ML20128C640

Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE Code)

ML20128C640
Person / Time
Issue date:	04/30/1985
From:	Chandrasekaran, Jun Lee, Willis C Office of Nuclear Reactor Regulation
To:
References
NUREG-0017, NUREG-0017-R01, NUREG-17, NUREG-17-R1, NUDOCS 8505280361
Download: ML20128C640 (208)

Text

_ __ _ . . -

NUREG-0017 '

Rev.1

, r Calculation of .

Releases of Radioactive Materials in Gaseous and Liquid Effluents i from Pressurized Water Reactors PWR-GALE Code U.S. Nuclear Regulatory Commission Offico of Nuclear Reactor Regulation l T. Chandrasekaran, J. Y. Lee, C. A. Willis forcoq i  !\ g 5: *

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g 52 g g 850430 0017 R PDR i

c_, y NOTICE I

Availability of Reference Materials Cited in NRC Publications

~

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l NUREG-0017 Rev.1 l

l Calculation of L Releases of Radioactive Materials

. in Gaseous and Liquid Effluents from Pressurized Water Reactors PWR-GALE Code t2Pu sh pri T. Ch ndrasekaran, J. Y. Lee, C. A. Willis Divl2 ion of Systems Integration Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Wnhington, D.C. 20665

ABSTRACT This report revises the original issuance of NUREG-0017, " Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE-Code)" (April 1976), to incorporate more recent operating data now available as well as the results of a number of,.in-plant measurement programs at operating pressurized water reactors.

The PWR-GALE 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). The U.S. Nuclear Regulatory Commission uses the PWR-GALE Code to determine conformance with the requirements of Appendix I to 10 CFR Part 50.

iii

TABLE OF CONTENTS Page ABSTRACT ..... . . . . . . . . . . . . . . . . . . . . . . . . . . iii l~ ACKNOWLEDGEMENTS . . . . . . . . . . . . . . . . . . . . . . . . . xi EXECUTIVE

SUMMARY

. . . . . . . . . . . . . . . . . . . . . . . . . . xi i i CHAPTER 1.- ' PWR-GALE CODE ,

1.1 Introduction . . ...................1-1 1.2 Definitions . . . . . . . . . . . . . . . . . . . . . . 1-3

- 1.3 Gaseous Source Terms . -. ...............1-6 1.4 Li qui d ' Sou rce Te rms . - . . . . . . . . . . . . . . . . . . 1 -7 1.5 Instructions for Completing PWR-GALE Code Input Data Ca rd s . . . . . . . . . . . - . . . . . . . . . . . . . . . 1-8

1. 5.1 Parameters Included in the PWR-GALE Code . . . 1 -8 1.5.2 Parameters Required for the PWR-GALE Code . . . 1-13 CHAPTER 2. PRINCIPAL. PARAMETERS USED IN PWR SOURCE TERM

- CALCULATIONS AND THEIR BASES

- 2.1 Introduction . . . . . . . . . . . . . . . . . . . . . 2-1 2.2 Principal Parameters and Their Bases . . . . .-. . . . 2-1 2.2.1 Th e rma l P owe r Le vel . . . . ~ . . . . . . . . . . 2-1 2.2.2 Plant Capacity Factor . . . . . . . . . . . . . 2-1 2.2.3 Radionuclide Concentrations in the Primary and Secondary Coolant . . . ... . . . . . . . . 2-2 2.2.4 Iodine Releases from Building Ventilation Systems . . . . . . . ... . . . . . . . . . . 2-33 2.2.5 Radioactive Particulates Released in Gaseous Effluents . . . . . . . . . . . . . . . . . 2-40 2.2.6 Noble Gas Releases from Building Ventilation Systems -. . . . . . . . . . . . . . . . . . 2-42 2.2.7 Steam Generator Blowdown Flash Tank Vent . . . 2-48

' 2.2.8 Iodine Releases from Main Condenser Air Ejector 4 Exhaust . . . . . . . . . . . . . . . . . . 2-48 1 2.2.9 Containment Purge Frequency . . . . . . . . . . 2-51 2.2.10 Containment Internal Cleanup System . . . . . . 2-55 2.2.11 Radioiodine Removal Efficiencies for Charcoal Adsorbers and Particulate Removal Efficiencies for HEPA Filters . . . . . . . 2-56 2.2.12 Waste Gas System Input Flow to Pressurized Storage Tanks . . . . . . . . . . . . . . . 2-57 2.2.13 Holdup Times for Charcoal Delay System . . . . 2-60

-2.2.14 Liquid Waste Inputs . . . . . . -. . . . . . . . 2-61 2.2.15 De t e rge nt Wa s te . . . . . . . . . . . . . . . . 2- 61

- 2.2.16 Chemical Wastes from Regeneration of Condensate Demineralizers . . . . . . . . . 2-68 v

I - .- +.m7 m + m. e - ..- - anw,

Page 2.2.17 Tritium Relea'ses . . . . . . . . . . . . . . . 2-68 2.2.18 Decontamination Factors for Demineralizers . . 2-77 2.2.19 Decontamination Factors for Evaporators . . . . 2-79 2.2.20 Decontamination Factors for Liquid Radwaste Filters . . . .-. . . . . . . . . . . . . . 2-80 l

.2.2.21 Decontamination Factors for Reverse Osmosis . . 2-80 l 2.2.22 Guideline for Calculating Liquid Waste Holdup Tines . . . . . . . . . . . . . . . . . . . 2-81 2.2.23 _ Adjustment to Liquid Radwaste Source Terms for Anticipated Operational Occurrences . . . . 2-86 2.2.24 Atmospheric Steam Dump . . . . . . . . . . . . 2-89 2.2.25 Carbon-14 Releases . . . . . . . . . . . . . . 2-90 2.2.26 Argon-41 Releases . . . . . . . . . . . . . . . 2-93 CHAPTER 3. INPUT FORMAT, SAMPLE PROBLEM, AND FORTRAN LISTING OF THE PWR-GALE CODE 3.1 Introduction . . . . . . . . . . . . . . . . . . . . . 3-1 3.2 Input Data . . . . . . . . . . . . . . . . . . . . . . 3-1 3.2.1 Explanation of the Inputs for the Sample Problem . 3-1 3.2.2 Input Coding Sheets . . . . . . . . . . . . . . 3-5 3.3 Sample Problem -- Input and Output . . . . . . . . . . 3-5 3.4 Listing of PWR-GALE Code . . . . . . . . . . . . . . . 3-5 3.4.1 Nuclear Data Library . . . . . . . . . . . . . . 3-5 3.4.2 FORTRAN Program Listing . . . . . . . . . . . . 3-15 CHAPTER 4. DATA FOR RADI0 ACTIVE SOURCE TERM CALCULATIONS FOR PRESSURIZED WATER REACTORS 4.1 General . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.2 Primary System . . . . . . . . . . . . . . . . . . . . 4-1 4.3 Secondary System . . . . . . ... . .-. . . . . . . . . 4-1 4.4 Liquid Waste Processing Systems . . . . . . . . . . . . 4-2 4.5 Gaseous Waste Processing System ...........4-3 4.6 Ventilation and Exhaust Systems . . . . . . . . . . . . 4-3 '

APPENDIX A Liquid Source Term Calculational Procedures for Regenerant Wastes from Demineralizers Other Than Condensate Demi ne ra l i ze rs . . . . . . . . . . . . . . . . . . . . . . . . A-1 REFERENCES R-1 vi

LIST OF TABLES Page Table 1-1 .Radioiodine Releases from Building Ventilation Systems Prior to Treatment . . . . . . . . . . . . . . . . . 1-9 1-2 Radioactive Particulate Releases from Building Ventilation Systems Prior to Treatment . . . . . . . 1-10 1-3 'PWR Liquid Wastes . . . . . . . . . . . . . . . . . . .

1-18 1-4 Decontamination Factors for PWR Liquid Waste T rea tme nt Sy s tems . . . . . . . . . . . . . . . . . . . 1 -21 1-5 Assigned Removal Efficiencies for Charcoal Adsorbers for Radiciodine Removal . . . . . . . . . . . . . . . 1-29 2-1 Plant Capacity Factors at Operating PWR's . . . . . . . 2-3 2-2 Numerical Values - Concentrations in Principal Fluid Streams of the Reference PWR with U-Tube Steam Generators . . . . . . . . . . . ... . . . . . . . . 2-4 2-3 Numerical Values - Concentrations in Principal Fluid Streams of the Reference PWR with Once-Through Steam Generators . . . . . . . . . . . . . . . . . . . 2-6 2-4 Parameters Used to Describe the Reference-Pressurized.

Water Reactor with U-Tube Steam Generators . . . . . 2-8 2-5 Parameters Used to Describe the Reference Pressurized Water Reactor with Once-Through Steam Generators . . 2-9 2-6 Values Used in Determining Adjustment Factors for Pressurized Water Reactors .............2-10

.2-7 Adjustment Factors for PWR's with U-Tube Steam Generators .....................2-12 2-8 Adjustment Factors for PWR's with Once-Through Steam Generators ..................2-13 I

2-9 Summary of Iodine-131 and Iodine-133 Primary

. Coolant Concentrations in PWR's . . . . . . . . . . . 2-21 2-10 Summary of Radionuclide Primary Coolant Concentrations in PWR's . . . . . . . . . . . . . . . . . . . . . . 2-22 2-11 Monthly Average Primary / Secondary Leakage . . . . . . . 2-26 2-12 Moisture Carryover in Recirculating U-Tube Steam Generators . . . . . . . . . . . . . . . . . . . . . 2-32 vii

Table Page l l

2-13 Annual Iodine Normalized Releases from Containment Ventilation Systems . . . . . . . . . . . . . . . . . 2-34 2 Annual Iodine Normalized Rele.tses from Auxiliary Bldg.

Ventilation Systems . . . . . . . . . . . . . . . . . 2-35 2-15 Annual' Iodine Normalized Releases from Refueling Area Ventilation Systems . . . . . . . . . . . . . . . . . 2-36 2-16 Annu'al Iodine Normalized Releases from Turbine Building Ventilation Systems . . . . . . . . . . . . . . . . . 2-37 2-17 Particulate Release Rate for Gaseous Effluents . . . . 2-42 2-18 Measured Release Upstream of HEPA Filters -

Containment . . . . . . . . . . . . . . . . . . . . .-. 2-43 2-19 Measured Releases Upstream of HEPA Filters - Auxiliary

' Building . . . . . . . . . . . . . . . . . . . . . . 2-44 2-20 Measured Releases Upstream of HEPA Filters - Fuel P o ol A re a . . . . . . . . . . . . . . . . . . . . . . . 2-45 2-21 Measured Releases Upstream Filters - Waste Gas System . .-. . . . . . . . ... . . . . . . . . . . . 2-46 2-22 Annual Iodine Normalized Releases from Main Condenser Ai r Ejector Exhaust . . . . . . . . . . . . . . . . . 2-49 2-23 PWR Containment Purging and Venting Experience . . . . 2-52 2-24' Waste Gas System Input Flow to Pressurized Storage Tanks and PWR's Without Recombiners . . . . . . . . . 2-58

' 2-25 Waste: Gas System Input Flow to Pressurized Storage Tanks for PWR's with Recombiners . . . . . . . . . . 2-59 2-26 PWR Li quid Wa stes . . . . . . . . . . . . . . . . . . . 2-65 1

2 Calculated Annual Release of Radioactive Material in Untreated Detergent Waste . . . . . . . . . . . . . . 2-67 2-28 Radionuclide Distribution of Detergent Waste . . . . . 2-69 l 2-29 Tritium Release Data from Operating PWR's with .

Zi rca l oy-Cl a d F u el s . . . . . . . . . . . . . . . . . . 2-71 2-30 Tritium Release Rate from Operating PWR's as a Function of Number of Years of Operation . . . . . . . . . . . 2-75 L

viii

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. 2-32 Distribution of Tritium Release in Gaseous Effluents . . 2-78 9.J l _js .;

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.. c Frequency and Extent of Unplanned Liquid Radwaste S '

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LIST OF FIGURES Figure Page 2-1 Removal Paths for Pressurized Water Reactor with U-Tube Steam Generators . . . . . . . . . . . . . . . 2-14 2-2 Removal Paths for Pressurized Water Reactor with Once-Through Steam Generators . . . . . . . . . . . . 2-15 2-3 Krypton and Xenon K Values as a Function of Reciprocal Temperature . . . . . . . . . . . . . . . 2-62 2-4 Effect of Moisture Content on the Dynamic Adsorption Coefficient . . . . . . . . . . . . . . . . . . . . . 2-63 2-5 Charcoal Moisture as a Function of Relative Humidity . 2-64 3-1 Input Coding Sheets for Sample Problem ........3-6 3-2 Printout of Input and Output for the Sample Problem . . 3-8 3-3 Program Listing for Gaseous Determination . . . . . . . 3-16 3-4 Program Listing for Liquid Determination . . . . . . . 3-26

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ACKNOWLEDGMENTS Many-individuals contributed to the preparation of Revision 1 of this

-document. In' particular, R. L. Bangart and F. P. Cardile of the Nuclear Regulatory Commission, and A. E. Refre of the Philippine Atomic Energy l Commission were the principal contributors. J. W. Mandler and F. Y. Tsang j of EG8G Idaho' edited this revision and have incorporated the results of a number of in-plant measurement programs at operating pressurized water reactors under an INEL Technical Assistance contract to the Division of

Systems Integration, U. S. Nuclear Regulatory Commission (FIN A-6460). The Revision 0 of this document was prepared by L. G. Bell, M. J. Bell, R. R.

Bellamy, J. S. Boegli, W. C. Burke, = F. P. Cardile, J. T. Collins, J. Y. Lee,

'P. G. Stoddart, W. D. Travers, and R. A. Weller of the Nuclear Regulatory Commi ssion.

I 5

xi

EXECUTIVE

SUMMARY

The average quantity of radioactive material released to the environment from a nuclear power reactor during normal operation including anticipated operational occurrences is called the " source term,"* since it is the source or initial number used in calculating the environmental impact of radioactive releases. The PWR-GALE (Pressurized Water Reactor - Gaseous and Liquid Effluents) Code is a computerized matWematical model Tor calcillating the releases of radioactive material in gaseous and liquid effluents (i.e., the ..

gaseous and liquid source terms) from pressurized water reactors. The calcu-lations are based on data generated from operating reactors, field and labo-ratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment during normal operation, including anticipated operational occurrences.

The U.S. Nuclear Regulatory Commission uses the PWR-GALE Code to determine conformance with the requirements of Appendix I to 10 CFR Part 50. The first issue of this NUREG report was published in April 1976. In order to use the best available data for improving the calculational models used by the Com-mission staff to determine conformance with Appendix I to 10 CFR Part 50, Revision 1 is being issued to update NUREG-0017. This revision incorporates more recent operation data now available and also incorporates.the results of a number of in-plant measurement programs at operating pressurized water reactors.

Chapter 1 of this report gives a step-by-step procedure for using the PWR-GALE Code along with a description of the parameters which have been built into the Code for use with all PWR source term calculations. These parameters, which apply generically to all PWR's, have been incorporated into the Code to eliminate the need for their entry on input data cards. Other parameters are required to be entered on input cards used by the Code. Explanations of the data require, along with acceptable means for calculating such data, are given for each input data card.

Descriptions of the principal parameters used in source term calculations and explanations of the bases for each parameter are given in Chapter 2. The -

parameters have been derived from reactor operating exper.ience where data were available. Where operating data were inconclusive or not available, informa-tion was drawn from laboratory and field tests and from engineering judgment.

The bases for the source term parameters explain the reasons for choosing the numerical values listed. A list of references used in developing the parame-ters is also included. The source term parameters used are believed to provide a realistic assessment of reactor and radwaste system operation.

• " Source term" as discussed in this report differs from " accident source term," which deals with potential releases resulting from nuclear reactor .

accidents.

xiii

l Chapter 3 contains sample input data together with an explanation of the input to orient the user in making the required entries. Also included is a l listing of the input data for a sample problem, a discussion of the nuclear data library used, and a FORTRAN listing of the PWR-GALE Code.

Chapter 4 lists the information needed to generate source terms for PWR's. .The information is proved by the applicant and is consistent with the contents of the Safety Analysis Report (ER) of the proposed PWR. This information constitutes the basic data required in calculating the releases of radioactive material in liquid and gaseous effluents.

xiv

CHAPTER 1. PWR-GALE CODE

1.1 INTRODUCTION

In promulgating Appendix I to 10 CFR Part 50, the U. S. Nuclear Regulatory Commission indicated its desire to use the best available data for improving the calculational models used by the Commission Staff to determine conformance with the requirements of the regulation. The first issue of this NUREG Report was published in April 1976. Revision 1 is being issued to update NUREG-0017 by incorporating more recent operating data now available and also by incorporating the results of a number of in-plant measurement programs at operating pressurized water reactors (PWR's).

The PWR-GALE (Pressurized Water Reactor - Gaseous and Liquid Effluents)

Code is a computerized mathematTcal model for calculating the releases of radioactive material in gaseous and liquid effluents from pressurized water reactors. The calculations are based on data generated from operating reactors, field and laboratory tests, and plant-sr,ecific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment during normal operation, including anticipated operational occurrences.

The average quantity of radioactive material released to the environ-ment from a nuclear power reactor during normal operation is called the

" source term" since it is the source or initial number used in calculating the environmental impact of radioactive releases. The calculations performed by the PWR-GALE Code are based on (1) American Nuclear Society (ANS) 18.1 Working Group recommendations (Ref.1) for 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) information received on the operation of nuclear power plants.

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, and (4) activity removed because of primary coolant leakage. These mechanisms are described briefly in the following paragraphs.

1-1

The primary coolant is continuously purified by passing a side stream through filters and demineralizers in the reactor coolant treatment systems (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 and/or improve fuel economy. 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 degassification are normally collected in pressurized storage tanks and held for radioactive decay prior to recycle or release to the environment. Alternative treatment methods include charcoal delay systems and cryogenic distillation.

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 is dependent upon the temperature and pressure at the point where the leakage occurs. Most of the noble gases enter the gas phase, whereas iodine 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 leakir.g from the primary coolant concentrate in the liquid phase in the '

steam generator. The degree of concentration is controlled by the steam '

generator blowdown rate and condensate demineralizer flow rate.

Since there is no liquid reservoir in a once-through steam generator, s the primary coolant leakage boils to steam when it enters the secondary 1-2

side of the steam generator. Secondary coolant purity is maintained by a condensate demineralizer system and there is no steam geierator blowdown.

The concentration of radioactivity in the secondary coolant is controlled by the condensate demineralizer flow rate.

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/

condensate systems.

In this chapter, a step-by-step procedure for using the PWR-GALE Code is given alonn with a description of the parameters which have been built into the Code for use with all PWR source term calculations. These parameters, which apply generically to all PWR's, have been incorporated into the Code to eliminate the need for their entry on input data cards.

Other parameters are required to be entered on input data cards used by the Code. Explanations of the data required, along with acceptable means for calculating such data, are given for each input data card.

Chapter 2 gives the principal source term parameters developed for use with the PWR-GALE Code and explains the bases for each parameter.

