3.4E-02 7.1E-09 Xe-138 1.2E-01 2.5E-08 Halogens Br-84 1.6E-02 1.8E-08 I-131 4.5E-02 5.2E-08 I-132 2.1E-01 2.4E-07 I-133 1.4E-01 1.6E-07 I-134 3.4E-01 3.8E-07 I-135 2.6E-01 3.0E-07 Cesium & Rubidium Rb-88 1.9E-01 6.0E-07 Cs-134 7.1E-03 3.0E-08 Cs-136 8.7E-04 3.6E-09 Cs-137 9.4E-03 3.9E-08 Water Activation Products N-16 4.0E+01 1.0E-06 Tritium H-3 1.0E+00 1.0E-03 Other Radionuclides Na-24 4.7E-02 1.1E-07 Cr-51 3.1E-03 6.9E-09 a These concentrations are from Table 7 of ANSI/ANS-18.1-1984.

b The concentrations given are for reactor coolant entering the letdown line. These concentrations are obtained from ANSI/ANS-18-1-1984.

c The concentrations are based on a primary-to-secondary leak of 7.5E+01 lb/d. The concentrations given are for steam leaving a steam generator.

4-85

Table 4-48 Radionuclide concentrations in PWRs with once-through steam generators

- ANSI/ANS-18.1-1984 (cont.)

Reactor Coolantb Secondary Coolantc Radionuclidea (Ci/g) (Ci/g)

Mn-54 1.6E-03 3.5E-09 Fe-55 1.2E-03 2.7E-09 Fe-59 3.0E-04 6.7E-10 Co-58 4.6E-03 1.0E-08 Co-60 5.3E-04 1.2E-09 Zn-65 5.1E-04 1.1E-09 Sr-89 1.4E-04 3.1E-10 Sr-90 1.2E-05 2.7E-11 Sr-91 9.6E-04 2.1E-09 Y-91m 4.6E-04 9.7E-10 Y-91 5.2E-06 1.2E-11 Y-93 4.2E-03 9.3E-09 Zr-95 3.9E-04 8.7E-10 Nb-95 2.8E-04 6.2E-10 Mo-99 6.4E-03 1.4E-08 Tc-99m 4.7E-03 1.0E-08 Ru-103 7.5E-03 1.7E-08 Ru-106 9.0E-02 2.0E-07 Ag-110m 1.3E-03 2.9E-09 Te-129m 1.9E-04 4.2E-10 Te-129 2.4E-02 5.1E-08 Te-131m 1.5E-03 3.3E-09 Te-131 7.7E-03 1.5E-08 Te-132 1.7E-03 3.8E-09 Ba-140 1.3E-02 2.9E-08 La-140 2.5E-02 5.6E-08 Ce-141 1.5E-04 3.3E-10 Ce-143 2.8E-03 6.2E-09 Ce-144 3.9E-03 8.7E-09 W-187 2.5E-03 5.6E-09 Np-239 2.2E-03 4.9E-09 a These concentrations are from Table 7 of ANSI/ANS-18.1-1984.

b The concentrations given are for reactor coolant entering the letdown line. These concentrations are obtained from ANSI/ANS-18-1-1984.

c The concentrations are based on a primary-to-secondary leak of 7.5E+01 lb/d. The concentrations given are for steam leaving a steam generator.

4-86

Table 4-49 Adjustment factors for PWRs with U-tube steam generators - ANSI/ANS-18.1-1984 Adjustment Factorsa Element Reactor Waterb Secondary Coolant Class (f) Water Steam 1.62E+02 (P) 9.0E-04 + 1.5E+07 Noble gases -- (f)

WP Rn + FS 1.62E+02 (P) 6.7E-02 + 4.5E+05 1.7E-01 + 4.5E+05 1.7E-01 +

Halogens (f) (f)

WP Rn + WS r + WS r +

Cesium & 1.62E+02 (P) 3.7E-02 + 4.5E+05 1.5E-01 + 4.5E+05 1.5E-01 +

Rubidium (f) (f)