Chapter 3 contains a sample data input sheet and a Fortran IV listing of the PWR-GALE Code. Chapter 4 lists the information needed to generate source terms that an applicant is required to submit with the application.

1.2 DEFINITIONS The following definitions apply to terms used in this report: g.pg yb Activation Gases: The gases (including oxygen, nitrogen, and argon) gQ that become radioactive as a result of irradiation in the core. T 9;3 c: n Anticipated Operational Occurrences: Unplanned releases of radioactive [

materials from miscellaneous actions such as equipment failcre, operator r. 3 error, administrative error, that are not of consequence to be considered C.. y an accident. p /;

Normally liquids that contain relatively high fyo;f Chemical Waste Steam:

concentrations of decontaminants, regenerants, or chemical compounds k[y-j B .,

other than detergents. These liquids originate primarily from resin 6 4.

• regenerant and laboratory wastes. @h Clean Waste System: Normally tritiated, nonaerated, low-conductivity liquids consisting primarily of liquid waste collected from equipment h

'ff; a leaks and drains and certain valve and pump seal leakoffs. These liquids -Q#. 4 originate from systems containing primary coolant and are normally ;Q W reused as primary coolant makeup water. .

a 1-3 l

Decontamination Factor (DF): The ratio of the initial amount of a nuclide in a stream (specified in terms of concentration or activity of radioactive materials) to the final amount of that nuclide 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 a low radioactivity content.

Dirty Waste Stream (Floor Drains): Normally nontritiated, aerated, high-conductivity, non-primary-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 100% licensed power to produce the integrated thermal power output during a calendar year, i.e.,

I Ef fective Full Power Days = Integrated Thermal Power i PT jj Licensed Power Level "P total where P

j is the ith power level, in MWt; P

total is the licensed power level, in MWt; and Tj is the time of operation at power level Pj , in days.

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

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

Liquid Effluent Stream: Processedliquidwastescontainingr$dioactive materials resulting from the operation of a nuclear power reactor.

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

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

1-4

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 primary coolant activity is considered to be constant over a range of power levels, coolant and cleanup flows, and coolant volumes.

Radionuclide concentrations given in this NUREG are based on a recent compilation of available operating data. Therefore, the concentration values in NUREG-0017, Rev. I differ from the ANSI N237 values (Ref.1).

Provisions are made in the PWR-GALE Code, in accordance with the recommendations of the standard, for adjusting coolant concentrations should the. plant be designed to parameters that are outside the ranges considered in the standard. The radionuclide concentrations used are considered to' be representative of measured values based on the available

' operating data. The radionuclides are divided into the following categories:

1. Noble gases

2. Halogens (Br,I)

3. Cs, Rb

4. Water activation products

5. Tritium

6. Other nuclides (as listed in Tables 2-2 and 2-3 of Chapter 2 of this document)

Radioactive Halogens: The isotopes of fluorine, chlorine, bromine, and iodine. The radicactive isotopes of iodine are the key isotopes considered in dose calculations.

Radioactive Noble Gases: The radioactive isotopes of helium, neon, argon, krypton, xenon, and radon, which are characterized by their chemical inactivity. The radioactive isotopes of 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 anticipated operational occurrences.

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.

1-5

Source' Term: The calculated average quantity of radioactive material released to the environment from a nuclear power reactor during normal operation, including anticipated operational occurrences. The source term is the isotopic distribution of radioactive materials used in evaluating the impact of radioactive releases on the environment.

Steam Generator Blowdown: Liquid removed from a steam generator in

. order to maintain proper water chemistry.

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 d rains.

1.3 GASEOUS SOURCE TERMS The following sources are considered in calculating the releases of radioactive materials (noble gases, radioactive particulates, carbor.-14, tritium, argon-41, and iodine) in gaseous effluents from normal operation, including anticipated operational occurrences:

1. Waste gas processing system;

2. Stean 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; and

6. Steam leakage from the secondary system.

The releases of radioactive materials in gaseous effluents from the followi and 10~ng sources C1/yr are calculated of iodine-131. Therefore, to be less thethan 1 Ci/yrreleases following of noble~are gases considered negligible:

1. Steam releases due to steam dumps to the atmosphere and low-power physics testing and

2. - Ventilation air from buildings not covered in 5. above.

The calculational model considers inputs to the waste gas processing

. system' from both continuous stripping of the primary cool::at during normal operation and from degassing the primary coolant for two cold shutdowns per year. For plants equipped with steam generator blowdown systems, theLmodel considers iodine present in gases leaving the system 1-6

- -vent. .The PWR-GALE Code calculates the releas'e. rates'of noble gases and-iodine to building atmospheres' based on coolant leakage rates to buildings.

Radiciodine releases are related to the iodine-131 coolant concentrations for-the PWR being evaluated. . Particulate release rates are based on

~

measurements at operating PWR's. -

Chapter 2 provides iodine and particulate decontamination factors for removal equipment and parameters for calculating holdup times for noble gases and for calculating tritium, argon-41 and carbon-14 releases.

1.4 LIQUID SOURCE TERMS The following sources are considered in calculating the. release of radioactive materials in liquid effluents from normal operation, including anticipated operational occurrences:

1. Processed water generated from the baron recovery system to maintain plant water balance or for tritium control;

2. . . Processed liquid waste discharged from the dirty waste or

~

miscellaneous waste systems;

3. Processed liquid waste discharged from the steam generator blowdown treatment system;

4. Processed liquid waste discharged from the chemical' waste and condensate demineralizer regeneration system;

5. Liquid waste discharged from the turbine building floor drain sumps; and

. 6. - Detergent waste.

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

!PCA is based on the recommendations of the American National Standard (ANSI N237) Source Term Specification (Ref.1), with the changes as noted in Section 1.2 under the Primary Coolant definition.

Radionuclide removal by the liquid radwaste treatment system is

, based on the following parameters:

1. Decay during collection and processing and

2. Removal by the proposed treatment systems, e.g., filtration, ion exchange, evaporation,- reverse osmosis, and plateout.

For hWR's using a deep-b'ed 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 1-7

is processed through the demineralizers and radionuclide removal using the decontamination factors given in Chapter 2. The activity on the condensate demineralizer resins will also include the steam generator blowdown activity if the blowdown is recycled to the condensate demin-eralizers. The radioactivity content of the demineralizer regenerant solution is obtained by considering that all the radioactivity is removed from the resins at the interval dictated by the regeneration frequency.

Methods for calculating collection and processing times and the decontamination factors for radwaste treatment equipment are given in this chapter. The liquid radioactive source terms are adjusted to compensate for equipment downtime and anticipated operational occurrences.

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

1.5 INSTRUCTIONS FOR COMPLETING PWR-GALE CODE INPUT DATA CARDS 1.5.1 PARAMETERS INCLUDED IN THE PWR-GALE CODE The parameters listed below are built into the PWR-GALE Code since they are generally applicable to all PWR source term calculations and do not require entry on input data cards.

1.5.1.1 The Plant Capacity Factor 0.80(292 effective full power days per year).

1. 5.1. 2 Radionuclide Concentrations in the Primary Coolant, Secondary Coolant, and Main Steam See Section 2.2.3 of Cnapter 2 of this document.

1.5.1.3 Radiciodine Releases from Building Ventilation Systems Prior to Treatment See Table 1-1. For a discussion of the normalization techniques see Section 2.2.4.

1. 5.1. 4 Radioactive Particulate Releases from Building Ventilation Systems Prior to Treatment See Table 1-2.

1.5.1.5 Noble Gas Releases from Building Ventilation Systems Noble Gas Releases from the containment building are based on a leakage rate of 3%/ day of primary coolant noble gas inventory. Releases from the auxiliary building are based on 160 lb/ day primary coolant leakage. Releases from the turbine building are based on 1700 lb/hr steam leakage.

1-8

TABLE 1-l i*

RADIOI0 DINE RELEASES FROM BUILDING VENTILATION SYSTEMS PRIOR TO TREATMENT (C1/yr/pC1/g)

Containment Auxiliary Turbine Building Building ** Building ***

Annual Normalized

• Iodine Release Rate 3

Power Operation 8.0 x 10-4ti 0.72 i 3.8 x 10 2

Refueling / Maintenance 0.32** 2.59 4.2 x 10 Outages t* The values in this table come from Tables 2-13 through 2-16.

• The normalized release rate, during different modes of operation, represents the effective leak rate for radiciodine. It is the combination of the reactor water leakage rate into the building and }.%j 4 ~

the partitioning of the radiciodine between the water phase in the 1 leakage and the gas phase where it is measured. For the turbine

-";,:ct; building the effective leak rate must consider the carryover for radiciodine from water to steam in the steam generator. ] q(

.s .

• • To obtain the actual iodine release from these buildings in Ci/yr, ET' multiply the normalized release by the iodine coolant concentration @ '* .

in pCi/g.

($.k;fd

• • • To obtain the actual iodine release from the turbine building in Ci/yr, multiply the normalized release by the secondary coolant hQ; f.Le; concentration in pCi/g and by the partition coefficient (NS) from [ .l Table 2-6. g t Includes contribution from the fuel pool area. Er : .

q tt This release rate is expressed in %/ day of leakage of primary ,

m coolant inventory of iodine and represents the effective leak  ? C~

rate for radiciodine. It is the combination of the reactor water jf.3,; .

leakage rate into the buildings, and the partitioning of the .g1 radioiodine between the water phase in the leakage and the gas g.M.

phase where it is measured. In order to obtain the releases in h?y curies / year during power operations from the containment building d ,e.;.

of a particular PWR, the normalized leak rates in Table 1-1, are 70i multiplied in the PWR-GALE Code by the iodine concentration in the 16 W reactor coolant for that particular PWR, and then this leak rate >

1s considered along with the containment purging method for that "%"'J ,

particular PWR. Tb l-9 II  ;

M(., 9

1 TABLE l-2 RADI0 ACTIVE PARTICULATE RELEASES FROM BUILDING VENTILATION SYSTEMS PRIOR TO TREATMENT *

(Ci/yr)/ Unit Auxiliary Fuel Pool Waste Gas Nuclide Containment Building Area System C r-51 9.2(-3)i 3.2(-4) 1.8(-4) 1.4(-5)

Mn-54 5.3(-3) 7.8(-5) 3.0(-4) 2.1 (-6)

Co-57 8.2(-4) NA NA NA Co-58 2.5(-2) 1.9(-3) 2.1(-2) 8.7(-6)

Co-60 2.6(-3) 5.1(-4) 8.2(-3) 1.4(-5)

Fe-59 2.7(-3) 5.0(-5) NA 1.8(-6)

Sr-89 1.3(-2) 7.5(-4) 2.1 (- 3) 4.4(-5)

Sr-90 5.2(-3) 2.9(-4) 8.0(-4) 1.7(-5)

Zr-95 NA 1.0(-3) 3.6(-6) 4.8(-6)

Nb-95 1.8(-3) 3.0(-5) 2.4(-3) 3.7(-6)

Ru-103 1.6(-3) 2.3(-5) 3.8(-5) 3.2(-6)

Ru-106 NA 6.0(-6) 6.9(-5) 2.7(-6)

Sb-125 NA 3.9(-6) 5.7(-5) NA Cs-134 2.5(-3) 5.4(-4) 1.7(-3) 3.3(-5)

Cs-136 3.2(-3) 4.8(-5) NA 5.3(-6)

Cs-137 5.5(-3) 7.2(-4) 2.7(-3) 7.7(-5)

Ba-140 NA 4.0(-4) NA 2.3(-5)

Ce-141 1.3(-3) 2.6(-5) 4.4(-7) 2.2(-6)

NA - No release observed from this source. Release assumed to be less than 1.01, of total, t 9.2(-3) = 9.2 x 10-3 ,

• The values in this table come from Tables 2-17 through 2-21.

1-10

E 1.5.1.6 Containment Building Purge Frequency Two purges at cold shutdown per year plus a continuous purge specified

-by the applicant in his containment design.

1.5.1.7 Primary System Volumes Degassed per Year Two coolant volumes per year for cold . shutdowns plus volumes degassed due to continuous stripping.

1.5.1.8 Steam Generator Partition Coefficient (PC)

Once-through PC Iodine 1.0 Nonvolatiles 1.0 Recirculation U-Tube Iodine 0.01 Nonvolatiles 0.005 1.5.1.9 Radiciodine Releases from the Main Condenser Air Ejector Exhaust Prior to Treatment The normalized release rate of radioiodine from the main condenser air ejector exhaust prior to treatment is 1.7 x 103 Ci/yr/pCi/g. The normalized release rate represents the effective release rate for radio-iodine. It is the combination of the steam flow to the main condenser, the partitioning of radiciodine between the main condenser and the air

. ejector exhaust where it is measured, and the partition coefficient for radiciodine from water to steam in the steam generator. To obtain the actual iocine release from the main condenser air ejector exhaust in

- Ci/yr, multiply the normalized release by the secondary coolant concen-tration in pCi/g and by the iodine partition coefficient (NS) from Table 2-6.

1.5.1.10 Containment Internal Cleanup System For systems using an internal cleanup system, the PWR-GALE Code calculates the iodine concentration in the containm4nt atmosphere based on 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> of system operation prior to purging, an iodine removal efficiency for the charcoal adsorbers corresponding to Table 1-5, a particulate DF of 100 for HEPA filters and an internal mixing efficiency of 7 0%.

1.5.1.11 Detergent Wastes The radionuclides listed in Table 2-27 of Chapter 2 are assumed to be released unless treatment is provided or laundry is not processed on site.

1-11

1.5.1.12 Tritium Releases The tritium releases through the combined liquid and vapor pathways are 0.4 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 1.0 pCi/ml up to a maximum of 0.9 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.

1.5.1.13 Argon-41 Releases The annual quantity of argon-41 released from a pressurized water reactor is 34 Ci/yr. The argon-41 is released to the environment via the containment vent when the containment is vented or purged.

1.5.1.14 Carbon-14 Releases The annual quantity of carbon-14 released is 7.3 Ci/yr, of which the releases from the containment, auxiliary building and waste gas system are 1.6, 4.5 and 1.2 C1/yr, respectively.

1.5.1.15 Decontamination Factors for Condensate Demineralizer Demineralizer Anion Cs, Rb Other Nuclides Deep Bed 10 2 10 Powdex 10 2 10 Note: For a system using filter /demineralizers (Powdex), a zero is entered for a regeneration frequency as explained later in Section 1.5.2.10.

1.5.1.16 Primary Coolant Purification System Demineralizers Demineralizer Anion Cs, Rb Other Nuclides Mixed Bed 100 2 50 Cation 1 10 10 1.5.1.17 Releases of Radioactive Material in Liquid Waste from the Turbine Building Floor Drain System 7200 gal / day at main steam activity.

1.5.1.18 Regeneration of Condensate Demine.alizers Flow rates and concentrations of radioactive materials routed to the liquid radwaste treatment system from the chemical regeneration of the condensate demineralizers are based on the following parameters:

1-12

t

m

1. Liquid flow to the demineralizer is based on the radioactivity

! of the main steam and the fraction of radioactivity which does not bypass the condensate demineralizer if there is pumped i forward flow. The steam generator blowdown radioactivity is

, added to the condensate radioactivity if the blowdown is L processed through the condensate demineralizer.  ; y i

[ 2. All radionuclides removed from the secondary coolant by the l demineralizer resins are removed from the resins during chemical a

regeneration. The radioactivity in the regenerant wastes is adjusted for radionuclide decay during demineralizer operation.

) 1.5.1.19 Adjustment to Liquid Radwaste Source Terms for Anticipated 3 Operational Occurrences

1. The calculated source term is increased by 0.16 Ci/yr per reactor using the same isotopic distribution as for the calculated source term to account for anticipated occurrences such as operator errors resulting in unplanned releases.

2. Evaporators are assumed to be unavailable for two consecutive

days per week for maintenance. If a two-day holdup capacity or i an alternate evaporator is available, no adjustment is needed.

If less than a two-day capacity is available, the waste excess is assumed to be handled as follows:

a. Clean or Dirty Waste - Processed through an alternative

'_' system (if available) using a discharge fraction consistent with the lower purity system,

b. Chemical Waste - Discharged to the environment to the 4 extent holdup capacity or an alternative evaporator is I not available.

1.5.2 PARAMETERS REQUIRED FOR THE PWR-GALE CODE Complete the cards designated in the sections below by "(SAR/ER)"

from information given in the Safety Analysis and Environmental Reports.

Complete the remaining cards (i.e., those not designated below as "(SAR/ER)"

j cards), using the principal source term parameters specified below and j3

[ discussed in Chapter 2 of this document.

1.5.2.1 Card 1: Name of Reactor (SI,R/ER) 7 Enter in spaces 33-60 the name of the reactor.

L- Enter in spaces 78-80 the type of reactor, i .e. , PWR.

1.5.2.2 Card 2: Thermal Power Level (SAR/ER)

Enter in spaces 73-80 the maximum thermal power level (in MWt) evaluated for safety considerations in the Safety Analysis Report.

1-13

1.5.2.3 Card 3: Mass of Coolant in Primary System (SAR/ER) 3 Enter in spaces 73-80 the mass of coolant (in 10 lb) in the primary l system at operating temperature and pressure.

i 1.5.2.4 Card 4: Primary System Letdown Rate (SAR/ER)

Enter in spaces 73-80 the average letdown rate (gal / min) from the primary system to the purification demineralizers.

1.5.2.5 Card 5: Letdown Cation Demineralizer Flow Rate (SAR/ER)

Enter in spaces 73-80 the annual average flow rate (gal / min) 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 on Card 4) by the fraction of time the cation demin-eralizers are in service to obtain the average cation demineralizer flow rate.

1.5.2.6 Card 6: Number of Steam Generators (SAR/ER)

Enter in spaces 73-80 the number of steam generators.

1.5.2.7 Card 7: Total Steam Flow (SAR/ER)

Enter in spaces 73-80 the total steam flow (in 610 lb/hr) for all steam generators. ,

1.5.2.8 Card 8: Mass of Liquid in Each Steam Generator (SAR/ER)

Enter in spaces 73-80 the mass of liquid (in 103 lb) in each steam generator.

1.5.2.9 Card 9: Steam Generator Blowdown Rate and Blowdown Treatment Method (SAR/ER)

Enter in spaces 37-44 the steam generator blowdown rate as given in the applicants SAR or ER.

Enter total blowdown rate in thousands of Ib/hr in spaces 37-44.

For a once-through steam generator, leave spaces 37-44 blank.

Describe the Blowdown Treatment Method as follows:

1. Enter 0 in space 80 if the blowdown is recycled to the condensate system af ter treatment in the blowdown system whether or not there are condensate demineralizers.

2. Enter 1 in space 80 if the steam generator blowdown is recycled directly to condensate system demineralizers without prior treatment in the blowdown system.

1-14

i-I h

L 3. Enter 2 in space 80 if the steam generator blowdown is not L recycled to the condensate system.

l?

If the plant has once-through steam generators, leave space 80 blank.

1.5.2.10 Card 10: Condensate Demineralizer Regeneration Time For deep-bed condensate demineralizers which do not use ultrasonic resin cleaner, use a 1.2-day regeneration frequency. Multiply the f requency by the number of demineralizers and enter the calculated number of days in spaces 73-80; for deep-bed condensate demineralizers which use ultrasonic resin cleaning, use an 8-day regeneration frequency. -

For filter /demineralizers (Powdex) or if condensate demineralizers are not used, enter zeros in spaces 73-80. -

1.5.2.11 Card 11: Fraction of Feedwater Through Condensate Demineralizer (SAR/ER)

E Enter in spaces 73-80 the fraction of feedwater to the steam generator processed through the condensate demineralizers. If condensate demineralizers are not used, enter 0.0 in spaces 73-80.

1.5.2.12 Cards 12-29: Liquid Radwaste Treatment System Input Parameters Six liquid radwaste inlet streams are considered in the PWR-GALE Code:

[ 1. Shim Bleed, Cards 12-14.

F

2. Equipment Drain Waste, Cards 15-17.

[

(

3. Clean Waste, Cards 18-20.

4. Dirty Waste, Cards 21-23,

5. Blowdown Waste, Cards 24-26.

6. Regenerant Wastes, Cards 27-29.

Three input data cards are used to define the major parameters for each of the six waste streams. Essentially the same information is needed on the three input data cards used for each of the six waste streams. The instructions given in this section are applicable to all six waste streams with the following exception: The inlet waste activity is not entered for Cards 12, 24, and 27 for the shim bleed, blowdown wastes, or regenerant ,

wastes since that activity for these wastes is calculated by the PWR-GALE Code.

Cards 12-14 are used only for the shim bleed stream. For reactor designs that combine the shim bleed with other reactor grade wastes prior to processing, the other wastes are entered as equipment drain wastes on Cards 15-17.

1-15

W The entries required on the first card (12, -15,18, 21, 24, and 27) for each of the six waste streams, respectively, considered in the PWR-GALE Code are outlined below and described in more detail in Section 1.5.2.15.1.

1. Enter 'in spaces 17-39 the name of the waste stream (Card 24 spaces:17-44).

2. Enter in spaces'42-49 the flow rate (in gal / day) of the inlet stream (except on Cards 24 and 27).

~3. Enter in spaces 57-61 the activity of the inlet stream expressed as a fraction of primary coolant activity (PCA) 1(except on Cards 12, 24 and 27).

The second card (13,16,19, 22, 25, and 28) for each waste stream contains the overall system decontamination factors for the three categories of radionuclides, .as follows:

1. Enter in spaces 21-28 the DF for iodine.

2. Enter in spaces 34-41 the DF for cesium and rubidium.

3. Enter in spaces 47-54 the DF for other nuclides.

The following entries are. required on the third card (14,17, 20, 23, 26, and 29) for each waste stream:

1. - Enter!in spaces 28-33 waste collection time (in days) prior to processing.

2. Enter in spaces 48-53 waste processing and discharge times (in days).

3. Enter in spaces 72-77 the average fraction of wastes to be discharged after processing.

Cards 24-26 are for waste inputs due to steam generator blowdown.

1. Card 24

a. For recirculating U-tube steam generator systems, enter the fraction of the blowdown stream processed in spaces 73-80. The PWR-GALE Code will calculate releases based on steam generator blowdown wastes.

b. For once-through steam generator systems, leave spaces 73-80 blank.

2. Card 25

a. If the steam generator blowdown is not recycled to the condensate system, enter blowdown system DF's as explained for Card 13.

1-16

=: , ,

3

?4f

.b. If the steam generator blowdown is recycled directly to the condensate system demineralizers without prior treatment L- in the blowdown system, enter DF of 1.0 for iodine in spaces 21-28, DF of 1.0.for cesium and rubidium in spaces 34-41, and DF of 1.0 for other nuclides in spaces 47-54.

c. If.the steam generator blowdown is recycled to the condensate

' system demineralizers after treatment in the blowdown system, enter blowdown system DF's as explained for Card 13.

3. Card 26.

Complete Card 26 as explained for Card 14.

Cards 27-29 are for waste inputs due to regenerant wastes.

1. Card 27

a. For recirculating U-tube steam generator systems that do not utilize condensate demineralizers in the secondary system, leave spaces 73-80 blank.

b. For once-through steam generator systems and for recirculating U-tube steam generator systems that utilize condensate demineralizers in the. secondary system, enter the regenera-tion solution waste flow (gal / day) in spaces 73-80. The inlet waste activity is not needed since the activity is calculated by the PWR-GALE Code.

.2. Cards 28 and 29 Complete Cards 28 and 29 as explained for Cards 13 and 14.

The following sections explain in more detail the use of.the parameters in this report and the information given in the SAR/ER to make the data entries on Cards 12-92 listed above.

1.5.2.12.1 Liquid Waste Flow Rates and Activities (Cards 12, 15, 18, 21, 24 and-27)

Flow rates and activity are calculated, using the waste volumes and activities given in Table 1-3. To the input flow rates given in the table, add expected flows and activities more specific to the plant design as given.in the SAR/ER. With the exception of the shim bleed, the individual streams are combined based on the radwaste treatment system described

, in the SAR/ER.

Waste streams processed with the shim bleed are entered as equipment drain wastes on Cards 15-17. Input activities are based on the weighted average activity of the composite stream entering the waste collection 1-17

- ~

. . y' , ig s

• ~-

, e s.

c , -

J t

t',

k

/

-TABLE l_3- . ..

PWR LIQUID WASTES

. EXPECTED DAILY AVERAGE INPUT  ! FLOW RATEL(in Gal / day) '

. Type of treatstent. cf blowdown recycled ,to secondary .i

. system (U-tube steam generator plants) or type of. -

treatment of condensate (once-through steam generatorplants) . Plant with!

blowdown treat--

Deep-bed cond. '

ment. Product i

Deep-bed cond. .demineralizers not recycled to-demineralizers without condenser or. ~ . FRACTION OF

.with ultrasonic. ultrasonic Filter- .seconda ry ' coolant ' PRIMARY. COOLANT SOURCE resin cleaner resin cleaner demineralizer. ~ system- ~ ACTIVITY (PCA)

l. . REACTOR CONTAIMENT-4 -

'20 0.1 f I a. Primary coolant pump seal lekage

20 20 20

b. Primary coolant leakage, 10 10 10 10 .1.67*

miscellaneous sources

c. Primary coolant equipment -.500 500 500 500 0.0 01 d rains .

1 I <

2. PRIMARY COOLANT SYSTEMS (OUTSIDE OF CONTAIMENT)

a. Primary coolant system 80 80 80 80- 1.0 i equipment drains

b. Spent fuel pit liner drains 700 700 700 700 0.0 01

c. Primary coolant sampling 200 200 200 200. 0.05 system drairs

d. _ Auxiliary building floor 200 200 200 200 0.1 drains -

1 1

r m .. w,.-

~

yi a

. TABLE.1-3'(Continued)- ,,

. 3. SECONDARY COOLANT SYSTEMS -

'a. Secondary coolant sampling 1400 1400- 1400- Ll400 10-4 system drains

b. Condensate demineralizer 3000 12000- -- -- 10

-8 ,

rinse and transfer solutions

c. Condensate demineralizer 850 3400 - -

Calculated in regenerant solutions . GALE Code-

d. Ultrasonic resin cleaner 15000 - - - 10-6, solutions

e. Condensate filter- - -

8100 -

- 2 x 10-0 demineralizer backwash

f. Steam generator blowdown - - '- Plant dependent ** Plant dependent **

7 g. Turbine building floor 7200 7200 7200 7200 Calculated in G drains GALE Code

4. DETERGENT AND DECONTAMINATION SYSTEMS *

a. On-site laundry facility 300 300 300 300 See Table'2-26'

b. Hot showers Negligible Negligible Negligible Negligible -

c. Hand wash sink drains 200 200 200 200 See Table 2-26

d. Equipment and area 40 40 40 40 See Table 2-26 decontamination TOTALS 29,700 26,300 19,000 10,000

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

• • Input parameter.

m ,_

tanks. . For example, if the inlet streams A, B, and C enter the dirty waste collector tank at average rates and PCA as listed below, Stream A 1,000 gal / day at 0.0lPCA Stream B 2,000 gal / day at 0.lPCA Stream C 500 gal / day at 1.0PCA the composite A, B, C activity would be calculated as follows:

(1,000 gal / day)(0.0lPCA) + (2,000 gal / day)(0.lPCA) + (500 gal / day)(1.0PCA) = 0.2PCA (1,000 gal / day + 2,000 gal / day + 500 gal / day)

The entries on Card 21 for this example would then be: spaces 17-33,

" Dirty Waste"; spaces 42-49, 3500.; spaces 57-61, " 0. 2" .

The input flow rates and activities are entered in units of gal / day and fractions. of PCA, respectively.

1.5.2.12.2 Decontamination Factors for Equipment Used in the Liquid Radwaste Treatment System (Cards 13,16,19, 22, 25, and 28)

.The decontamination factors (DF's) given in this' document are used in the PWR-GALE Code. The DF's represent the expected equipment performance averaged over 'the life of the plant, including downtime. The following-factors should be considered in calculating the overall decontamination factors for the various systems:

, 1. DF's are categorized by one of the following types of radionuclides:

a. Halogens

b. Cs, Rb

c. 'Other Nuclides Note: A DF of.1 is assumed by the PWR-GALE Code for tritium. Noble gases and water activation products, e.g., N-16, are not considered in the liquid code.

2. The system DF for each inlet stream is the product of the '

individual equipment 0F's in each of the subsystems.

3. Equipment that is used optionally (as required) and not included in the normal flow scheme should not be considered ,

in calculating the overall system DF. 1 Table 1-4 shows the decontamination factors to be used for PWR systems.

L 1-P0 e __

f_ t TABLE l-4 DECONTAMINATION FACTORS FOR PWR LIQUID WASTE TREATMENT SYSTEMS r..

TREATMENT SYSTEM DECONTAMINATION FACTOR Demineralizer Anion Cs, Rb Other Nuclides

< Mixed Bed-

~

Primary coolant letdown (CVCS) 100 2 50 2 2 Radwaste (H+0H-) 10 (10)* 2(l'0) 10 (10)

Evaporator condensate polishing' 5 'l 10 Boron recycle 10' 2 10 2 2 Steam generator blowdown 10 (10) 10(10) 10 (10)

Cation bed (any system) 1 (1 ) ' 10(10) 10(10) 2 Anion bed (any system) 10 (10) 1(1) 1(1)

Powdex (any system) 10(10) 2(10) 10(10)

All Nuclides Evaporators Except Iodine Iodine 3 2 Miscellaneous radwaste 10 10 3 2 Boric acid recovery 10 10 Reverse Osmosis All Nuclides Laundry wastes 30 Other liquid wastes 10 Filters DF of I for all nuclides o

• For demineralizers in series, the DF for the second demineralizer is l-- -given'in parentheses.

L i

! 1 - 21 L

^

The following example illustrates the calculation of the decontamination factor for a dirty waste treatment system: Assume that dirty wastes are collected; processed through a filter, an evaporator, and a mixed-bed polishing demineralizer; and collected for sampling. If required to meet discharge criteria, the contents of the waste sample (test) tank are processed through a mixed-bed demineralizer for additional radionuclide removal. This example may be summarized graphically as:

Demineralizer 2 Dirty waste - Filter - Evaporator - Demineralizer 1 - Waste sample collector tank tank Extracting from Table 1-4 gives the following values for the example:

Demineralizer Demineralizer Filter Evaporator 1 2 Product 2 2 Iodine 1 10 5 1 5 x 10 3 3 Cs, Rb 1 10 1 1 10 3 4 Other Nuclides 1 10 10 1 10 These values are obtained as follows:

e A DF of 1.0 is applied to all nuclides for the filter.

2 3 e A DF of 10 for iodine and 10 for Cs, Rb, and other nuclides is applied for the radwaste evaporator, e A DF of 5 is applied for iodine, a DF of 1 for Cs, Rb and a

. DF of 10 for the evaporator condensate polishing demineralizer, e A DF of 1 is applied to the second demineralizer since this demineralizer's used is optional, and it is not used for normal operations.

e The product of the DF's is obtained by multiplication of the first four columns for each nuclide.

Thus on Card 22, the following would be entered: in spaces 21-28, "500.0"; in spaces 34-41, "1000.0"; and in spaces 47-54, "10000.0".

1.5.2.12.3 Collection Time for Liquid Wastes (Cards 14, 17, 20, 23, 26, and 29 - ' Spaces 29-33)

Collection time prior to processing is based on the input flow

, calculated above. Where redundant tanks are provided, assume the collection l tank to be filled to 80% design capacity. If only one tank is provided, I

j 1-22 i

assume the tank to be filled to 40% design capacity. For example, if flow from a 1,000-gal / day floor drain is collected in two 20,000-gallon tanks prior to processing, collection time would be calculated as follows:

Collection time (Tc ) " = 16 days l 9 /d Then, for example, "16.0" should be entered in spaces 29-33 on Card 23.

1.5.2.12.4 Processing and Discharge Time (Cards 14, 17, 20, 23, 26, and 29 -- Spaces 48-53)

Decay during processing and discharge of liquid wastes is shown graphically as follows:

o Tank A - BR b - Tank C -

R - Discharge Canal c

where A is the capacity of initial tank in flow scheme, in gal; B is the limiting process based on equipment flow capacity, dimensionless; C is the capacity of final tank in flow scheme prior to discharge, in gal; R is the equipment flow capacity of Process B, in gal / day; b

R is the flow capacity of Tank C discharge pump, in gal / day; and c

R is the rate of flow of additional waste inputs to Tank C, in g

gal / day.

Tp , the process time credited for decay, is calculated as follows, in days:

0.8A for redundant tanks, or T =

0 4A for a single tank T =

p p Td , the discharge time (50% credited for decay), is calculated as follows, in days:

T d" R for redundant tanks, or Td= j for a single tank.

C C l-23

After performing the above two calculations, calculate whether credit may be taken for decay during discharge by determining whether 0.8C > Tp (Rb + R,) for redundant tanks, or 0.4C > Tp (Rb + 9R ) for a single tank.

-If so, then Decay = T + 0.5T p d where " Decay" is the new processing and discharge time to be entered in spaces 48-53 of the third card for each input stream (Cards 14, 17, 20 23, 26, and 29).

If, however, 0.8C (or 0.4C, as appropriate) < Tp(Rb + Rg ), T p is used for the holdup time during processing, since Tank C may be discharged before Tank A has been completely processed. In this case, the T value should be p

entered in spaces 48-53 of the third card.

For example, for the following input waste stream:

FLOOR DRAINS 1.000 GAL / DAY I

t t FLOOR DRAIN FLOOR DR AIN DAY 540

. TANK A TANK B 20.000 GAL 20.000 GAL WASTE

• SAMPLE TANK A 40,000 GAL g .

• ~ VAP RATOR 00 AL/

MIN 1s GAL / MIN

, WASTE

• SAMPLE TANK 8' '

40.000 GAL DISCHARGE PUMP 10 GAL /

MIN.

1-24

i

Decay time during processing and discharge is calculated as follows:

Process Time (Tp ) = = 0.7 day (15 imin 0 / day)

Discharge Time (T d ) " (10 imi 0 / day) = 2.2 days 1

Then, checking for decay credit, 0.8C/(Rb + Rg) = 1.45 days, which is greater than Tp ; therefore, credit is taken for (Tp+ 0.5T d) or 1.8 days for processing and discharge. The input in spaces 48-53 to the Code is 1.8 days for processing and discharge time.

1.5.2.12.5 Fraction'of Wastes Discharged (Cards 14, 17, 20, 23, 26, and 29 -- Spaces 72-77)

The percent of the wastes discharged after processing may vary between 10% and 100%, except as noted below, based on the capability of the system to process liquid waste during equipment downtime, waste volume surges, tritium control requirements, and tank surge capacity. A minimum value of 10% discharge for the liquid radioactive waste treatment system ,

is used when the system is designed for maximum waste recycle, when the system capacity is sufficient to process wastes for reuse during equipment downtime and anticipated operational occurrences, and when a discharge route is provided. For steam generator blowdown treatment systems, less than 10% discharge should be considered on a case-by-base basis, depending on system capacity.

The PWR-GALE Code calculates the release of radioactive materials in liquid waste from the following systems after processing. The quantity released is shown on the printout.

1. Boron Recovery System - Combined releases from both shim bleed and equipment drains.

2. Miscellaneous Liquid Waste System - Combined releases from both clean and dirty waste subsystems.

3. Secondary Waste System - Releases from steam generator blowdown system, regenerant wastes from demineralizer regenerations, or both according to the plant design.

'4. Turbine Building Floor Drain System - Releases of liquid from l the turbine building floor drain system are calculated assuming no treatment prior to release. Straight decay time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is built into the code.

5. Detergent Waste System - Combined releases from laundry operations, equipment decontamination solutions, and personnel decontamination showers.

1-25

1.5.2.13 Card 30: Letdown System

1. Enter 0 in space 80 if there is not continuous gas stripping of the full letdown flow. (This sets Y = 0.0.) i j

2. Enter 1 in space 80 if there is continuous degassification ,

of the full letdown flow to the gaseous radwaste system via a gas stripper. (This sets Y = 1.0.)

3. Enter 2 in space 80 if there is continuous purging of the volume control tank. (This sets Y = 0.25.)

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, caver gas) is calculated in the PWR-GALE Code. (For definition of "Y", she Tables 2-4 & 2-5.)

1.5.2.14 Cards 31-33: Holdup Time for Fission Gases Stripped from Primary Coolant 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 parameters:

1. Pressurized Storage Tanks

a. One storage tank is held in reserve for back-to-back shutdowns, one tank is in the process of filling, and the remainder are used for storage. The PWR-GALE Code will calculate the effective holdup time for filling and add it to the holdup time for storage.

b. Calculations are based on the waste gas input flow rate to the pressurized storage tanks, and a storage tank pressure 70% of the design value.

c. If the calculated holdup time exceeds 90 aays, assume the remaining gases are released after 90 days.

The holdup time (T h

) and fill time (T f) are calculated as follows:

PV f"T T PV(n-2) h= F

-uhere n is the number of tanks; 1-26

n-2 is the correction to subtract the tank being filled and the tank held in reserve; P -is the storage pressure, in atmospheres (dimensionless in this partic.ularcalculation);

T is the time required to fill one tank, in days; f

T is the holdup time, in days; h

V is the volume of each tank, in ft (STP);and F 'is the waste gas flow rate to pressurized storage tanks. This flow rate should be supplied by the applicant for the specific type of waste gas system design. In the absence of specific data supplied by the applicant, we will use the data given in Section 2.2.12.1, in which the average value for_ the PWR's listed in Table 2-24 is 170 ft3 / day (STP) per reactor for-PWR's without recombiners; and for PWR's with recombiners, the average

'value for the PWR's listed in Table 2-25 is 30 ft3/ day (STP) per reactor.

Enter on Card 31 the holdup time, in days, for Xe in spaces 73-80.

Enter on Card 32 the holdup time, in days, for Kr in spaces 73-80.

Enter on' Card 33 the fill time, in days, in spaces 73-80.