WP Rn + WS r + WS r +

Water 4.5E+05 4.5E+05 activation 1.0E+00 products WS WS Tritium d d d Other 1.62E+02 (P) 6.6E-02 + 4.5E+05 1.7E-01 + 4.5E+05 1.7E-01 +

radionuclide (f) (f) s WP Rn + WS r + WS r +

a These values are from Table 2-7 of NUREG-0017, Revision 1, and Table 11 of ANSI/ANS-18.1-1984.

b f is the reactor water adjustment factor and is used in the secondary coolant adjustment factors.

c is the isotopic decay constant (h-1).

d The concentration of tritium is a function of (1) the inventory of tritiated liquids in the plant, (2) the rate of production of tritium caused by activation in the reactor coolant as well as releases from the fuel, and (3) the extent to which tritiated water is recycled or discharged from the plant. The tritium concentrations given in Tables 4-47 and 4-48 are representative of PWRs with a moderate amount of tritium recycle and can be used to calculate source terms in accordance with RG 1.112, Revision 1.

4-87

Table 4-50 Adjustment factors for PWRs with once-through steam generators -

ANSI/ANS-18.1-1984 Adjustment Factorsa Element Class Reactor Waterb Secondary Coolant (f) 1.62E+02 (P) 9.0E-04 + 1.5E+07 Noble gases (f)

WP Rn + FS 1.62E+02 (P) 6.7E-02 + 1.0E+05 2.7E+01 +

Halogens (f)

WP R + WS r +

1.62E+02 (P) 3.7E-02 + 1.0E+05 7.5E+00 +

Cesium & Rubidium (f)

WP Rn + WS r +

1.5E+05 Water activation products 1.0E+00 WS Tritium c c 1.62E+02 (P) 6.6E-02 + 1.0E+05 1.4E+01 +

Other radionuclides (f)

WP Rn + WS r +

a These values are from Table 2-8 of NUREG-0017, Revision 1, and Table 12 of ANSI/ANS-18.1-1984.

b f is the reactor water adjustment factor and is used in the secondary coolant adjustment factors.

c The concentration of tritium is a function of (1) the inventory of tritiated liquids in the plant, (2) the rate of production of tritium caused by activation in the reactor coolant as well as releases from the fuel, and (3) the extent to which tritiated water is recycled or discharged from the plant. The tritium concentrations given in Tables 4-47 and 4-48 are representative of PWRs with a moderate amount of tritium recycle and can be used to calculate source terms in accordance with RG 1.112, Revision 1.

4-88

4.2.3 ANSI/ANS-18.1-2016 Source Term Parameters Parameter As mentioned in Section 4.2, in GALE-PWR 3.2, the user can also select the ANS-18.1 Version-2016 for the reactor coolant source term, radionuclide concentrations in the primary and secondary coolant, as shown in Figure 3-4. The ANS-18.1 Version-2016 option should be described with sufficient detail to allow the NRC to conduct an independent evaluation. Tables 4-51 and 4-52 list the expected radionuclide concentrations in the reactor coolant and steam for PWRs from ANSI/ANS-18.1-2016. Figures 4-4 and 4-5 depict the relationship of the design parameters.

Bases The values in Tables 4-51 and 4-52 provide a set of typical radionuclide concentrations in the primary and secondary systems for reactor designs within the parameters specified in Tables 4-39 and 4-40. The reactor designs within the parameters specified in Tables 4-39 and 4-40 are the same for ANSI/ANS-18.1-1999 and ANSI/ANS-18.1-2016. The values in Tables 4-51 and 4-52 are those determined to be representative of radionuclide concentrations in a PWR over its lifetime based on the ANSI/ANS-18.1-2016 data and models. The secondary coolant concentrations in Tables 4-51 and 4-52 are calculated by using the reference parameters in Table 4-41 and the equations in Tables 4-53 and 4-54. Some systems will have design parameters that are outside the ranges specified in Tables 4-39 and 4-40. For that reason, the concentrations are adjusted to the actual design parameters by a method similar to that used for the adjustments discussed in Section 4.2.1.2, with the exception of using the ANSI/ANS-18.1-2016 values in Tables 4-51 through 4-54.

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Table 4-51 Radionuclide concentrations in PWRs with U-tube steam generators -

ANSI/ANS-18.1-2016 Secondary Coolantc Reactor Coolantb Radionuclidea Waterd Steame (Ci/g)

(Ci/g) (Ci/g)

Noble Gases Kr-85m 1.7E-03 3.6E-10 Kr-85 5.3E-02 1.1E-08 Kr-87 2.9E-03 6.0E-10 Kr-88 3.4E-03 7.0E-10 Xe-131m 7.3E-01 1.5E-07 Xe-133m 1.5E-03 3.2E-10 Xe-133 3.7E-02 7.7E-09 Xe-135m 7.3E-03 1.5E-09 Xe-135 1.3E-02 2.4E-09 Xe-137 3.4E-02 7.1E-09 Xe-138 8.8E-03 1.8E-09 Halogens Br-84 1.6E-02 7.5E-08 7.5E-10 I-131 1.2E-03 4.8E-08 4.8E-10 I-132 7.7E-03 1.1E-07 1.1E-09 I-133 4.8E-03 1.7E-07 1.7E-09 I-134 1.2E-02 9.0E-08 9.0E-10 I-135 8.3E-03 2.1E-07 2.1E-09 Cesium & Rubidium Rb-88 6.3E-04 1.8E-09 8.8E-12 Cs-134 5.5E-04 1.2E-09 6.2E-12 Cs-136 8.4E-05 3.3E-09 1.7E-11 Cs-137, Ba-137mf 3.5E-04 5.6E-11 2.8E-13 Water Activation Products N-16 4.0E+01 1.0E-06 1.0E-07 Tritium H-3 1.2E+00 1.2E-03 1.2E-03 Other Radionuclides Na-24 1.2E-03 3.9E-08 2.0E-10 Cr-51 2.2E-03 7.1E-08 3.6E-10 a These concentrations are from Table 6 of ANSI/ANS-18.1-2016.

b The concentrations given are for reactor coolant entering the letdown line. These concentrations are obtained from ANSI/ANS-18.1-2016.

c These values are based on a primary-to-secondary leak of 7.5E+01 lb/d.

d The concentrations given are for water in a steam generator.

e The concentrations given are for steam leaving a steam generator.

f These nuclides are in secular equilibrium; other radionuclide concentrations are those of the parent.

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Table 4-51 Radionuclide concentrations in PWRs with U-tube steam generators -

ANSI/ANS-18.1-2016 (cont.)

Secondary Coolantc Reactor Coolantb Radionuclidea Waterd Steame (Ci/g)

(Ci/g) (Ci/g)

Mn-54 6.0E-04 3.2E-09 1.6E-11 Fe-55 1.2E-03 4.9E-08 2.5E-10 Fe-59 3.6E-04 9.2E-09 4.6E-11 Co-58 2.1E-02 3.8E-07 1.9E-09 Co-60 4.9E-04 4.4E-10 2.2E-12 Zn-65 1.5E-04 9.8E-10 4.9E-12 Sr-89 1.4E-04 5.7E-09 2.9E-11 Sr-90 1.2E-05 4.9E-10 2.4E-12 Sr-91 9.6E-04 2.8E-08 1.4E-10 Y-91m 4.6E-04 3.2E-09 1.6E-11 Y-91 5.2E-06 2.1E-10 1.1E-12 Y-93 4.2E-03 1.2E-07 6.1E-10 Zr-95 3.4E-04 6.9E-09 3.5E-11 Nb-95 3.0E-04 8.6E-09 4.3E-11 Mo-99 4.5E-05 1.8E-09 8.9E-12 Tc-99m 1.7E-05 4.1E-10 2.1E-12 Ru-103 7.5E-03 3.1E-07 1.6E-09 Ru-106 9.0E-02 3.7E-06 1.8E-08 Ag-110m 1.3E-03 5.3E-08 2.7E-10 Te-129m 1.9E-04 7.8E-09 3.9E-11 Te-129 2.4E-02 2.2E-07 1.1E-09 Te-131m 1.5E-03 5.4E-08 2.7E-10 Te-131 7.7E-03 2.9E-08 1.5E-10 Te-132 1.7E-03 6.6E-08 3.3E-10 Ba-140 1.3E-02 5.2E-07 2.6E-09 La-140 2.5E-02 9.3E-07 4.6E-09 Ce-141 1.5E-04 6.1E-09 3.1E-11 Ce-143 2.8E-03 1.0E-07 5.1E-10 Ce-144 4.0E-03 1.6E-07 8.2E-10 W-187 3.4E-04 1.2E-08 6.0E-11 Np-239 1.2E-06 4.7E-11 2.4E-13 a These concentrations are from Table 6 of ANSI/ANS-18.1-2016.