2. Charcoal Delay Systems Charcoal delay system holdup times are based on the following equation:

T = 0.011 MK/F where F is the system flow rate, in ft / 3min; (see 1.5.2.14.1.c, above)

K ' is the dynamic adsorption coefficient, in cm3 jg; M is the mass of charcoal adsorber, in thousands of pounds; and T is the holdup time, in days.

The dynamic absorption coefficient, K, for Xe and Kr and based on 1-27

the system design noted below.

3 DYNAMIC ABSORPTION COEFFICIENT, K (cm fg)

Operating 77 F Operating 77*F Operating 77*F Operating 0*F Dew Point 45*F Dew Point 0*F Dew' Point -40* Dew Point -20*

Kr 18.5 25 70 105 Xe 330.0 440, 1160 2410 Enter on Card 31 the holdup time, in days, for Xe in spaces 73-80.

Enter on Card 32 the holdup time, in days, for Kr in spaces 73-80.

Leave Card 33 blank.

13. Cover Gas Recycle System For this' system or other systems designed to' hold gases indefinitely, the calculations are based on a 90-day holdup time.

Enter on Card 31 the holdup time (90 days) for Xe in spaces 73-80.

Enter on _ Card 32 the holdup time (90 days) for Kr in spaces 73-80.

Enter on Car'd 33 the fill-time (0 days) in spaces 73-80.

1.5.2.15 Card 34: Waste Gas System Particulate Releases Card 34 identifies the treatment provided for particulate removal from the waste gas system effluent.

1. If ventilation exhaust air is treated through HEPA filters which satisfy the guidelines of Regulatory Guide 1.140 (Ref. 2), enter a removal efficiency of 99. for particulates in spaces 39-41..

2. If no treatment is provided for the ventilation exhaust air to i remove particulates or if the HEPA filters do not satisfy the l

guidelines of Regulatory Guide 1.140 (Ref. 2), enter 0.0 in '

spaces 39-41.

1.5.2.16 Cards 35 and 36: Fuel Handling and Auxiliary Buildings Releases Cards 35 and 36 indicate the fractions of airborne iodine and radio-active particulates released from the fuel handling and auxiliary buildings, respectively.

l l-28

r-TABLE l-5

~

ASSIGNED REMOVAL EFFICIENCIES FOR CHARC0AL ADSORBERS FOR RADIOI0 DINE REMOVAL b

Removal Efficiencies

- ActisatedCarbonaBed Depth for Radioiodine %

. 2. inches. Air filtration system designed to 90.

operate inside reactor containment 2 inches. Air filtration system designed to 70.

operate outside the reactor containment and relative humidity is controlled at 70%

4 ' inches. - Air filtration system designed to 90.

operate outside the reactor containment and relative humidity is controlled at 70%

6 inches. Air filtration system designed to 99.

operate outside the reactor containment and

-relative humidity is controlled to 70%

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

b The removal efficiencies assigned'to HEPA filters for particulate removal and charcoal adsorbers for radiciodine removal are based on the design, testing, and maintenance criteria recommended in Regulatory Guide 1.140, " Design, Testing and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear _ Power Plants" (Ref. 2).

t f

h 1-29

l. If vIntilation exhaust air is traat d through charcoal adsorbers which satisfy the guidelines of Regulatory Guide 1.140 (Ref. 2),

enter the appropriate removal efficiency in spaces 47-49 for radioiodine corresponding to the depth of charcoal'as indicated.

in Table 1-5.

2. If ventilation exhaust air is treated through HEPA filters which satisfy the guidelines of Regulatory Guide 1.140 (Ref. 2), enter a removal efficiency of 99, for particulates in spaces 56-58.

3. If no treatment is provided for the ventilation exhaust air to remove radioiodine, enter 0.0 in spaces 47-49; if no treatment

.is provided.to remove particulates, enter 0.0 in spaces 56-58.

1.5.2.17 Card 37: Containment Free Volume (SAR/ER) 6 Enter the containment volume (in 10 ft 3) in spaces 73-80.

1.5.2.18 Card 38: Containment Internal Cleanup System (SAR/ER)

If the containment internal cleanup system uses charcoal adsorbers

~

1. -

which satisfy the guidelines of Regulatory Guide 1.140 (Ref. 2),

enter the appropriate removal efficiency in spaces 47-49 for radioiodine corresponding to the depth of charcoal as indicated in Table 1-5.

2. If the containment internal cleanup system uses HEPA filters which satisfy the guidelines of Regulatory Guide 1.140 (Ref. 2), enter a removal efficiency of 99. for particulates in spaces 56-58.

3. If there is no containment internal cleanup system, enter 0.0 in spaces' 47-49 and in spaces 56-58.

3

4. Enter the' flow rate (in 10 ft/3 min)throughtheinternalcleanup system in spaces 73-80.

The airborne concentration calculations are based on the following parameters:

a. A primary coolant leakage rate corresponding to the normalized release rate given in Table 1-1.

b. A continuous normal ventilation flow rate as specified by the applicant.

c. Operation of the cleanup system for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> prior to purging.

d. A DF for the charcoal adsorber corresponding to the values in Table 1-5, a DF of 100 for the HEPA filters, and a mixing efficiency of 70%. The mixing efficiency is an effective removal efficiency which takes into account the effects of incomplete mixing in the containment.

1-30

Continuous leakage of primary coolant during the operation of

~

e.

the internal cleanup system. l 1.5.2.19 Card 39: Containment Building Iodine Releases - During large Volume Purge System Operation Card 39 indicates the fraction of airborne iodine and radioactive particulates released during purging of the containment building with the large volume containment purge system.

Note: Treatment referred to below does not- include the internal recirculation system.

1. If ventilation exhaust air is treated through charcoal adsorbers which satisfy the guidelines of Regulatory Guide 1.140 (Ref. 2),

enter the appropriate removal efficiency in spaces 47-49 for radiciodine corresponding to the depth of charcaal as indicated in Table 1-5.

2. If ventilation exhaust air is treated through HEPA filters which satisfy the guidelines of Regulatory Guide 1.140 (Ref. 2), enter a removal efficiency of 99. for particulates in spaces 56-58.

3. If no treatment is provided- for the ventilation exhaust air to remove radioiodine, enter 0.0 in spaces 47-49; if no treatment is provided to remove particulates, . enter 0.0 in spaces 56-58.

4. Enter the number of purges per year during power operations in spaces 78-80. (Note: The 2 purges at shutdown are stored in the PWR GALE Code and need not be entered on card 39.)-

1.5.2.20 Card 40: Containment Building Iodine Releases - Low Volume Purge During Power Operation Card 40 indicates the fraction of airborne iodine in the containment atmosphere that is released during the low volume purge of the containment building while the reactor is at power.

Note: Treatment referred to below does not include the internal recirculation system.

1. If ventilation exhaust air is treated through charcoal adsorbers which satisfy the guidelines of Regulatory Guide 1.140 (Ref. 2),

y enter the appropriate removal efficiency in spaces 47-49 for radioiodine corresponding to the depth of charcoal as indicated in Table 1-5.

2. If ventilation exhaust air is treated through HEPA filters which satisfy the guidelines of Regulatory Guide 1.140 (Ref. 2), enter a removal efficiency of 99.. for particulates in spaces 56-58.

3. If no treatment is provided for the ventilation exhaust air to remove radiciodine, enter 0.0 in spaces 47-49; if no treatment is provided to remove particulates, enter 0.0 in spaces 56-58.

1 - 31

5 I

A

4. Enter the continuous _ containment purge rate (ft 3/ min) in spaces F 73-80..

1.5.2.21 Card 41: Steam Generator Blowdown Tank Vent

1. Enter 0.0 in spaces 73-80 if the gases from the blowdown flash tank are. vented.through a condenser prior to release.

2. Enter 0.0 in spaces 73-80 if the blowdown flash tank is vented to the main condenser air ejector.

. 3. : Enter 0.0 'in spaces .73-80 for a once-through steam generator system.

4. For older plants which still use flash tanks which vent directly

. to_the atmosphere an iodine partition factor of 0.05 is used.

T* .

1.5.2.22- Card 42: Percentage of Iodine Removed by the Condenser Air Ejector Offgas Treatment System

51. If, prior to release, the offgases from the condenser air ejector are processed through charcoal adsorbers which satisfy the guidelines of Regulatory Guide 1.140 (Ref. -2), enter the removal efficiency in spaces 73-80 for radiciodine corresponding to the depth of charcoal as indicated in Table 1-5.

2. If- the _ of fgases' are released from the condenser air ejector without treatment, enter 0.0 in spaces 73-80.

1.5.2.23 Card 43: Detergent Wastes

~

1.1 . If the plant does not have an onsite laundry, enter 0.0 in spaces 73-80.

2. If the plant has an onsite laundry and detergent wastes are released without treatment, enter 1.0 in spaces 73-80.

3. If detergent wastes are treated prior to discharge,-enter the fraction of radionuclides remaining after treatment (1/DF) in spaces 73-80. The parameters in Chapter 2 of this document should be used in determining the DF for the treatment applied i

to detergent wastes.

t 1-32 l-

L CHAPTER 2..' PRINCIPAL PARAMETERS'USED IN PWR SOURCE TERM CALCULATIONS AND THEIR BASES

' 2.1 - INTRODUCTION .

l The? principal' parameters used in source term calculations have been -I

. compiled to standardize-the calculation of radioactive source terms.

r

. The following sections describe parameters used in the evaluation of ~ radwaste treatment systems.x The parameters have been derived from-

' reactor operating experience where data were available.t Where operating

' data:were inconclusive or not available, information was drawn from laboratory. and field tests -and from engineering judgment. The bases _for the; source; term parameters explain the reasons for choosing the. numerical e values listed. A list of references used in developing the parameters is 2also' included.

.The parameters in the PWR-GALE Code will be updated periodically and

published-in revisionsito this NUREG as additional operating data become available. . The source. term parameters used are believed to provide a realistic assessment of. reactor and radwaste system operation.

~

[2.2 -PRINCIPAL PARAMETERS AND THEIR BASES 2.2.1 THERMAL POWER LEVEL

2. 2.1.1 - Parameter

The maximum thermal power level (MWt) evaluated for safety- considera-tions-in the Safety Analysis Report.

' 2.2.1.2 Bases The power level ~used in the source term PWR-GALE Code is the maximum-

. power-level evaluated for safety considerations in the Safety' Analysis Report. Using'this value,1the evaluation of the radwaste management systems need- not be repeated when the applicant applies for a stretch power: license at^atl ater date.. Past experience indicates that most utilities-request.

approval--to operate at maximum power soon after reaching commercial.

operation.

, s J 2.2.2- PLANT CAPACITY FACTOR

2.2.2.1 --Parameter A plant' capacity- factor of 80% is used, i.e., 292 effective full

power days.

t 2-1

. . + . . - - -- .. .- - . . - -- .- -

y  ; ' ,

n, a

- 12.2.2;2 Bases.

LThe= source term calculations are based on a plint capacity factor of 80%' averaged over the ..30-year ~ operating'; life of. the plant, i.e. , the plant noperates at 100% power 80% of the time. The plant capacity factors

? experienced at PWR's *~

a're listed in Table 1 for the period 1972 through b l977. ~

~

'The: average _' plant capacity factors shown in Table 2-1_ indicate that:the 80%. factor assumed is higher than the average; factors experienced.

However, it is expected'that the major maintenance problemsc and extended refueling outages that ha've. contributed .to the lower plant capacity factors t ~ 3 will be overcome and that the plants will achieve the 80% capacity factor when averaged ~ over 30 years of operation.

2.2.3.' RADIONUCLIDE CONCENTRATIONS IN THE PRIMARY"AND SECONDARY COOLANT i a

2.2.3.1 .

Parameter t

s As used -in the'PWR-GALE Code, Tables 2-2 and 2-3 list the expected a*~

.radionuclide' concentrations in the reactor coolant and steam for PWR's with design' parameters within the ranges listed in Tables 2-4 and 2-5. Should.

- any design parameter be outside the range in Tables 2-4 and 2-5, the PWR-

' GALE Code adjusts the concentrations in Tables 2-2 and 2-3, using the

'factorstin Tables 6, 2-7, and 2-8. Figares 2-1 and 2-2 show the graphical

' relationship of the design parameters.

s2.2.3.2. Bases-

.The radionuclide concentrations, adjustment . factors, and. procedure for effecting adjustments are based on the values and methods in American

. National: Standard ANSI N237, Source Term Specification, (Ref.il) but have

-been' updated based on a recent compilation of available operating data '

.concerning; primary, coolant concentrations, steam generator tube leakage,

-and secondary side radionuclide behavior. Therefore, the concentration

<. Evalues in NUREG-0017,> Rev. l' differ from the ANSI N237 ~ values, g ,' The : values in' Tables. 2-2' and 2-3 provide a set of typical radionuclide

- concentrations in the primary and secondary systems for reactor designs

within the parameters specified in Tables 2-4 ano 2-5. - The values in

Tables 222 and 2-3 are those determined to be representative of radio-

nuclide concentrations in a PWR over its lifetime based on the currently:

' :available data and models. The secondary coolant concentrations given in

' Tables 2-2 and 2-3 are calculated by using the reference parameters given

-in Table ~ 2-6 land the~ equations given in Tables 2-7 and .2-8. . It is E~ 1 recognized that some systems will have design parameters that are outside p .

[the. ranges lspecified in Tables 2-4 and 2-5. .For that reason, a means of.

adjusting the concentrations to the actual design parameters has been 1 . provided in' Tables 2-6 through 2-8. The adjustment factors in Table 2-6 through.2-8 are based on the following expression; g;

I' 2-2 I.  : , - , .- . ~.. - _ . - . ,_ _._. _ _ . _ _ _. .. . _ . _ . _ _ _ _ _ .

o TABLE 2-1 4.- PLANT CAPACITY FACTORS AT OPERATING PWR's'

~

Date of Commercial c b

' FACILITY Operation :1972 1973 '1974, _1975 1976 1977 d

Haddam Neck 1/68 86 48 89 84 81 82 San Onofre 1 1/68 72 60 83 85' 66 62 d

R. E. Ginna 7/70 58 81 50 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 d

. Palisades . 12/71 61 40' d 46' 50 78

Point Beach 2. 10/72 72 77 88 86- 82 Turkey Point 3 12/72 55 61 76 75- 78 d

Surry 1_ 12/72 51 50 60 67 78 f

Maine Yankee 12/72 17 54 69 77 d 91' 65 Surry 2~ 5/73 40 76 51 Oconee l 7/73 54 71 54 54 d

Indian Point 2 8/73 51 68 31 73 Turkey Point 4- 9/73 71 68 64 62 Fort Calhoun 9/73 54 -57 76 61' 83 Prairie Island 1 12/73 36 83 73 Zion 1 12/73 499 68 55 58 LKewaunee 6/74 75 75 77.

-Three Mile Island 1 9/74 79 63 79 d

Oconee 2 9/74 68 58 53 Zion 2 9/74 68 54' 71

-Oconee 3 12/74 69 64 71 Arkansasl .

12/74 69 54 73 Prairie Island 2 12/74 73 69 87 Rancho Seco 4/75 28 9 75 Calvert Cliffs 1 5/75 88 65 f

Cook 1 8/75 75 54 Millstone 2 12/75 68 63 Trojan 5/76 71 Indian. Point 3 8/76- 72' Beaver Valley 1 10/76 44

'St. Lucie 1 12/76 78 AVERAGE 71 64 69 72 69 74 a

.From nonthly Operating Units Status Reports. i b -Indian ' Point- 1 and Yankee Rowe are not included since they are small reactors

[< 700 MW(t)].

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

d Not . included due to. extended outage for refueling /amintenance.

' Not included due to extended maintenance / repair to the secondary system.

I Not' included due to extended operation at reudced power.

9--Not included due to extended maintenance outage to repair generator.

2-3

e ig

, TABLE 2-2

. NUMERICAL VALUES - CONCENTRATIONS IN PRINCIPAL FLUID STREAMS l 0F-THE REFERENCE PWR WITH U-TUBE STEAM GENERATORS l

(pCi/g) j Seconda ry ~ Coolant

• Isotope: Reactor Coolant ** Water *** Steam it

" Noble Gases Kr-85m :1.6(-1)iti 3.4(-8)

'Kr '

4.3(-1): 8.9(-8)

Kr 1.5(-l) 3.0(-8)

Kr-88. 2.8(-1 ) 5.9(-8)

Xe-131m 7.3(-l) 1.5(-7 Xe-133m 7.0(-2) 1.5(-8

.Xe-133 2.6(0) 5.4(-7

' Xe-135m 1.3(_1) cXe-135 2.7(-8)

.8.5(-1 )- 1.8(-7)

Xe-137 .3.4(-2) 7.1 (-9)

Xe-138 1.2(-1) 2.5(-8)

Halogens. '

'Br-84 1.6(-2) '7.5(-8) 7.5(-10) -

I-131 4.5(-2) 1.8(-6) l'8(-8)

I-132- 2.1(-1) 3.1(-6) 3.1(-8).

I-133 1.4(-1) 4.8(-6) 4.8(-8) 1-134 3.4(-1) 2.4(-6) 2.4(-8)

I-135 2.6(-1)' 6.6(-6) 6.6(-8)

Cs, Rb Rb-88 1.9(-1) 5.3(-7) 2.6(-9)

- Cs-134 7.1 (-3) 3.3(-7) 1.7(-9)

Cs-136 8.7(-4) 4.0(-8) 2.0(-10)

Cs-137 9.4(-3) 4.4(-7) 2.2(-9)

Water Activation Products N-16_ 4.0( +1 ) 1.0(-6) 1.0(-7)

Tritium H-3 1.0(0) 1.0(-3) 1.0(-3)

' Other Nuclides

Na-24 4.7(-2) 1.5(-6) 7.5(-9) i C r-51 3.1(-3) 1.3(-7) 6.3(-10)

Mn-54 1.6(-3) 6.5(-8) 3.3(-10) 2-4

7..

1 TABLE 2-2 (continue'd)

Secondary.. Coolant

• ii

. Isotope Reactor Coolant ** ' Water *** Steam Fe 1.2(-3) 4.9(-8) 2.5(-10)

Fe-59 3.0(-4 1.2(-8) 6.1(-11)

' Co-58: 4.6(-3 1.9(-7) 9.4(-10)-

Co-60 5.3(-4 2.2(-8) 1.1 (-10)

Zn-65 5.1 ( -'4) 2.1(-8)~ 1.0(-10)

Sr-89 1.4(-4) 5.7(-9) 2.9(-11 )

Sr,-90 1.2(-5) 4.9(-10) 2.5(-12)

_ S r-91 9.6(-4) 2.8(-8) 1.4(-10)

Y-91m 4.6(-4) 3.2(-9) 1.6(-11)

Y-91 5.2(-6) 2.1(-l0) 1.1(-l2)

-Y-93 4.2(-3) 1.2(-7) 6.1 (-10)

Zr-95 3.9(-4) 1.6(-8) 7.9(-11)

Nb-95 2.8(-4) 1.1(-8) 5.7(-11)

Mo-99; 6.4(-3) 2.5(-7) 1.2(-9)

~ Tc-99m 4.7(-3 1.1 -7) 5.7(-10)

Ru-103 7.5(-3 3.1 -7) 1.6(-9)

Ru-106 9.0(-2 3.7 -6) 1.8(-8)

Ag-110m 1.3(-3) 5.3(-8) 2.7(-10)

- Te-129m 1.9(-4) 7.8(-9) 3.9(-11)

Te-129 2.4(-2) 2.2(-7) 1.1(-9)

Te-131 m . 1.5(-3) 5.4(-8) 2.7(-10)

- Te-131 7.7(-3)' 2.9(-8) 1.5(-10)

Te-132 1.7(-3) ,6.6(-8) 3.3(-10)

Ba-140 1.3(-2) 5.2(-7) 2.6(-9)

La-140 2.5(-2 9.3 -7) 4.6(-9)

Ce-141. 1.5(-4 6.1 -9) 3.1 (111)

Ce-143 2.8(-3 1.0 -7) 5.1 (-10)

- Ce-144 3.9(-3 1.6 -7) 8.2(-10)

W-187 2.5(-3 8.7(-8) 4.4(-10) -

Np-239 2.2(-3) 8.4(-8) 4.2(-10)

• Based on a' primary-to-secondary leak of 75 lb/ day.

'** The concentrations given are for reactor coolant entering the letdown ,

line. These concentrations are obtained from Tables 2-9 and 2-10.

N-16 and H-3 concentrations are obtained from Reference 1.

• • • The concentrations given are for water in a steam generator.

tt The concentrations given are for steam leaving a steam generator, ttt 1.6(-1) = 1.6 x 10-I .

2-5

.. . -- _ ~ . . . . . . . . . .-.

TABLE 2-3

NUMERICAL VALUES - CONCENTRATIONS IN PRINCIPAL FLUID STREAMS OF THE REFERENCE PWR WITH ONCE-THROUGH STEAM GENERATOR _S

, (pCi/g)

Isotope Reactor Coolant

• Secondary Coolant **

Noble Gases Kr-85m 1.6(-1 ) 3.4(-8) )

.Kr-85 4.3 -1) 8.9(-8 Kr-87 '- 1.5 -1 ) = 3.0(-8 lKr-88 2.8 -1 ) 5.9(-8 Xe-131m 7.3(-1) 1.5(-7)

Xe-133m - 7.0(-2). 1.5(-8) c Xe-133.. 2.6(0) 5.4(-7)~

Xe-135m 1.3-1) .2.7(-8)

~Xe-135 8.5 -1) 1.8(-7) ,

Xe-137 3.4 -2) 7.1(-9  ;

Xe-138 1.2 -1) 2.5(-8

-Halogens

.Br-84 1.6(-2) 1.8(-8)

I;131-4.5(-2) 5.2(-8)

I-132 2.1 -1 ) 2.4 -7)

I-133 1.4 -1) 1.6 -7)

I-134 3.4 -1) 3.8 -7)

I-135 2.6 (-1 ) 3.0(-7)

Cs, Rb Rb 1.9h1) 6.0(-7)

Cs-134- 7.1(-3) 3.0(-8)-

Cs-136 8.7(-4) 3.6(-9)

Cs-137'- 9.4(-3) 3.9(-8)

Water Activation Products N-16 4.0(+1 ) 1.0(-6)

-Tritium l

H-3 ~ 1.0(0) 1.0(-3) ,

f. -

Other Nuclides Na-24 -4.7(-2) 1.0(-7)

C r-51 3.1(-3) 6.9(-9)

Mn-54 .1.6 -3) 3.6(-9) 1 Fe'-55 1.2 -3) 2.7(-9) 3 Fe-59 3.0 -4) 6.7(-10)

Co-58 4.6(-3) u3(-8) e Co 5.3(-4) 1.2(-9)

L 2-6

4 TABLE 2-3 (continued)

-Isotope Reactor' Coolant * ' Secondary Coolant **

Zn-65. 5.1(-4) 1.1(-9) e Sr-89' .1.4(-4) 3.1(-10)

Sr-90 u 1.2(-5). 2.7(-11)-

Sr-91 9.6(-4) 2.1 (-9) -

' ^ ~

Y-91 m 4.6(-4) 9.7(-10)-

Y-91 5.2(-6). 1.2(-11) t: .Y 4.2(-3) 9.3(-9)

Zr-95 ^ 3.9(-4) 8.7(-10)

- Nb-95 2.8(-4) 6.2(-10)

- Mo-99 6.4(-3) 1.4(-8)-

Tc-99m 4.7(-3)- 1.0(-8)

- Ru-103 7.5(-3) 1.7(-8)

' Ru-106 9.0(-2) 2.0(-7)

- Ag-110m 1.3(-3)' 2.9(-9)

. Te-129m 1.9(-4) 4.2(-10) 4 Te-129 2.4(-2) 5.1(-8) -

-Te-131m 1.5(-3) 3.3(-9)

Te-131 : 7.7(-3) 1.5(-8)

Te-132 1.' 7 (- 3) 3.8 -9)- -

- Ba-140 1.3(-2) 2.9 -8)

' La-140 2.5(-2) 5.6 -8)

Ce-141 1.5(-4) 3.3(-10) 2.8(-3) 6.2(-9)

Ce-143

-Ce-144 3.9(-3) 8.7(-9)

-y-187. -2.5(-3) 5.6(-9)

.Np-239- 2.2(-3) 4.9(-9)-

• The concentrations given are reactor coolant entering the letdown line.

. . , These concentrations are obtained from Tables 2-9 and 2-10. N-16 and i H-3 concentrations are obtained from Reference 1.

• • Based on primary-to-secondary leakage of 75 lb/ day. The concentrations t

^

given are for steam leaving a steam generatnr.

l l

l

{

i

2-7

TABLE 2-4 PARAMETERS USED TO DESCRIBE THE REFERENCE PRESSURIZED WATER REACTOR WITH U-TUBE STEAM GENERATORS i Nominal Range ~

Parameter- Symbol ~ Units Value Maximum Minimum Th;rmal Power P MWt 3,400 3,800 3,000

'Stgam flow rate FS lb/hr 1.5(7) 1.7(7) -1.3(7)

W2ight of water in reactor coolant WP .lb 5.5(5) 6.0(5) 5.0(5) system Weight- of water. in all steam - WS lb 4.5(5) 5.0(5) 4.0(5)~

,g:nerators i

Reactor coolant letdown flow FD lb/hr 3.7(4) 4.2(4) 3.2(4)

(purification)

R actor. coolant letdown flow (yearly FB lb/hr 500 1,000 250 average for boron control)

Steam generator blowdown flow (total) FBD lb/hr 75,000 100,000 50,000

Fraction of radioactivity in blowdown NBD --

1.0* 1.0 0.9 stream that is not returned to the srcondary coolant system Flow through the purification system FA lb/hr 3,700 7,500 0.0-cation demineralizer :

Ratio'of condensate demineralizer flow NC --

0.0** 0.01 0.0

' rate to the total steam flow ~ rate Ratio' of the total amount of noble Y -- 0.0 0.01 0.0 gases' routed to-gaseous radwaste

from the purification system to the .

total amount of noble gases routed from the primary coolant system to  ;

the purification system (not in-cluding' the boron recovery system) '

• This value is. based on a nominal case of blowdown through blowdown demineralizers back to the main condenser (no condensate demineralizers). Value taken from blow-down demineralizer DF's in Section 2.2.18. Value for cesium and rubidium is 0.9.

oo This value is based on a nominal case of no condensate demineralizers. For a U-tube steam generator PWR with full flow condensate demineralizers, a value of NC = 1.0:is used by the PWR-GALE Code. For a U-tube steam generator PWR with condensate demineralizers and pumped forward feedwater heater drains, the value.

for NC used by the PWR-GALE Code is 0.2 for iodine, and 0.1 for Cs, Rb and other nuclides as discussed on page 2-20.

2-8

r TABLE 2-5 PARAMETERS USED TO DESCRIBE THE REFERENCE PRESSURIZED WATER REACTOR WITH ONCE-THROUGH STEAM GENERATORS Nominal Range Parameter Symbol . Units Value Maximum Minimum Th;rmal Power P MWt 3,400 3,800 3,000 '

'St:am flow rate FS lb/hr 1.5(7) 1.7(7) 1.3(7)

Weight of water in reactor coolant WP lb 5.5(5) 6.0(5) 5.0(5) system Weight of water.in all steam WS lb 1.0(5) *

• generators R; actor coolant letdown flow FD lb/hr 3.7(4) 4.2(4) 3.2(4)

'(purification)

Rsactor coolant letdown flow (yearly FB lb/h r 500 1,000 250 av: rage for boron control)

Flow through the purification system FA lb/hr 3,700 7,500 0.0 cation demineralizer

-Ratio of condensate demineralizer NC --

0.65** 0.75 0.55 ,

. flow rate to the total steam flow rate Ratio of the total amount of noble- Y --

0.0 0.01 0.0 gas:s routed to gaseous radwaste from the purification system to the total amount routed from the primary coolant system to the  ;

purification system (not including the boron recovery system)

• The secondary coolant inventory is not of importance in a once-through steam g:ncrator plant because decay is not an important removal mechanism for most

-of the isotopes.

N

• • For 3 PWR that is within the range indicated above, i.e., a PWR with pumped forward feedwater heater drains, the value for NC used by the PWR-GALE Code is 0.2 for iodine and 0.1 for Cs, Rb and other nuclides, as discussed on page 2-20.

For a PWR that has full flow condensate demineralizer, a value of NC = 1.0 is us;d by the PWR-GALE Code.

2-9 m

1

^

)

~

l TABLE 2-6

~

VALUES'USED IN DETERMINING ADJUSTMENT FACTORS FOR PRESSURIZED WATER REACTORS .

Element Class Water Noble Activation Other Sy:bol Description Gases Halogens Cs, Rb Products M Nuclides NA Fraction- of material 0.0- 0.0 ~ 0.9 0.0 0.0 0.9*

removed in'. passing through.the cation demineralizer '

.NB Fraction of material 0.0 0.99 0.5 0.0 0.0 0.98-

~

removed ~1n passing through the purification demine,ralizer.

R- r 0.0009 0.067 0.037 0.0 *** 0.066 Removal rate j )*eactor coolant (Hr~

• NS Ratio of concentration in steam to that in water in the steam generator

.U-tube steam generator- t 0.01 0.005 tt 1.0 0.005 Once-through steam t 1.0 1.0 1.0 1.0 1.0 generator

-NX- Fraction of activity 0.0~ 0.9 0.5 0.0 0.0 0.9 removed in passing through the condensate

~

demineralizers r s Removal rate j )econdary coolant (Hr- ttt U-tube steam generator t 0.17 0.15 tt *** 0.17 Once-through steam t 27 7.5 tt *** 14 2, generator FL Primary-to-secondary 75 75 75 75 75 75 leakage (lb/ day)

A These represent effective removal terms and include mechanisms such as plateout.

--Plateout would be applicable to nuclides such as Mo and corrosion products.

2-10 -

m , _ _ . _ _ _ . _ _ . _ . .

J s < -

TABLE-2-6 (continued) t o .

'00. tThese values of R apply to the reference PWR's whose parameters are given in

Tables 2-4 and 2-5 and have :been used in developing Tables 2-7 and 2-8. For

'PWR'.s1not included in Tables'2-4 and 2-5, the appropriate'value for l R may be i d;termined by the .following equations.

^

. R = FB +'(FD -'FB)Y' for noble gases -

. WP.

R = (FD)(NB)'+ (1 - NB)(FB + (FA)(NA)) for halogens, Cs, Rb, and other nuclides The concentration of tritium is a function of (1) the inventory of tritiated

~

L***

liquids-in the plant, (2) the rate of production of tritium due to 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 concen-

. trations given in. Tables 2-2 and 2-3 are representative of PWR's with a moderate, Lamount of tritium recycle and can be used to calculate. source terms in accordance-

twith Regulatory Guide 1.112. " Calculations of Releases of Radioactive Materials in Gaseous- and Liquid Effluents from Light-Water-Cooled Power Reactors."

't ONoble gase's 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.

- tt Water activation products exhibit . varying chemical and physical properties in reac. tor coolants that are not well defined. Most are_not effectively removed by the demineralizers, but their concentrations are controlled by decay.

< :ttt These values of r apply to the reference PWR's whose parameters are.given in LTables 2-4 and 2-5 and have been used in developing Tables 2-7 and 2-8. For PWR's

~n ot included in Tables 2-4 and 2-5, the appropriate value_ for r may be determined by the'following equation: )

j-

~

J -r = (FBD)(NBD) + NS)(FS)(NC)(NXl for halogens, Cs, Pb, and other nuclides .

i a +

n.

4 w 4 1

2-11 1

..s% -

..w... - , , , , , , , , - , , . . , - - -,,-,,..e ,,,y,,..--w,,,.m,.m y_m ,, , -r - - - - -----~_~--+t e ce = - ve- - -g -y -n ww *w%, 4--w=t---

,, _ v 4

4" TABLE 2-7 ADJUSTMENT FACTORS FOR PWR's WITH U-TUBE STEAM GENERATORS Adjustment Factors-Secondary Coolant Element Clas's Reactor Water (f)* Water Steam Noble gases 162P 0.0009 + A ** __

l .5 x 10 7 f WP R+A FS Halogens 162P 0.067 + A 4.5 x 10 0.17 + A 4.5 x 1050.17 + A WP. R+A WS r+A WS r+A 5

Cs, Rb 162P 0.037 + A 4.5 x 10 . 0.15 + A f 4.5 x 1050.15 + A f y WP R+A WS r+A WS r+A 5

4.5 10 4.5 10 Water activation 1.0 products Tritium *** *** ***

Other nuclides 1 62 0.066 + A 4.5 x 10 0.17 + A f 4.5 x 1050.17 + A f WP R+A WS r+A WS r+A

• f is the reactor water adjustment factor and is used in the secondary coolant adjustment factors.

• • A is the isotopic decay constant (hr ).

• • • The concentration of tritium is. a function of (1). the inventory of' tritiated liquids in the plant, (2) the' rate of production of tritium due to 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 2-2 and 2-3 are representative of PWR's with a moderate amount of tritium recycle and can be used to calculate source terms in accordance with Regulatory Guide 1.112. ,

ee - = _

TABLE 2-8 ADJUSTMENT FACTORS FOR PWR's WITH ONCE-THROUGH STEAM GENERATORS Adjustment Factors Nuclide Reactor Water-(f)*~ Secondary Coolant 162P 0.0009 + A 1.5 x 10 Noble gases R+A FS 7

WP 5

162P 0.067 + A Halogens WP R+A 10 (27r+A WS

+A}

5 Cs, Rb 162P 0.037 + A 10 WP R+A WS 7.5r+A

+A}7 5

1.0 1.0 x 10 Water activation WS products Tritium 5

162P 0.066 + A 10 j4, A ) f

- Other nuclides WP R+A WS r+A

• f is the reactor water adjustment factor and is used in the secondary coolant adjustment factors.

• • The. concentration of tritium is a function of (1) the inventory of tritiated liquids in the plant, (2) the rate of production of tritium due to 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 2-2 and 2-3 are representative of PWR's with a moderate amount of tritium recycle and can be used to calculate source terms in accordance with the Regulatory Guide 1.112.

I 2-13

a t

t LEAKAGE w#

FS NS _ _

STEAM STEAM OFF GAS REACTOR STEAM - MAIN GENERATORS GENERATORS VESSEL PRIMARY SECONDARY FS(1-NC)

SIDE SIDE

/

d, FEEDWATER \ I'.

8 NC.FS I ,,

BLOWDOWN I

- _. /b _J FD FBD l

CONDENSATE DEMINERAllZER Y

E NX l

l PURIFICATION FA CATION DEMINERALIZER DEMINERALIZER SYMBOLS ARE DEFINED IN TABLES 2 4 AND 2 6 l

l NB

!  ; FB

TO BORON RECOVERY SYSTEM i

PURIFICATION y SYSTEM  % DEGASSING j TANK l

l FIGURE 21 REMOVAL PATHS FOR PRESSURIZED WATER REACTOR WITH U TUBE STEAM GENERATORS

LEAKAGE

• d FS NS REACTOR STEAM MAIN ESSEL GEN RA ORS GENERATORS PRIMARY SECONDARY ps(1.NC)

SIDE SIDE

- /

o FEEDWATER \

,, NC FS FD CONDENSATE DEMINERALIZER 7 NX G

_ PURIFICATION FA _ CATION DEMINERALIZER '

DEMINERALIZER SYMBOLS ARE DEFINED IN TABLES 2-5 AND 2-6 NB FB

TO BORON RECOVERY SYSTEM PURIFICATION Y SYSTEM e DEGASSING TANK FIGURE 2-2 REMOVAL PATHS FOR PRESSURIZED WATER REACTOR WITH ONCE THROUGH STEAM GENERATORS l

i I. . -

j r -

s C=

w(A + R)K where C is the specific activity .(in pCi/g) I K is a conversion factor, 454 g/lb i 1

A . R is the removal rate of the isotope from the system due to ~

demineralization, leakage, etc. (hr 1). .

(If considering secondary coolant R = r).

s is the rate of release to and/or production of the isotope.

in the system (in pCi/hr) w- is.the fluid weight (in Ib), and is the decay constant (hr-1 ).

'+

The following sample calculations illustrate the method by which the PWR-GALE Code will adjust the radionuclide concentrations in Tables 2-2 and 2-3. . As indicated in Tables 2-7 and 2-8, adjustment factors will be calculated for noble gases, halogens, Cs, Rb, and other nuclides.

7 As an example,- the sample case parameters shown below compare with the range of values in Table 2-4 as follows.

Parameter (U-tube steam' generator PWR) Value Range Thermal power level, MWt 3800 3000 - 3800 0 6 Steam flow rate, Ib/hr 17 x 10 13 x 106 - 17 x 10

. Mass of reactor coolant, Ib 5.5 x 10 5 5 5.0 x 10 - 6.0 x 10 5

5 5 Water weight in all steam generators, lb 4.4 x 10 5 4.0 x 10 - 5.0 x 10 4

Reactor coolant letdown, Ib/hr 4.9 x 10 4 3.2 x 10 - 4.2 x 10 4

Cation demineralizer flow, lb/hr 4.9 x 10 3 0 - 7.5 x 10 3 Shim bleed rate - yearly average, .lb/hr 650 250 - 1000 Steam generator blowdown flow, . lb/hr 60,000 50,000 - 100,000 Fraction of blowdown activity not 0.99 0.9 - 1.0 returned to secondary system Cation demineralizer flow, Ib/hr 4900 0.0 - 7500 Condensate demineralizer flow fraction 0.0 0.0 - 0.01

-Y (see definition in Table 2-4 and page 1-26) 2-16

Since in this example the parameter for reactor coolant letdown

- rate (4.9 x 10 lb/hr) is outside the range specified in Table 2-4 (3.2 - 4.2 x 10 lb/hr), and the sample case employs continuous purging of the volume control tank, the primary coolant activity is recalculated using the actual design value for all parameters employing the methods described below.