b The concentrations given are for reactor coolant entering the letdown line. These concentrations are obtained from ANSI/ANS-18.1-2016.

c These values are based on a primary-to-secondary leak of 7.5E+01 lb/d.

d The concentrations given are for water in a steam generator.

e The concentrations given are for steam leaving a steam generator.

f These nuclides are in secular equilibrium; other radionuclide concentrations are those of the parent.

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Table 4-52 Radionuclide concentrations in PWRs with once-through steam generators

- ANSI/ANS-18.1-2016 Reactor Coolantb Secondary Coolantc Radionuclidea (Ci/g) (Ci/g)

Noble Gases Kr-85m 1.7E-03 3.6E-10 Kr-85 5.3E-02 1.1E-08 Kr-87 2.9E-03 6.0E-10 Kr-88 3.4E-03 7.0E-10 Xe-131m 7.3E-01 1.5E-07 Xe-133m 1.5E-03 3.2E-10 Xe-133 3.7E-02 7.7E-09 Xe-135m 7.3E-03 1.5E-09 Xe-135 1.3E-02 2.8E-09 Xe-137 3.4E-02 7.1E-09 Xe-138 8.8E-03 1.8E-09 Halogens Br-84 1.6E-02 1.8E-08 I-131 1.2E-03 4.2E-10 I-132 7.7E-03 2.7E-09 I-133 4.8E-03 1.7E-09 I-134 1.2E-02 4.4E-09 I-135 8.3E-03 3.0E-09 Cesium & Rubidium Rb-88 6.3E-04 3.9E-10 Cs-134 5.5E-04 3.5E-10 Cs-136 8.4E-05 5.4E-11 Cs-137, Ba-137md 3.5E-04 1.5E-10 Water Activation Products N-16 4.0E+01 1.0E-06 Tritium H-3 1.2E+00 1.2E-03 Other Radionuclides Na-24 1.2E-03 4.3E-10 Cr-51 2.2E-03 7.9E-10 a These concentrations are from Table 7 of ANSI/ANS-18.1-2016.

b The concentrations given are for reactor coolant entering the letdown line. These concentrations are obtained from ANSI/ANS-18-1-2016.

c These values are based on a primary-to-secondary leak of 7.5E+01 lb/d. The concentrations given are for steam leaving a steam generator.

d These nuclides are in secular equilibrium; other radionuclide concentrations are those of the parent.

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Table 4-52 Radionuclide concentrations in PWRs with once-through steam generators

- ANSI/ANS-18.1-2016 (cont.)

Reactor Coolantb Secondary Coolantc Radionuclidea (Ci/g) (Ci/g)

Mn-54 6.0E-04 2.1E-10 Fe-55 1.2E-03 2.7E-09 Fe-59 3.6E-04 1.3E-10 Co-58 2.1E-02 7.3E-09 Co-60 4.9E-04 1.7E-10 Zn-65 1.5E-04 5.3E-11 Sr-89 1.4E-04 3.1E-10 Sr-90 1.2E-05 2.7E-11 Sr-91 9.6E-04 2.1E-09 Y-91m 4.6E-04 9.7E-10 Y-91 5.2E-06 1.2E-11 Y-93 4.2E-03 9.3E-09 Zr-95 3.4E-04 1.2E-10 Nb-95 3.0E-04 1.1E-10 Mo-99 4.5E-05 1.6E-11 Tc-99m 1.7E-05 5.9E-12 Ru-103 7.5E-03 1.7E-08 Ru-106 9.0E-02 2.0E-07 Ag-110m 1.3E-03 2.9E-09 Te-129m 1.9E-04 4.2E-10 Te-129 2.4E-02 5.1E-08 Te-131m 1.5E-03 3.3E-09 Te-131 7.7E-03 1.5E-08 Te-132 1.7E-03 3.8E-09 Ba-140 1.3E-02 2.9E-08 La-140 2.5E-02 5.6E-08 Ce-141 1.5E-04 3.3E-10 Ce-143 2.8E-03 6.2E-09 Ce-144 3.9E-03 8.7E-09 W-187 3.4E-04 1.2E-10 Np-239 1.2E-06 4.4E-13 a These concentrations are from Table 7 of ANSI/ANS-18.1-2016.