1. Noble' Gases (Xe-133 is used as an example)

Using the equation for noble gases in' Table 2-7, the adjustment factor,. f, is calculated as follows:

f = 162P 0.0009 + A (I)

WP R+A where the terms in the equations are defined in Tables 2-4 and 2-6.

In calculating f, the variable R is calculated first by using the equation given.in Table 2-6 for noble gases R = FB + (FD - FB)(Y) 2) where the terms of the equation are as defined in Tables 2-4 and 2-6.

Use the sample case parameters given above and the noble gas parameters given in Table 2-6 and substitute in Equation (2) above.

4 R = 650 + (4.9 x 10 - 650) x 0.25 = 0.023 5.5 x 10 Use the value of R in Equation (1) above.

7 ,162 x 3800 0.0009 + 5.5 x 10-3 0.25 5.5 x 10 5 0.023 + 5.5 x 10-3 The adjusted Xe-133 primary coolant concentration

= (adjustment factor) x (standard Xe-133 concentration)

= 0.25 x 2.6 pCi/g = 0.65 pCi/g

2. Halogens (I-131 is used as an example)

Using the equation for halogens in Table 2-7, the adjustment factor, f, is calculated as follows:

2-17

te-I '" 162P ' 0.067 ' + A (3)

WP R+A -

where the terms in the equations are defined in Tables 2-4 and 2-6. l

--In calculating f, the variable R is calculated first by using the I equation given in Table 2-6.

. R' = (FD)(NB) + (1 - NB)(FB + -(FA)(NA)) (4)-

WP where the terms in the equation are as defined in Tables 2-4 and 2-6.

_Use the sample case parameters given above and the halogen parameters given in Table.2-6 and substitute in Equation (4) above.

4 R'= (4.9 x 10 x 0.99) +-(1 - 0.99)(650 5

+ (4900)(0.0)) = 0.088 5.5 x 10 Use the value of R in Equation (3) above.

_7 , 162(3800) 0.067 + 3.6 x 10-3 0.86 5

5.5 x 10 0.088 + 3.6 x 10-3 The adjusted'I-131 concentration

= (adjustment factor) x (standard I-131 concentration)

= 0.86 x 0.045 pCi/g = 0.039 pCi/g

3. Cs, Rb (Cs .137 is used as an example)

Using the equation for Cs and Rb in Tab 1'e 2-7, the adjustment factor, f, is calculated as follows: ,

f " 162P WP 0.037 R+A+ A (5) where the terms in the equation are as defined in Tables 2-4 and 2-6.

In calculating f, the variable R is calculated first by using Equation (4) above. The Cs and Rb parameters given in Table 2-6 and the sample case parameters given in Table 2-9 are used in.the equation.-

4 R = (4.9 x 10 x 0.5) + (0.5)(650 +-(4900)(0.9)) = 0.05 5.5 x 10 2-18

Use the value of R in Equation (5) above.

f ,162(3800) 0.037 + 2.6 x 10-6 = 0.83 5

5.5 x 10 0.05 + 2.6 x 10-6 i.

The adjusted Cs-137 concentration

= (adjustment' factor) x (standard Cs-137 concentration)

= 0.83 x 9.4 x 10-3 p Ci/g = 7.8 x 10-3 pCi/g 4.. Other Nuclides (Te-132 is used as an example)

Using the equation for other nuclides in Table 2-7, the adjustment factor, f, is calculated as follows:

I " 162P WP 0.066 R+A +A }

where the terms in the equation are as defined in Tables 2-4 and 2-6.

In calculating f, the variable R is calculated first by using Equation

.(4) above. The parameters for other nuclides given in Table 2-6 and the sample case parameters given in Table 2-9 are used in the equation.

4 R = (4.9 x 10 ) (0.98) + (1 - 0.98)(650 + (4900)(0.9)) = 0.087 5.5 x 10 5 Use the value of R in equation (6) above.

f , 162 (3800) (0.066 + 8.9 x 10-3) = 0.87 5 -3 5.5 x 10 0.087 + 8.9 x 10 The adjusted concentration of Te-132

= (adjustment factor) x (standard Te-132 concentration)

= 0.87 x 1.7 x 10-3 pCi/g = 1.5 x 10-3 Ci/g A similar method is used in the PWR-GALE Code to adjust secondary coolant concentrations for reactors with parameters outside the ranges specified in Tables 2-4 and 2-5.

The radionuclide primary coolant concentrations in Tables 2-2 and 2-3 are based on data submitted by utilities with operating PWR's (Ref. 3).

The data are also based on measurements taken by the NRC at Ft. Calhoun (Ref. 4), Zion 1 and 2 (Ref. 5), Turkey Point 3 and 4 (Ref. 6), Rancho Seco (Ref. 43), .and Prairie Island 1 and 2 (Ref. 42); by EPRI (Ref. 7) at Three Mile Island 1 and Calvert Cliffs; and by measurements at various other_ PWR's (Ref. 8, 9, and 39).

2-19

These data are summarized in Table 2-9 and Table 2-10 indicating the average value of the nuclide concentration for each plant, the years over which the data was obtained, and the total number of years of data for each nuclide.

The secondary coolant concentrations are based on the primary coolant concentrations as obtained above, on 75 lb/ day primary-to-secondary leakage in the steam generators, on appropriate steam generator carryover f actors, on the appropriate main steam flow, steam generator blowdown flow and fraction of a blowdown flow returned to the secondary coolant, as defined in the plant design, and on the fraction of the nuclides in the main steam which return to the steam generators.

The secondary coolant concentrations are based on 75 lb/ day primary-to-secondary leakage. The pr nary-to-secondary leakage rate experience for 79 years of experience at cperating PWR's is given in Table 2-11.

The average primary-to-secondary leakage rate in Table 2-11 is 75 lb/ day.

Westinghouse estimates that the data in Table 2-11 are accurate within

+ 25% (Ref. 8, 39).

For plants using recirculating U-tube steam generators, carryover due to mechanical entrainment is based on 0.5% moisture in the steam.

Table 2-12 provides measured values for moisture carryover at five operating PWR's that use recirculating U-tube steam generators. Based on data from Turkey Point 3 and 4 (Ref. 6) a value of 1% iodine carryover with the steam is used in our evaluations. For once-through steam generators, it is assumed that 100% of both nonvolatile and volatile species is carried over with the steam since this type of steam generator has no liquid reservoir and 100% of the feed is converted to steam.

For PWR's that use condensate demineralizers in the secondary system, the nominal value oi 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 which utilizes a pumped forward model, that is, one in which the reactor steam flow is split with 65% flowing to the low pressure turbines and the main condenser, and 35% pumped forward to the feedwater. The fraction pumped forward to the feedwater does not undergo any treatment in the condensate demineralizers. We have determined that the iodine, Cs, Rb, and "Other Nuclides" of Table 2-2 and Table 2-3 preferentially go with the " pumped forward" fraction. The reason for this is that these nuclides show a tendency to go with the condensed steam in the moisture separator-reheater drains and with the extraction steam lines from the high pressure turbines to the feedwater system.

Based on data provided in Ref. 6, 7,12 and 13 for Turkey Point, Point Beach and Brunswick, the percentages used in the PWR-GALE Code for the amount of activity which is pumped forward and which bypasses the condensate demineralizers is 80% for iodine and 90% for Cs, Rb, and "Other Nuclides" of Table 2-2 and Table 2-3. Since the remainder of the nuclides listed in Tables 2-2 and 2-3 are not removed in the condensate demineralizers, we have not considered the magnitude of bypass for those nuclides.

2-20

m. <

. -j

_; r

,a >

1

^

. TABLE 219-

SUMMARY

OF ~I-131 AND-I-133 PRIMARY COOLANT CONCENTRATIONS IN PWR'S*

.(pCi/g)

H.B. Robinson 2 Arkansas 1 ' D.C. Cook 1 Trojan Palisades Point' Beach 1/2 R.E. Ginna Isotope (1973-1978)**' (1976) -) (1976-1978) _ (1977-1978) (1972-1976) (1972-1979) (1971-1978)

I-1 31 3.1E-03 7.3E-03 7.8E-03 1.3E-02 1.2E-02 7.7E-02 2.2E-01 1-133 .1.2E-02 *** 1.8E-02 .l.5E-02 1.6E-02. 3.6E-01 . 6.9E-01 Fort Calhoun 1 Zion 1/2 Turkey Point 3/4 Three Mile Island-1 Calvert Cliffs 1. Beaver Valley 1 Isotope (1976-1977) (1975-1978) (1974-1978) (1975-1977)-' (1976) (1977-1978)

I-131 1.8E-01 2.3E-02 2.1E-02 2. l E-02 3.6E-02 1.8E-03

• • • *** 5.5E-03

'? I-133 1.6E-01 6.9E-02 6.9E-02 I

.-Indian Point 2/3 Kewaunee Prairie Island 1/2 Surry 1/2 J.M. Farley 1 Yankee Rowe Rancho Seco Isotope (1975-1978) (1975-1978)_ (1975_19311 , (1973,-1978) (1978) (1975) (1978)

I-131 .2.5E-02 5.3E-03 8.lE-03 2.1E-02 5.4E-04 4.9E-03 1.3E-02 I-133 5.1E-02 1.3E-02 1.0E-02 3.2E-02 1.9E-03 2.4E-02 4.7E-02

• . Data in this table are based on I-131 and I-133 primary coolant concentrations in Ref. 3 through 9, 42 and 43, and have been adjusted to the NSS parameters listed in Table 2-4 of'this report. These adjustments were made using the individual plant parameters and the nominal plant parameters (Table 2-4) and adjusting the actual coolant concentration using the equations in Table 2-7 of this report.

• • Data in this table were gathered during the indicated inclusive dates. It does not necessarily imply that data were available during each of the years covered by the period, nor does it mean that the number of data points should be the same for each radionuclide.

• • • No value reported.

S

' ?}

f TABLE '2-10 ,

~

SUMMARY

OF RADIONUCLIOE PRIMARY COOLANT CONCENTRATIONS IN PWR'S*

(uC1/g)

H.B. Robinson 2 Arkansas.1 D.C. Cook 1 Trojan Palisades Point Beach 1/2 R.E. Ginna Isotcpe (1973-1978)** ' (1976) (1976-1978) -(1977-1978)1 -(1972-1976) (1972-1979) '(1971-1978)-

Kr-85m 1.8E 8.0E-03 3.4E-02 2.2E-02 4.0E-01 2.2E-01 -2.4E-01 ,

Kr-85 *** 5.2E-03 *** . 3.5E+00 8.3E-04 2.9E-02 ***

Kr-87 1.7E-02 6.7E-03 .4.7E-02' 4.2E-02 4.6E-01 1. l E-01 3.8E-01 Kr-88 2.3E-02 1.3E-02 5.7E-02 4.0E 7.5E-01 3.0E-01 6.3E-01 X:-131n *** *** *** *** 4.4E+00 9.4 E-01 ***

X:-133m 1.6E-03 3.3E-03 1.5E-02 2.7E-03 1. 9E-01 6.5E-02 ***

Xa-133 2.3E-01 2.1E-01 5.8E-01 5.7E-01 l4.9E+00 2.8E+00 6.0E+00 -

Xs-135m 2.1 E-02 ***! *** *** . 1.0E-02 1.4E-01 1.5E-01 Xa-135 7.8E-02 2.7E-02 1.9E-01 1.2E-01 1.1E+00- 1.lE+00 2.2E+00 X:-137 *** *** *** *** *** '*** ***

Xs-138 *** *** *** 2.2E-03 1.7E-01 ***

' 6.7E-02 7' Br-84 *** *** *** *** *** ***- ***

R$ I-132 1.5E-02 *** *** 1.9E-02 7.1 E-03 3.6E-01. 7.3E-01 1-134 3.2E-02 *** 2.4E-02 2.2E-02 1.0E-02 6.2 E-01 1.2E+00 I-135 1.9E-02 *** 2.0E-02 1.7E-02 9.2E-03 5.7E-01 6.6E-01 Rb-88 *** *** *** 3.5E-02 2.6E-02 1.7E-01 ' 3.7E-01 Cs-134 1.9E-03 5.6E-04 2. 7E-03 6.0E-04. 1.7E-04 1.4E-02 1.1E-02 Cs-136 3.1E-04 *** 5.2E-03 7.2E-04 4.6E-05 2.2E-03 ***

Cs-137 2.3E-03 1.5E-03 4.9E-03 1.3E-03 2.6E-04 1.1E-02 3.1E-02 N-16 *** *** *** *** *** *** ***

H-3 *** 6.3E-02 2.lE-01 *** 7.5E-02 6.7E-01 5.0E-01 Na-24 1.3 E-01 8.7E-02 1.2E-02 1.3E-02 5.6E-03 7.6E-02 ***

C r-51 3.5E-04 3.2E-03 *** *** 9.lE-03 *** 1.1E-04 Mn-54 3.4E-04 7.6E-04 8.3E-03 9.7E-04 1.1E-04 2.8E-03 2.5E-05 Fe-55 *** *** *** *** *** *** ***

Fe-59 1.4E-05 1.6E-03 -*** *** 1.6E-04 *** 2.6E-05 Co-58 1.3E-03 7.0E-03 1.4E-02 2.2E-03 3.4E-03 9.6E-03 7.6E-04 Co-60 3.5E-04 6.4E-04 4.5E-03 3.4E-05 1.1E-04 2.3E-04 1. 6E-04 Zn-65 1.7E-05 *** *** '*** 7.0E-05 *** ***

Sr-89 2.3E-05 *** *** ~ *** *** 2.4E-04 ***

Sr-90 5.2E-06 *** *** *** 1.1E-04 *** ***

l

' TABLE 2-10 (continued)

SUMMARY

OF RADIONUCLIDE PRIMARY COOLANT CONCENTRATIONS IN PWR'S*

(pCi/g)

H.B. Robinson 2 Arkansas l' D.C. Cook 1 Trojan ' Palisades ' Point Beach 1/2- R.E. Ginna Isotope (1973-1978)** (1976) (1976-1978) (1977-1978) (1972-1976) (1972-1979) (1971-1978) ~

Sr-91 4.9E-04. *** *** *** 1.1E-04 *** ***

Y-91m *** *** ***. ***- - *** *** ***

y.gj ... ... ... ... ... .... . . . -

Y- 93 *** *** *** *** *** *** ***

Zr-95 1.3E-05 3.4E-04 '4.5E-03 *** 1.0E-04 *** 1.5E-03 Hb-95 1.3E-05 3.lE-04 2.4E-03 *** 7.6E-05 3.6E-04 . 8.lE-05 Mo-99 *** 7.2E-05 *** *** 5.7E-04 3.8E-02 4.lE-04 Tc-99m *** *** *** *** 7.3E-04 2.5E-02 ***

Ru-103 *** *** *** *** *** *** 1.2E-03

,, Ru-106 *** *** *** *** *** *** ***

g, Ag-110m *** *** *** *** *** 8.8E-03 ***

us To-129m *** *** *** *** *** *** ***

Lo-129 *** *** *** *** *** *** ***

Te-131m *** *** *** *** *** *** ***

Te-131 *** *** *** *** *** *** ***

Te-132 *** . 1.3E-03 *** *** 6.6E-05 8.8E-03 ***

Ba-140 2.0E-04 *** *** *** 6.2E- 06 1.6E-01 5.9E-05 La-140 9.2E-05 *** *** *** 3.0E-05 5.2E-01 ***

Ce-141 *** *** *** *** *** *** ***

Cs-143 *** *** *** *** *** *** ***

Cc-144 2.6E-04 1.4E-03 *** *** *** 4.5E-02 ***

W-187 3.4E-04 *** *** - *** 5.8E-04 *** ***

Np-239 *** *** *** *** *** *** 2.0E-03

TABLE 2-10 (continued)

SUMMARY

OF RADIONUCLIDE PRIMARY COOLANT CONCENTRATIONS IN PWR'S*

(pCi/g)

Fort Tu rkey Indian Yankee Calvert Three Mile Prairie Rancho Calhoun 1 Zion 1/2 Point 3/4 Pt 2/3 Rowe' Cliffs 1 Island 1 Island 1/2 Seco Isotope (1976-1977) (1975-1978) (1974-1978) (1975-1978) (1975) (1976) (1975-1977) (1981) (1979)

Kr-85m 1.9E-01 *** 7.8E-02 3.4E-02 5.7E-03 *** *** 4.9E-04 5.5E-02 Kr-85 3.4E-02 *** *** *** *** *** *** 3.3E-04 2.2E-01 Kr-87 1.9E-01 *** 9.0E-02 *** 7.6E-03 *** *** 1.1E-03 5.9E-02 Kr 3.2E-01 *** 1.3E-01 7.3E-02 1.9E-02 *** *** 1.1E-03 9.9E-02 XI-131 m 6.8E-02 *** 1.2E-03 *** *** *** *** 4.2E-05 3.5E-03 X2-133m 1.6E-01 *** 9.1E-03 *** *** *** *** 7.2E-05 4.5E-02 Xa-133 6.7E+00 *** 8.8E-01 8.3E-01 2.1 E-01 *** *** 2.2E-03 1.5E+00 Xr-135m 9.5E-02 *** 1. 7E-01 1.0E-01 *** *** *** 1.4E-03 6.0E-01 X2-135 9.3E-01 *** 5.1E-01 1.9E-01 3.0E-02 *** *** 3.6E-03 4.6E-01

'? X -137 *** *** 3.4E-02 *** *** *** *** *** ***t-

% Xa-138 1.8E-01 *** 7.6E-02 *** *** *** *** 2.9E-03 1. 7E-01 Br-84 *** *** 1.1E-02 *** *** *** *** 1.0E-03 5.5E-02 I-132 7.1E-02 9.6E-02 9.3E *** 1.8E-02 *** *** 5.1E-03 5.3E-02 I-134 3.8E-02 1.3E-01 1.5E-01 *** *** *** *** 9.0E-03 8.3E-02 1-135 7.4E-02 1.1 E-01 8.6E-02 *** *** *** *** 5.8E-03 6.0E-02 Rb-88 5.0E-01 2.3E-01 1. 0E-01 *** *** *** *** 5.7E-03 1.5E-01 Cs-134 1.8E-02 9.4E-03 1.8E-03 1.9E-02 *** *** *** 2.2E-05 7.7E-03 Cs-136 1.7E-03 1.2E-03 1.1E-04 *** *** *** *** 3. 2E-06 1.9E-04 Cs-137 2.0E-02 1.2E-02 3.1E-03 2.4E-02 *** *** *** 6.7E-05 9.4E-03 N-16 *** *** *** *** *** *** *** *** ***

H-3 1.3E-01 1. 5E-01 *** *** *** 4.4E-02 1.2E-01 2.9E-01 2.5E-01 Na-24 8.8E-03 1.0E-01 1.0E-02 3.6E-03 *** *** *** 9.0E-03 1.4E-02 Cr-51 1.5E-02 2.1 E-03 3.4E-04 *** 1.7E-03 *** *** 3.0E-05 6.4E-03 Mn-54 4.4E-03 2.2E-03 3.9E-05 1.5E-02 1.1E-04 *** *** 1.0E-05 6.8E-04 Fe-55 6.5E-04 1.6E-04 *** *** *** *** *** 2.1E-05 9.1E-03 Fe-59 5.2E-04 6.2E-04 2.3E-04 *** - 6.9E-04 *** *** 1.5E-06 5.2E-04 Co-58 1.4E-02 4.6E-03 6.7E-04 3.6E-03 5.8E-04 *** *** 8.0E-05 2.4E-02 Co-60 1.0E-03 7.8E-04 1.2E-04 3.1E-03 4.7E-04 *** *** 1.6E-05 '9.2E Zn-65 2.6E-03 2.4E-03 1.6E-05 *** *** *** *** 1.7E-06 2.2E-05 S r-89 6.8E-04 7.7E-05 6.8E-07 *** *** *** *** 6.6E-06 ***

Sr-90 4.2E-06 3.4E-06 1.6E-06 *** *** *** *** 5.4E-08 ***

.s .