b The concentrations given are for reactor coolant entering the letdown line. These concentrations are obtained from ANSI/ANS-18-1-2016.

c These values are based on a primary-to-secondary leak of 7.5E+01 lb/d. The concentrations given are for steam leaving a steam generator.

d These nuclides are in secular equilibrium; other radionuclide concentrations are those of the parent.

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Table 4-53 Adjustment factors for PWRs with U-tube steam generators - ANSI/ANS-18.1-2016 Adjustment Factorsa Element Reactor Waterb Secondary Coolant Class (f) Water Steam 1.62E+02 (P) 9.0E-04 + 1.5E+07 Noble gases -- (f)

WP Rn + FS 1.62E+02 (P) 6.7E-02 + 4.5E+05 1.7E-01 + 4.5E+05 1.7E-01 +

Halogens (f) (f)

WP Rn + WS r + WS r +

Cesium & 1.62E+02 (P) 3.7E-02 + 4.5E+05 1.5E-01 + 4.5E+05 1.5E-01 +

Rubidium (f) (f)

WP Rn + WS r + WS r +

Water 4.5E+05 4.5E+05 activation 1.0E+00 products WS WS Tritium d d d Other 1.62E+02 (P) 6.6E-02 + 4.5E+05 1.7E-01 + 4.5E+05 1.7E-01 +

radionuclides (f) (f)

WP Rn + WS r + WS r +

Zn-65e 1.0E+01 1.0E+01 1.0E+01 Co-58f 1.0E+01 1.0E+01 1.0E+01 a These values are from Table 11 of ANSI/ANS-18.1-2016.

b f is the reactor water adjustment factor and is used in the secondary coolant adjustment factors.

c is the isotopic decay constant (h-1).

d The concentration of tritium is a function of (1) the inventory of tritiated liquids in the plant, (2) the rate of production of tritium caused by activation in the reactor coolant as well as releases from the fuel, and (3) the extent to which tritiated water is recycled or discharged from the plant. The tritium concentrations given in Tables 4-51 and 4-52 are representative of PWRs with a moderate amount of tritium recycle and can be used to calculate source terms in accordance with RG 1.112, Revision 1.

e Adjustment factors are for zinc addition plants using natural zinc. Use of depleted zinc would result in a lower adjustment factor, and the decrease is a function of the reduction of zinc-64.

f Adjustment factors are for zinc addition plants using natural or depleted zinc.

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Table 4-54 Adjustment factors for PWRs with once-through steam generators -

ANSI/ANS-18.1-2016 Adjustment Factorsa Element Class Reactor Waterb Secondary Coolant (f) 1.62E+02 (P) 9.0E-04 + 1.5E+07 Noble gases (f)

WP Rn + FS 1.62E+02 (P) 6.7E-02 + 1.0E+05 2.7E+01 +

Halogens (f)

WP Rn + WS r +

1.62E+02 (P) 3.7E-02 + 1.0E+05 7.5E+00 +

Cesium & Rubidium (f)

WP Rn + WS r +

1.5E+05 Water activation products 1.0E+00 WS Tritium c c 1.62E+02 (P) 6.6E-02 + 1.0E+05 1.4E+01 +

Other radionuclides (f)

WP Rn + WS r +

a These values are from Table 12 of ANSI/ANS-18.1-2016.

b f is the reactor water adjustment factor and is used in the secondary coolant adjustment factors.

c The concentration of tritium is a function of (1) the inventory of tritiated liquids in the plant, (2) the rate of production of tritium caused by activation in the reactor coolant as well as releases from the fuel, and (3) the extent to which tritiated water is recycled or discharged from the plant. The tritium concentrations given in Tables 4-51 and 4-52 are representative of PWRs with a moderate amount of tritium recycle and can be used to calculate source terms in accordance with RG 1.112, Revision 1.

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4.3 PWRfixed-parameters.txt file This section discusses the user option to modify certain GALE-PWR fixed modeling parameters by means of the PWRfixed-parameters.txt file. The GALE-PWR 3.2 code zip file contains a sample PWRfixed-parameters.txt file (Figure 3-1), which allows the user to define 19 of the fixed modeling parameters from the default GALE86 code and ANSI/ANS-18.1-1999 values in the code.

The user can open and edit the sample PWRfixed-parameters.txt file in any text editor program (e.g., NotePad), as shown in Figure 4-6. The text file should start with $user and end with

$end and comments can be included after any !. Each line entry in the sample PWRfixed-parameters.txt file (Figure 4-6) contains a comment line, which indicates the reference source for the values used in the line entry in the sample PWRfixed-parameters.txt file.

Figure 4-6 PWRfixed-parameters.txt file Table 4-55 lists the fixed modeling parameters that can be defined using the PWRfixed-parameters.txt file. Figure 4-7 shows an example of a modified PWRfixed-parameters.txt file.

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Table 4-55 PWRfixed-parameters.txt file modifiable parameters Location in Code Parameter GALE-PWR Default Valuesa Line Code 3.1 Plant Capacity Factor (fraction) GE & LE 8.0E-01 Once-through steam GE generator (noble gases & See Table 4-38 halogens)

Coolant and Main Steam U-tube steam generator 3.2 Radionuclide Concentration (Ci/g) GE (noble gases & See Table 4-37 halogens)

Primary and secondary LE See Table 4-37 coolant Normal 8.0E-06 Containment operation Shutdown 3.2E-01 Normal Auxiliary 6.8E-01 operation building Radioiodine Release from Ventilation Shutdown 2.5E+00 3.3 Systems Prior to Treatment GE (Ci/yr/Ci/g) Normal Fuel handling 3.8E-02 operation building Shutdown 9.3E-02 Normal Turbine 3.8E+03 operation building Shutdown 4.2E+02 Containment See Table 4-8 Particulate Release from Ventilation Auxiliary building See Table 4-8 3.4 Systems Prior to Treatment GE (Ci/yr for each radionuclide) Fuel pool area See Table 4-8 Waste gas system See Table 4-8 Primary coolant leakage to 1.6E+02 auxiliary building (lb/d)

Steam leakage to turbine building Noble Gas Releases from Building 1.7E+03 3.5 GE (lb/d)

Ventilation Systems Primary coolant noble gas inventory leakage to containment 3.0E-02 building (fraction/d) a The default values are from Section 4.1.1 of NUREG-0017, Revision 1, and ANSI/ANS-18.1-1999.

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Table 4-55 PWRfixed-parameters.txt file modifiable parameters (cont.)

Location in Code Parameter GALE-PWR Default Valuesa Line Code Containment Building Purge 3.6 GE & LE 2.0E+00 Frequency Primary System Volumes Degassed 3.7 GE 2.0E+00 per Year Nonvolatile PC for U-tube steam 5.0E-03 generator Radioiodine PC for U-tube steam 1.0E-02 generator 3.8 Steam Generator Partition Coefficient LE Nonvolatile PC for once-through 1.0E+00 steam generator Radioiodine PC for once-through 1.0E+00 steam generator Radioiodine Releases from the Air 3.9 Ejector Exhaust Prior to Treatment GE 1.7E+03 (Ci/yr/pCi/g)

System operation time (h) 1.6E+01 Containment Internal Cleanup System mixing efficiency 3.10 GE 7.0E-01 System (fraction)

System particulate DF 1.0E+02 Annual Releases in Untreated 3.11 LE See Table 4-21 Detergent Waste (Ci/yr)

Tritium activity in primary coolant 1.0E+00 (Ci/yr) 3.12 Tritium Releases GE & LE Total tritium release (Ci/yr/MWt) 4.0E-01 Maximum fraction of tritium 9.0E-01 released through liquid pathway 3.13 Argon-41 Release (Ci/yr) GE 3.4E+01 3.14 Carbon-14 Release (Ci/yr) GE 7.3E+00 a The default values are from Section 4.1.1 of NUREG-0017, Revision 1, and ANSI/ANS-18.1-1999.