TABLE 2-10 (continued)

SUMMARY

OF RADIONUCLIDE PRIMARY COOLANT CONCENTRATIONS IN PWR'S*

(pC1/g)

Fort Turkey Indian Yankee Calvert Three Mile ~ Prai rie Ra'ncho Calhoun 1 Zion 1/2 Point 3/4 Pt 2/3 Rowe Cliffs 1 Island 1 Island 1/2 Seco Isotope (1976-1977) (1975-1978) (1974-1978) (1975-1978) (1975) (1976) (1975-1977) (1981) (1979)

Sr-91 *** 3.8E-03 3.7E-04 *** *** *** *** 3.3E-05 7.2E-04 Y-91m *** 8.7E-04 *** *** *** *** *** 5.0E-05 ***

Y-91 5.0E-06 4.4E-06 *** *** *** *** *** 4.3E-07 1.8E-05 Y-93 *** 7.9E-03 2.2E-03 *** *** *** *** 2.1E-04 2.6E-03 Zr-95 1.5E-03 4.2E-04 4.5E-05 *** 2.8E-04 *** *** 4.6E-06 2.9E Nb-95 1.3E-03 2.2E-04 3.8E-05 *** 2.4E-04 *** *** 3.9E-06 4.6E-04 Mo-99 5.7E-03 3.5E-03 8.1 E- 04 *** 5.0E-03 *** *** 1.3E-04 1.7E-03 Tc-99m 4.l E-04 *** 2.7E-06 *** 4.9E-03 *** *** *** ***

Ru-103 5.4E-02 1.8E-04 2.1E-05 *** *** *** *** 6.7E-07 7.0E-05 y

• • • 9.0E-02 *** *** *** *** *** *** ***

4 Ru-106 *** *** *** ***

ui Ag-110m 2.2E-04 3.1E-03 1.lE-05 3.7E-06 9.7E-05 2.1E-04 3.8E-04 1.9E-04 *** *** *** *** 2.0E-06 ***

Te-129m Te-129 *** *** 2.4E-02 *** *** *** *** *** ***

Tc-131m *** 2.l E-03 3.7E-04 *** *** *** *** *** ***.

Te-131 *** *** 7.9E-03 *** *** *** *** *** 7.4E-03 Tc-132 *** 1.8E-04 4.0E-05. *** *** *** *** 1.2E-06 3.1E-05 Ba-140 1.1E-03 1.0E-03 1.lE-04 *** *** *** *** 1.9E-05 2.5E-04 La-140 4.2E-04 1.8E-03 1.3E-04 *** *** *** . *** 1.4E-05 1.1E-04 Cz-141 4.3E-04 1.lE-04 1.7E-05 *** *** *** *** *** 4.6E-05 4.6E-04 4.9E-05 *** *** *** *** 1.9E-05 ***

C2-143 8.2E-03 Cr-144 *** 1.4E-04 1.2E-05 *** 2.6E-05 *** *** - 5.4E-06 4.6E-04 W-187 1.4E-02 3.1E-03 3.0E-04 *** *** -*** *** 1.lE-04 2.9E-03 Np-239 1.2E-02 9.3E-04 1.0E-04 *** *** *** *** 3.7E-06 7.6E-04

• See Footnote of Table 2-9.

• • See Footnote of Table 2-9.

• • • See Footnote of Table 2-9.

t Data unreliable.

-79

. TABLE 2-11

' MONTHLY AVERAGE

• PRIMARY / SECONDARY LEAKAGE'(REF.'8, 39)-

-(gal / day at: 70 F; density = 8.3 -lb/ gal).

'] .1970 Plant ~ J- F M A M- J J -A S 0 N D San Onofre 4 4 ~4 4 3 9 11 8 14 S** -S 0 Connecticut 0- 10. O S 0 0 'l20 10 20 'O 0 0

^ Yankee

-R. E. Ginna 0 -0 0 0. 0- 0 Point Beach 1. 0 1971' Plant' J F M A M J J A S 0 N D

San'0nofre' 0, 0 0 ,, 0 0 0 0 0 0 0' 0 0

. Connecticut 0 30 15 0 0 10 20- 20 15 40 40 40 Yankee

'R. E. Ginna 0 0 S S 0. 0 0 0 0 0 0 0

-H. B. Robinson S S -S S S S 0 50 55 20-Point Beach'l. 0 0 0 10 90 100 53 30 20 20 20 .20 1972

' Plant J F M A M J J A S 0 N D

San Onofre S 0 0 0 0 0 22 0 10 30 4 31 Connecticut 40 -40 40 40 40 S 0 0 0 'O 0 0 Yankee

~ R. E. Ginna 0 0 0 S S- 0 0 0 0 S 0 0

. H. B. Robinson 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 Leakage values listed begin with the first year.of commercial operation.

• • Shutdown:not included in average.

. NA - Not Available.

2-26

t- ,

.)

l TABLE 2-11L(contin 5ed)

MONTHLY AVERAGE

• PRIMARY / SECONDARY LEAKAGE (gal / day at 70 F; density = 8.3 lb/ gal) 1973 Plant J- F M."A M' 'II J A' 'S 0 N D.

. San Onofre- 3 3 0 0 0 0 0 0 0 0 S S Connecticutz 0- 0 0 0 10 E 0 S S -S S 0

' Yankee R. E. Ginna 0 0 0 0 0 0 0 0 0 -0 0 0 H 'S. Robinson 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 1974 Plant J F' M A M J- J A S 0 'N D

. San Onofre- 0 44 60 60 0 0 0 0 0 2 2 2 Connecticut 0 0 0 S 0. 0- 0 0 0 0 0 0

' Yankee 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 l

' Turkey Point 4 S 0 0 0 0 0 0 22 0 0 0 0 Zi on ' t S S~ S 0 0 0 S S 0 0 0 0

. Zi on .2 - 0 0 .0 0 Indian Point 2 0 0 0 0 0 Prairie Island 1 0 0 'O O O O 2-27

e--

~

igh ..

TABLE 22-11 (continued)

MONTHLY AVERAGE

• PRIMARY / SECONDARY LEAKAGE

-(gal / day at 70*F; density = 8.3.lb/ gal)

\

1975 Plant J F M. A. M J J A S 0 N D .

San Ong,fre .2 2 2 .2 3 -5 0 0 0 0 0 0 Connecticut- 0 0 0 0 0 S 0 0 0 0 0 0 Yankee 1

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- ,

Su rry -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 01 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 'O O O 0 0 0 0 0 0 0 0 0

4. Cook 1 0 0 0 0 t

4 2-28

~

(

(. ,

~

. TABLE 2-11. (continued)

MONTHLY AVERAGE

• PRIMARY / SECONDARY LEAKAGE ,

4 (gal / day at 70*F; density = 8.3 lb/ gal)

, 1976 Plant J- F M A -M -J - J .A S 0 N D

'46 San Onofre 0 0 -O 0 0 0- 0 0 S S S Connecticut 0 0 0 0' S S 0 0 0 S 0 0

. Yankee 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. O 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 2-29

l ,

' ^

TABLE 2-11 (continued) ~

l MONTHLY' AVERAGE

• PRIMARY / SECONDARY LEAKAGE .I (gaifday at 70*F; density = 8.3 lb/ gal)

I 1977 Plant J F M A M- J J A S 0 N D-San Onofre -S- -S S~ 0 0 1 2 2 S' "O 2 1 r- -Connecticut O' 0- 0-O O 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 'O 0 1 0 6 41 52 Pof_nt Beach 1 4 5 3 6 3 5 5 5 4 5 8 7 Point Beach 2. 25 35 33' S 0 0 0- 0 0 0 0 01  !

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 15 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 Salen 1 0 0 0 S S- .0 Farley 1. O i

<r e

t.

4 TA,BLE 2-11 (continued)

MONTHLY ~ AVERAGE

• PRIMARY / SECONDARY LEAKAGE (gal / day at 70 F; density = 8.3 lb/ gal).

1978 v- Plant J F M A M J Average,* gal / day San Onofre 1 1 1 S 1 1 4.6 Connecticut 0 'O 0 0 0. 0 5.7 Yankee 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 11 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 O 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

• 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.

2-31

^; _,: .

f 4;

TABLE 2-12 MOISTURE CARRY 0 VERS IN RECIRCULATING U-TUBE

, STEAM GENERATORS

• Facility Percent Carryovers ' Reference

.Pf~11sades 0.08 10, 11

.,. ~

-Kansai 0.05 10, 11 Point Beach '

O.2 8, 12 Turkey Point 3 0.6 6

. . Turkey Point 4 1.6 6 Average 0.5

• Measurement based on Na concentration.

4 :.

I  :

I

(

i 2-32 i <

The category "Other Nuclides" includes Mo, Y, and Tc which are generally present in colloidal suspensions or as " crud." Although the actual removal mechanism for Y, Mo, and Tc 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) and consequently plateout of these nuclides is included in the parameters for ion exchange.

2.2.4 IODINE RELEASES FROM BUILDING VENTILATION SYSTEMS 2.2.4.1 Parameter The iodine releases f rom building ventilation systems prior to treat-ment are calculated by the PWR-GALE Code using the data in Tables 1-1, Tables 2-2 through 2-8 and 2-13 through 2-16.

2.2.4.2 Bases The iodine-131 releases from building ventilation systems are based on measurements made at a number of operating reactors. The measurements were made during routine plant operation and during plant shutdowns. Work on identifying sources of radiciodine at PWR's has been conducted by C. Pelletier, et al. (Ref. 7) for the Electric Power Research Institute (EPRI), at three operating PWR's; Ginna, Calvert Cliffs 1, and Three Mile Island 1. Measurements have also been made by EG&G Idaho, Inc., Allied Chemical Corp., Idaho National Engineering Laboratory, for the U. S. Nuclear Regulatory Commission at Fort Calhoun (Ref. 4), Zion 1 and 2 (Ref. 5),

Turkey Point 3 and 4 (Ref. 6), Prairie Island (Ref.42), and Rancho Seco (Ref. 43).

These measurements indicate that iodine-131 building vent releases are directly related to the reactor coolant iodine-131 concentration. As i 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 pCi/g to give a " normalized" release rate of iodine-131 in Ci/yr/pC1/g as shown in the following equation:

R

=

A R

N C pg where R = normalized release rate of iodine-131, Ci/yr/pCi/g N

R A

= absolute (measured) iodine-131 release rate, Ci/yr CRW = measured reactor water iodine-131 concentration, pCi/g 2-33 ,

a TABLE 2-13 ANNUAL I0 DINE NORMALIZED RE' LEASES FROM CONTAINMENT VENTILATION SYSTEMSt NORMAL OPERATION LEAK RATE

• Normalized Release / Unit Data Source 10-3 %/ day

.Ft.'Calhoun (Ref. 4) 0.0014 Three Mile Island l'(Ref. 7) 2.5 Turkey Point 3/4 (Ref. 6) 0.9 Main' Yankee (Ref.16) 0.1 Ginna (Ref.19) 0.064 r Yankee Rowe (Ref.14,16) 1.0 Prairie Island 1/2 (Ref. 42) 0.005 Rancho Seco (Ref. 43) 2.56 Average 0.80 RELEASE FOR EXTENDED OUTAGES **

Normalized Release / Unit Data Source (Ci/yr/u ci/g)

Three. Mile Island 1 (Ref. 7) 0.44 Calvert Cliffs 1 (Ref. 7) 0.19 Average 0.32

The normalized release rate, expressed in %/ day of leakage of primary coolant inventory of iodine, 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 radiciodine between the water phase in the leakage and the gas phase where it is measured.

• • The normalized release rate, expressed in Ci/yr/pCi/g, represents the effective ^1eak rate for radioiodine. It is the combination of the reactor water iodine 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.

t. These results were obtained using 131 1 data. The normalized release rates are applicable to both 131 1 and 133g ,

2-34

TABLE 2-14 ANNUAL I0 DINE NORMALIZED RELEASES

• FROM AUXILIARY BUILDING VENTILATION SYSTEMSt NORMAL OPERATION Normalized Release / Unit Data Source (Ci/yr/pC1/g)

Zion 1/2 (Ref. 5) 1.0 Fort Calhoun (Ref. 4) 0.12 Ginna (Ref. 7) 0.032 Calvert Cliffs 1 (Ref. 7) 0.57 Three Mile Island 1 (Ref. 7) 0.034 Turkey Point 3/4 (Ref. 6) 1.85 l

Prairie Island 1/2 (Ref. 42) 0.013 Rancho Seco (Ref. 43) 0.97 Average 0.68 SHUTDOWN Normalized Release / Unit Data Source (Ci/yr/pCi/g)

Ginna (Ref. 7) 0.08 Calvert Cliffs 1 (Ref. 7) 0.016 Three Mile Island 1 (Ref. 7) 0.14 Turkey Point 3/4 (Ref. 6) 6.8 Rancho Seco (Ref. 43) 1.14 Average 2.50 L

• The normalized release rate, expressed in C1/yr/pCi/g during different modes of operation, represents the effective leak rate for radioiodine.

It is the combination of the reactor water iodine leakage rate into the buildings and the partitioning of the radiciodine between the water phase in the leakage and the gas phase where it is measured.

131 t These results'were obtained using 1 data. The normalized release 131 133 g, rates are applicable to both I and 2-35

TABLE 2-15 --

ANNUAL 10 DINE' NORMALIZED RELEASES *

.FROM REFUELING AREA VENTILATION SYSTEMSt i

NORMAL OPERATION Normalized Release / Unit Data Source (C1/yr/uCi/g)

. Ginna (Ref. 7) 0.008

, Calvert Cliffs 1 (Ref. 7) 0.049 Three Mile Island 1 (Ref. 7) 0.0012 Turkey Point 3 (Ref. 6) 0.16 Prairie Island 1/2 (Ref. 42) 0.01 9 RanchoSeco(Ref.43) 0.01 Average 0.038 SHUTDOWN Normalized Release / Unit Data Source (Ci/yr/uci/g)

Ginna(Ref.7) 0.014 Calvert Cliffs 1 (Ref. 7) 0.039 Three Mile Island 1 (Ref. 7) 0.06

, Turkey Point 3 (Ref. 6) 0.05 Rancho Seco (Ref. 43) 0.30 Average 0.093

• The normalized release rate, expressed in Ci/yr/pci/g during different modes of operation, represents the effective leak rate for radiciodine.

It is the combination of the reactor water iodine 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.

t These results were obtained using 131 I data. The normalized release rates are applicable to both 131 1 and 133 g, 2-36

,, , TABLE 2-16* . ,

" ANNUAL' IODINE NORMALIZED RELEASES ** i FROM TURBINE BUILDING VENTILATION SYSTEMSt

_a.

' NORMAL OPERATION .

l Normalized Release / Unit .t

' Data' Source (Ci/yr/p Ci/g)

Monticello' 3.1 x 10 3 .

Oyster Creek - 6.0 x 10 3 Vermont Yankee 0.35 x 10 3 Pilgrim 8.5 x 10 3 ,

Browns Ferry 1.3 x 10 3 3

References 3, 5 of Ref.15 3.3 x 10 Average 3.8 x 10 3  :

EXTENDED SHUTDOWN Normalized Release / Unit Data Source (Ci/yr/pCi /g)

Monticello 1.7.x 10 2 Oyster Creek 3.5 x 10 2

'. Vermont Yankee 0.63 x 10 2 Browns Ferry 1.3 x 10 2 References 3, 5 of Ref.15 1.4 x 10 3  ;

Average 4.2 x 10 2  ;

• The data in this table are taken from Table 2-8, NUREG-0016, Revision 1, January 1979 (Ref.15). [

, .** The normalized release rate, expressed in C1/yr/pCi/9 during different -

modes of operation represents the effective leak rate for radiciodine.

It is a function of iodine leak rate via steam and the partition r coefficient for.radiciodine from reactor water to steam in the reactor

• vessel.

t These' results were obtained using 131 1 data. The normalized release rates are applicable to both 131 1 and 133g ,

2-37

.y .

r

5 r, f

w7 ^ The normalized reactor water release rate, expressed in Ci/yr/pCi/g ,

represents an effective leak' rate for reactor water containing iodine. 1

-It-is the combination of the water leakage rate into the building and )

the effect of iodine partitioning between the water phase in the systems  ;

leakage and the vapor. phase in the building atmosphere.  ;

' For the turbine building, the secondary coolant iodine releases are <H eectly related to the secondary coolant iodine-131 concentration.

Therefore, for the turbine building, the normalized-iodine release, R ' N is detennined using the following expression:

1 R

' A RN~" C gg x PC .

. where  !

R N

= normalized release rate of secondary coolant water containing iodine-131, Ci/yr/pci/g

! Rg = absolute (measured) iodine-131 release rate, Ci/yr

, Cpy = measured secondary coolant iodine 131 concentration, pCi/g

PC = measured iodine partition coefficient from secondary coolant water to steam.

1 i

i The normalized release rate is used to estimate the release from PWR's' 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 PWR's.

i o Data on normalized release rates from the three reactors used in ,

c the EPRI study and the five reactors used in the NRC sponsored study are l given for normal operation and shutdown periods in' Tables 2-13 through i 2-15, for the containment building, auxiliary building and refueling ',

a rea, - respectively. Also given in Table 2-13 is the normalized value of L the iodine release data discussed in NUREG-0017, April 1976 (Ref.14).

For Table 2-16, it was considered that since the basic design and operation - 1 1- of PWR and BWR power generation equipment which is housed in the turbine i building is essentially identical, the turbine building leakage rates i f from PWR's and ra?'s should be similar. Therefore, for the PWR turbine building nonnalized iodine release rate, the values for BWR's given in Table 2-15 of NUREG-0016, Revision 1 (Ref.15) have been used and ,

reproduced here as Table 2-16 of this report. l The data in Tables 2-14 through 2-16 are expressed as total  !

I normalized releases during power operation of 300 days an<! the total i normalized releases during shutdowns of 65 days. Since the reactors j used in the EPRI study and the NRC study experienced several intermittent l' l 2-38 o

r shutdowns of short duration during the power operation measurement period, the iodine releases during these short duration outages are included under power operation.

Since the releases from the containment building are dependent on the method of containment purging (see Section 2.2.9, Containment Purging Frequency), the releases in Table 2-13 are expressed in terms of a leak rate (in %/ day of primary coolant inventory). In addition, the release f rom the containment building during extended outages is expressed as a total normalized release as discussed above for other buildings.