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Table 4-55 PWRfixed-parameters.txt file modifiable parameters (cont.)

Location in Code Parameter GALE-PWR Default Valuesa Line Code Anion DF for condensate 1.0E+01 demineralizers Other nuclide DF for condensate 3.15 DFs for Condensate Demineralizers LE 1.0E+01 demineralizers Cesium DF for condensate 2.0E+00 demineralizers Releases of Radioactive Material in 3.17 Liquid Waste from the Turbine LE 7.2E+03 Building Floor Drain System (gal/d)

Source Term Adjustments for AOOs 3.19 LE 1.6E-01 (Ci/yr) a The default values are from Section 4.1.1 of NUREG-0017, Revision 1, and ANSI/ANS-18.1-1999.

Figure 4-7 Example of a modified PWRfixed-parameters.txt file 4-99

1

5.0 REFERENCES

1. 10 CFR Part 50, U.S. Code of Federal Regulations, Title 10, Energy, Part 50, Domestic Licensing of Production and Utilization Facilities.

2. NUREG-0017, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE Code), U.S. Nuclear Regulatory Commission, Washington, DC, April 1976.

3. NUREG-0017, Revision 1, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE Code),

U.S. Nuclear Regulatory Commission, Washington, DC, April 1985. Available at ADAMS Accession No. ML112720411.

4. DC/COL-ISG-05, Interim Staff Guidance on the Use of the GALE86 Code for Calculation of Routine Radioactive Releases in Gaseous and Liquid Effluents from Boiling-Water Reactors and Pressurized-Water Reactors to Support Design Certification and Combined License Applications, U.S. Nuclear Regulatory Commission, Washington, DC, July 2008. Available at ADAMS Accession No. ML081710299.

5. Regulatory Guide 1.112, Revision 1, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Light-Water-Cooled Nuclear Power Reactors, U.S. Nuclear Regulatory Commission, Washington, DC, March 2007. Available at ADAMS Accession No. ML070320241.

6. NUREG-0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Chapter 11, Radioactive Waste Management, U.S. Nuclear Regulatory Commission, Washington, DC, March 2007. Available at ADAMS Accession No. ML070790010.

7. ANSI/ANS-18.1-1984, Radioactive Source Term for Normal Operation of Light Water Reactors, American National Standards Institute/American Nuclear Society, La Grange Park, IL, January 1984.

8. ANSI/ANS-18.1-1999, Radioactive Source Term for Normal Operation of Light Water Reactors, American National Standards Institute/American Nuclear Society, La Grange Park, IL, September 1999.

9. ANSI/ANS-18.1-2016, Radioactive Source Term for Normal Operation of Light Water Reactors, American National Standards Institute/American Nuclear Society, La Grange Park, IL, November 2016.

10. PNNL-24249, Software Quality Assurance Plan: Support for the Gaseous and Liquid Effluent (GALE) Computer Code Project, Pacific Northwest National Laboratory, Richland, WA, April 2015.

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11. PNNL-24250, Support for the Gaseous and Liquid Effluent (GALE) Computer Code Project: Configuration Management & Maintenance Plan, Pacific Northwest National Laboratory, Richland, WA, April 2015.

12. NUREG/BR-0167, Software Quality Assurance Program and Guidelines, U.S. Nuclear Regulatory Commission, Washington, DC, February 1993. Available at ADAMS Accession No. ML012750471.

13. PNNL-28818, GALE-3.2 Verification Report, Pacific Northwest National Laboratory, Richland, WA, June 2019.

14. Regulatory Guide 1.140, Revision 1, Design, Testing, and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants, U.S. Nuclear Regulatory Commission, Washington, DC, October 1979. Available at ADAMS Accession No. ML003740190.

15. Regulatory Guide 1.140, Revision 3, Design, Inspection and Testing Criteria for Air Filtration and Adsorption Units of Normal Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants, U.S. Nuclear Regulatory Commission, Washington, DC, August 2016. Available at ADAMS Accession No. ML16070A277.

16. EPRI NP-939, Sources of Radioiodine at Pressurized Water Reactors, Electric Power Research Institute, Palo Alto, CA, November 1978.

17. NUREG/CR-0140, In-Plant Source Term Measurements at Fort Calhoun Station Unit 1, U.S. Nuclear Regulatory Commission, Washington, DC, July 1978. Available at ADAMS Accession No. ML102560187.

18. NUREG/CR-0715, In-Plant Source Term Measurements at Zion Station, U.S. Nuclear Regulatory Commission, Washington, DC, February 1979. Available at ADAMS Accession No. ML102460323.

19. NUREG/CR-1629, In-Plant Source Term Measurements at Turkey Point Station Units 3 and 4, U.S. Nuclear Regulatory Commission, Washington, DC, September 1980. Available at ADAMS Accession No. ML102560188.

20. NUREG/CR-4397, In-Plant Source Term Measurements at Prairie Island Nuclear Generating Station, U.S. Nuclear Regulatory Commission, Washington, DC, September 1985.

21. NUREG/CR-2348, In-Plant Source Term Measurements at Rancho Seco Station, U.S. Nuclear Regulatory Commission, Washington, DC, October 1981. Available at ADAMS Accession No. ML10246032.

22. NUREG-0016, Revision 1, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Boiling Water Reactors (BWR-GALE Code),

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U.S. Nuclear Regulatory Commission, Washington, DC, January 1979. Available at ADAMS Accession No. ML091910213.

23. NUREG-75/087, Revision 1, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, U.S. Nuclear Regulatory Commission, Washington, DC, September 1975. Available at ADAMS Accession No. ML042080088.

24. ANSI/ANS-55.6-1979, Liquid Radioactive Waste Processing System for Light Water Reactor Plants, American National Standards Institute/American Nuclear Society, La Grange Park, IL, September 1979.

25. NUREG/CR-0143, The Use of Ion Exchange to Treat Radioactive Liquids in Light-Water-Cooled Nuclear Power Plants, U.S. Nuclear Regulatory Commission, Washington, DC, August 1978.

26. NUREG/CR-0142, The Use of Evaporation to Treat Radioactive Liquids in Light-Water-Cooled Nuclear Power Plants, U.S. Nuclear Regulatory Commission, Washington, DC, September 1978.

27. NUREG/CR-0141, The Use of Filtration to Treat Radioactive Liquids in Light-Water-Cooled Nuclear Power Plants, U.S. Nuclear Regulatory Commission, Washington, DC, September 1978.

28. NUREG/CR-0724, A Study of Reverse Osmosis Applicability to Light Water Reactor Radwaste Processing, U.S. Nuclear Regulatory Commission, Washington, DC, April 1979.

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NUREG-0017 Revision 2 Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized-Water Reactors July 2020 GALE-PWR 3.2 Code J.J. Tomon, K.J. Geelhood, and D.V. Colameco Technical Division of Systems Analysis Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555 Same as above.

John Tomon, NRC Project Manager This report is documentation for the release of the GALE-PWR 3.2 (Gaseous and Liquid EffluentsPressurized-Water Reactor 3.2) code. The GALE-PWR-3.2 code is a computerized mathematical model for calculating the releases of radioactive material in gaseous and liquid effluents (i.e., the gaseous and liquid source terms) from nuclear power plants (NPPs). The code is a tool that can be used to determine compliance with the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix I, Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the Criterion As Low as Is Reasonably Achievable for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents.

The purpose of the release of this version of the GALE-PWR code is to comprehensively verify the applicability of the current methodology described in NUREG-0017, Revision 1 (ADAMS Accession No. ML112720411), Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR GALE code), issued April 1985, to both the current U.S. PWR facilities and proposed NPP designs. Additionally, the GALE-PWR 3.2 code includes a graphical user interface (GUI) to facilitate easier use of the code.

Radioactive Effluents, Radioactive Waste, Radioactive Materials

NUREG-0017, Revision 2 Calculation of Releases of Radioactive Materials in Gaseous July 2020 and Liquid Effluents from Pressurized-Water Reactors

NUREG-0017, Revision 2 Calculation of Release of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized-Water Reactors

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NUREG-0017, Revision 2 Calculation of Release of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized-Water Reactors

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