In order to obtain the releases in curies / year from the auxiliary building and the refueling area of a particular PWR, the normalized release data in Tables 2-14 and 2-15, respectively, are multiplied in the PWR-GALE Code by the lodine concentrations in the reactor coolant for that particular PWR using the following expression:

RPWRi = RN xC pygj where RPWRi = calculated annual release rate for particular PWR for iodine isotope 1, Ci/yr R

N = n rmalized annual release rate of iodine from Tables 2-14 and 2-15, C1/yr/pCi/g CPWRi = calculated reactor water concentration for particular PWR for iodine isotope 1, pCi/g To obtain the release in curies / year from the turbine building of a particular PWR, the normalized release data in Table 2-16 are multiplied in the PWR-GALE Code by the iodine concentration in the secondary coolant water and the iodine partition coefficient from the water to steam in the steam generator for that particular PWR using the following expression:

RPWRi = RNx SCPWRi x PCPWR where RPWRi = calculated annual release rate for particular PWR for iodine isotope 1, Ci/yr R

N

= n rmalized annual release rate of iodine from Table 2-16, Ci/yr/pCi/g 2-39

SCPWRi = calculated secondary coolant concentration for particular  !

PWR for iodine isotope 1, pC1/g '

PC PWR

= partition coefficient from the secondary coolant water to steam for the particular PWR (see Table 2-6)

In order to obtain the releases in curies / year from the containment building of a particular PWR, the normalized leak rates in Table 2-13, are multplied in the PWR-GALE Code by the iodine 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.

To obtain the releases during shutdown, multiply 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 prior to shutdown.

lodine released from PWR building ventilation systems appear in one of the following chemical forms: particulate, elemental, hypoiodious acid (H01) and organic. Based on data in References 4, 5, 6, 7, 42 and 43, the fraction of the iodine appearing in each of the chemical forms for each building ventilation system is given below:

FRACTION OF I0 DINE APPEARING IN EACH CHEMICAL FORM FROM PWR BUILDING VENTILATION SYSTEMS Containment Auxiliary Turbine Fuel Handling Particulate 0.09 0.04

• 0.01 Elemental 0.21 0.21 0.78 0.17 0.22

• 0.57 H01 0.21 Organic 0.49 0.53

• 0.25

'* No data on breakdown of other species.

l 2.2.5 RADI0 ACTIVE PARTICULATES RELEASED IN GASE0US EFFLUENTS 2.2.5.1 Pa rameter Use the radioactive particulate release rates in gaseous effluents given in Table 2-17.

t 2-40

TABLE 2-17

^ PARTICULATE RELEASE RATE FOR GASE0US EFFLUENTS

• 1 (Ci/yr)/ unit Auxiliary Fuel Pool Waste Gas Nuclide Containment Building Area' System C r-51 9.2 - 3.2(-4) 1.8(-4) 1.4(-5)-

Mn-54 5.3 - 7.8(- 3.0(-4) 2.1(-

Co-57 8.2 - NA NA NA Co 2.5 - 1. 9 - 2.1(-2) 8.7 -

Co-60 '2.6 - 5.1 - 8.2(-3) 1.4 -

Fe-59 2.7 - 5.0 - NA 1.8 -

S r- 89*

• 1.3 - 7.5 - 2.1 - 4.4 -

S r- 90*

• 5.2 - 2.9 - 8.0 - 1.7 -

Zr-95 NA 1.0 - 3.6 - 4.8 -

Nb-95 1.8(-3) 3.0 - 2.4 - 3.7 -

Ru-103 1.6(-3) 2.3(-5 3.8 - 3.2 -

Ru-106 NA 6.0(-6 6.9 - 2.7 -

Sb-125 NA 3.9 -6 5.7-5) .NA Cs-134 2.5 - 5.4 - 1.7(-3) 3.3 -

Cs-136 3.2 - 4.8 - NA 5.3 -

Cs-137 5.5 - 7.2 - 2.7(-3)- 7.7 -

Ba-140 NA 4.0-4) NA 2.3 -

Ce-141 1.3(-3) 2.6 -5) 4.4(-7) 2.2 -

Particulate release rates are prior to filtration.

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

• • Data not available from Ref. 4, 5, 6 or 7, therefore Sr-89 and Sr-90 data were extracted from Semi-annual Effluent Release Reports. Release from each area above calculated by use of percent released from each area from Ref. 4, 5, 6 and 7 data.

f 2-41

2.2.5.2 B'a ses s

c Tables' 2-18 through 2-21. list measured particulate. releases at 12 oper'ating reactors (Ref. 4, 5, 6, 7, 42, and 43). 'The average annual release rates for each nuclide released from four sources within the plant have been calculated based on the data in Tables 2-18 through 2-21.

The measurements shown-in ' Tables 2-18 through 2-21 were taken upstream of HEPA filters on streams on which HEPA filters are located. Based on the data in Tables 2-18 through 2-21, 63% of the releases came from the containment, 5% from the auxiliary building, 31% from the fuel pool area, and less than 1% from the waste gas processing system.

2.2.6 N0BLE GAS RELEASES FROM BUILDING VENTILATION SYSTEMS

,y .

2.2.6.1 Pa rameter The noble gas releases from building ventilation systems are based on a daily leak rate of 3% of the noble gas inventory in the primary coolant released to the containment atmosphere; on a 160 lb/ day primary coolant leak to the auxiliary building; and on a 1700 lb/hr steam leak rate in the turbine building.

2.2.6.2~ Bases

'The containment building leakage rate is derived from xenon-133 measurements in the containment atmosphere at Ginna and Maine Yankee (Ref.17). The xenon-133 concentrations in the containment atmospheres at steady state were approximately 5 x 10-3 pC1/cc for Main Yankee and 7 x 10-3.pC1/cc-for Ginna. -The containment volumes at these fa'cilities are approximately 1.8 x 106 ft3 for Maine Yankee and 1 x 106 ft3 for Ginna. Based on these values, the total microcuries of xenon-133 in the containment building atmosphere are Maine Yankee 6 3 4 3 8 (5 x 10-3 C1/cc)(1.8 x 10 ft)(2.83x10 cc/ft ) = 2.5 x 10 Ci Xe-133 Ginna 4 0 (7x10-3 C1/cc)(1 x 106 ft)(2.83x10 3 cc/ft ) =3 2.0 x 10 pCi Xe-133 Based on the half-life of xenon-133 (5.3d) 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 Main Yankee

• 7 p ! ha o.fd)=3.3x10 pCi/ day Xe-133 leakage 2-42

J g hR TABLE 2-18 MEASURED RELEASES UPSTREAM OF HEPA FILTERS - CONTAlletENT.

(Ci/yr)

Turkey Prairie d

Three Mile Pcint- Calvert Island Rancho 3 , Island 1 Fort Calhoun Zion 1 & 2 3&4 Cliffs 1 Ginna 1&2 Seco ' 'Avera9e

. Nuclide (Ref. 7) (Ref. 4) (Ref. 5) (Ref. 6) (Ref. 7) (Ref. 7) (Ref. 42) (Ref. 43) (Ci/yr)/ unit Cr-51 5.5(-2) ND HD ND NA NA NA .NA 9.2(-3)-

i Mn-54 2.1(-2) 1.4(-8) 3.9(-6) NA NA NA NA NA - 5.3(-3)

Co-57 4.9(-3) ND ND: ND NA NA NA . NA 8.2(-4)' -

4 Co-58 2.2(-1) 5.6(-8) 1.5(-5) 3.2(-6) NA NA 6.6(-8) 2.5(-3) 2.5(-2)

Co-60 2.3(-2) 3.8(-8) 1.2(-5) 3.0(-5) NA NA 1.4(-7) 3.3(-4) 2.6(-3)

' Fe-59 1.6(-2) ND ND ND NA NA NA NA 2.7(-3)

? Zr NA NA NA NA NA NA ~NA NA . NA

, g Nb-95 1.1(-2) ND .ND ND NA NA - NA 1A 1.8(-3) i Ru-103 9.5(-3) ND ND ND NA NA NA: NA 1.6(-3)

Ru-iO6 NA NA NA NA NA NA NA NA NA -!

Sb-125 NA NA NA NA NA -NA NA NA NA '

, Cs-134 2.1 (-2) 3.2(-6) 2.3(-4) 7.7(-5) NA NA -3.2(-8) 1.5(-3) 2.5(-3)

Cs-136 1.9(-2) ND ND ND NA NA NA NA 3.2(-3) .j I

Cs-137 4.4(-2) 4.1(-6) 3.2(-4) 1.9(-4) NA NA 6.6(-8)- 5.0(-3) 5.5(-3)

Ba-14D NA NA NA NA NA NA NA NA NA Co-141 8.0(-3) ND ND ND NA NA NA NA 1.3(-3)

I i 'l MD = Not detected. For averaging purposes, a value of zero was assumed.

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

1 i

. _ , . _. - - __ . . _ - ,,. - m -. . _. ,. -

_+

4

<A TABLE 2-19 MEASURED RELEASES UPSTREAM OF HEPA FILTERS - AUXILIARY BUILDING (C1/yr)

Turkey Prairie Three Mile Point Calvert Island Rancho Island 1 Fort Calhoun Zion 1 & 2 3&4 Cliffs 1 Ginna 1&2 Seco Average Nuclide (Ref. 7) (Ref. 4) (Ref. 5) (Ref. 6) (Ref. 7) (Ref. 7) (Ref. 42) (Ref. 43) (C1/yr)/ unit Cr-51 1.4(-3) MD MA ND MA 1.9(-4) MA NA 3.2(-4)

Mn-54 1.1(-4) MA NA 6.3(-5) 3.0(-4) 6.7(-5) ,2.7(-6) MA- 7.8(-5)

Co-57 NA NA NA NA NA NA NA NA NA-Co-58 1.1(-3) 2.0(-3) NA 1.1(-3) 4.8(-4) 6.3(-4) 4.0(-5) 1.2(-2) 1.9(-3) '

Co-60 2.0(-4) 2.7(-4) NA 6.0(-4) -2.0(-3) 7.7(-4) 4.5(-5) 7.3(-4) 5.1(-4)

Fo-59 2.3(-4) MD ND ND NA 1.9(-5) NA NA 5.0(-5)

Zr-95 2.7(-4) MD ND ND 7.9(-3) 4.1(-5) 5.7(-6) NA 1.0(-3) m Nb-95 1.4(-4) ND ND ND NA 6.0(-5) 1.0(-5) NA 3.0(-5)

{Ru-103 9.1(-5) MD NA ND NA 6.9(-5) 2.7(-6) NA 2.3(-5)

Ru-106 NA ND MA ND NA 2.4(-5) NA . NA 6.0(-6)

Sb-125 NA NA NA NA NA NA 7.7(-6) NA 3.9(-6) l Cs-134 8.0(-5) 1.6(-3) NA 7.9(-4) 2.0(-3) 3.4(-4) 1.5(-6) 5.2(-5) ' 5.4(-4 l Cs-136 NA ND NA MD NA 1.9(-4) NA NA 4.8(-5 Cs-137 2.0(-4) 1.8(-3) MA 1.4(-3) 1.9(-3) 1.1(-3) 9.4(-6) 8.0(-5) 7.2(-4 Ba-140 NA ND ND ND NA 1.6(-3) NA NA 4.0(-4) l Ce-141 1.5(-4) MD NA ND NA 2.8(-5) 1.5(-6) MA 2.6(-5)

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

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

Measurecents were made downstream of the auxiliary building HEPA filter. Due to uncertainty in the DF's of the HEPA filter, the data is not considered.

l

TABLE 2-20 MEASURED RELEASES UPSTREAM OF HEPA FILTERS - FUEL' POOL AREA (C1/yr)/ unit Turkey Prai rie -

Three Mile Point Calvert . Island. Rancho Island 1- Fort Calhoun Zion 1 & 2 3&4 Cliffs 1 & 2 Ginna 1&2 . Seco Nuclide (Ref. 7) (Ref. 4) (Ref. 5)- (Ref. 6) (Ref. 7) (Ref. 7)- (Ref. 42) (Ref. 43) Average Cr-51 1.8(-4) MA NA NA NA NA NA NA 1.8( 8)

Mn-54 1.0(-5) NA NA NA 1.2(-3) MA .2.6(-6) NA 2.4(-4)

Co-57 NA NA NA .NA NA NA NA NA NA Co-58 8.5(-5) NA NA NA 1.1(-2) NA 8.8(-6) 6.7(-5) 1.8(-3)

Co-60 4.4(-5) M MA NA 5.0(-3) NA 6.9(-6) _ 7.6(-6) 8.4(-3)

Fo-59 NA NA - NA NA NA NA NA NA NA Zr-95 -NA NA NA .NA NA NA 7.2(-6) NA 3.6[-6)

Mb-95 3.0(-5) MA NA NA 9.5(-3) NA 1.7(-5) NA 1.9L-3) y Ru-103 9.8(-5) NA NA .NA NA NA 1.7(-5) NA g R3-106 3.8(-5) 6.9(-5) NA NA NA NA NA NA NA 6.9q-5)

Sb-125 1.7(-4) MA NA NA NA NA ND NA 5.71,-5)

Cs-134 9.0(-6) MA NA NA 2.2(-3) NA 9.8(-7) 9.6(-7) -3.7(-4)

Cs-136 NA NA NA NA NA NA NA NA NA Cs-137 2.4(-5) NA NA NA 5.6(-3) NA 4.1(-6) 7.4(-7) 9.4(-4)

Ba-140 NA NA NA NA NA NA NA NA NA Ce-141 NA NA NA NA NA NA 8.8(-7) NA 4.4(-7)

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

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

_ _ _ ___ _ _ = ..- -

TABLE 2-21 MEASURED RELEASES UPSTREAM OF HEPA FILTERS - WASTE GAS SYSTEM (Ci/yr)

Tu rkey Prairie Three Mile Point Calvert Island Rancho Island 1 Fort Calhoun Zion 1 & 2 3&4 Cliffs 1 _Ginna 1&2 . Seco Avera ge Nuclide (Ref. 7) (Ref. 4) (Ref. 5) (Ref. 6) (Ref. 7) (Ref 7) (Ref. 42) (Ref. 43) (Ci/yr)/ unit Cr-51 8.4(-5) ND ND ND NA NA NA NA 1.4(-5)

Mn-54 1.1(-5) ND 4.0(-6) ND NA NA NA 8.4(-9) 2.1(-6)

Co-57 NA NA NA NA NA NA NA NA NA Co-58 4.5(-5) 3.8(-6) 1.1(-5) 8.8(-7) NA NA NA 5.1(-8) 8.7(-6)

Cc-60 8.0(-5) NA 2.2(-6) 2.9(-7) NA NA NA 5.9(-8) 1.4(-5)

Fa-59 7.2(-6) 1.8(-6) 1.9(-6) ND NA NA NA NA 1.8(-6) m Zr-95 1.9(-5) ND ND NA NA NA NA NA 4.8(-6) i Nb-95 2.2(-5) ND ND ND NA NA HA NA 3.7(-6)

• Ru-103 1.9(-5) ND ND ND NA NA NA NA 3.2(-6)

Ru-106 1.6(-5) ND ND ND NA- NA NA NA 2.7(-6)

Sb-125 NA NA NA NA NA NA NA NA NA Cs-134 1.2(-4) 1.2(-6) 1.1(-4) 3.8(-8) NA NA NA 1.1(-8) 3.3(-5)

Cs-136 3.2(-5) NO ND ND NA NA NA NA 5.3(-6)

Cs-137 3.5(-4) NA 1.1(-4) 8.8(-8) NA NA NA -2.6(-8) 7.7(-5)

Ba-140 1.4(-4) ND ND ND NA NA NA NA 2.3(-5)

Cc-141 1.3(-5) ND ND ND NA NA NA NA 2.2(-6)

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

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

n

~

Ginna I

-(5 3 y0 3) = 2.6 x 107 ' Ci/ day-Xe-133 leakage .

.i Based on the xenon-133 concentration during power operation (Ref. 29) I and the masses of primary coolant of the two plants, the fraction of the

~

xenon-133 inventory in the containment released per day is Maine Yankee 7

3.3 'x 10 C1 day

= 0.01/ day = 1%/ day (10 pCi/cc x 28,300 cc/ft x.11,000 ft 3)

Ginna 7

. 2.6 x 10 Ci/ day

, = 0.005/ day - 0.5%/ day

.(30 pCi/cc'x28,300x.6,234ft)3 i

Reference 16 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 .two -orders of nagnitude. - The percent of xenon-133 inventory in the coolant released to the containment atmosphere varied from approximately 0.05%/ day to 0.5%/ day. Also from Ref. 43, this percent was determined to be 10.4.for. Rancho Seco.

On the basis of these data, we consider that 3%/ day of the noble

. gas inventory ~.in the primary coolant 'is released to the containment atmosphere.

.i In the auxiliary building, the source term calculation is based

( on an assumed primary coolant leakage rate of 160 lb/ day (20 gal / day).

In the ' absence of;available data, this value is based on engineering judgment and is consistent with values proposed in Environmental Reports.

I

-The reactor coolant concentrations for Xe-133 are measured values during 12/73 - 6/74 for Main Yankee and September and October of 197.1'.

for Ginna (Ref.16).

2-47

,- - -- ~ . . , , . .s , , _ . . . ,

In the turbine building, it is assumed that steam will leak to the turbine building atmosphere at a rate of 1700 lb/hr. The leakage is considered to be from many sources, each too small to be detected individually, but which, taken collectively, total 1700 lb/hr. The most significant leakage pathway is considered to be leakage through valve stem packings.

2.2.7 STEAM GENERATOR BLOWDOWN FLASH TANK VENT 2.2.7.1 Parameter

, le Pressurized water reactors, 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, iodine releases by this path are negligible and a partition factor of zero is used for the steam generator blowdown flash tank vent.

2. For older plants which still utilize flash tanks which vent directly to the atmosphere an iodine partition factor of 0.05 is used.

2.2.7.2 Bases Approximately one-third of the blowdown stream flashes to steam in the flash tank, provided there is a heat balance between steam generator operating conditions (550 F,1000 psia) and the blowdown flash tank conditions (240'F, sat.). Although the iodine species in the blowdown stream will be predominantly nonvolatile (volatile species are degassed '

in the st?am generator), significant iodine removal will occur because of entrainment by the flashing steam. A steam quality of 85% is considered in the evaluation. For currently designed PWR's which have provisions to prevent flashing (cooling blowdown below 212*F) or to condense the steam leaving the flash tank, the entrainment losses will be negligible, i.e., a partition factor of zero.

2.2.8 IODINE RELEASES FROM MAIN CONDENSER AIR EJECTOR EXHAUST 2.2.8.1 Parameter The iodine releases from the main condenser air ejector exhaust prior to treatment are calculated by the PWR-GALE Code using the data in Tables 2-2 through 2-8, and in Table 2-22.

2.2.8.2 Bases l

The iodine releases from the main condenser air ejector exhaust are based on secondary side measurements made by EPRI at Point Beach 2, (Ref.

=

7), by EG&G Idaho, Inc., for the NRC, at Turkey Point 3 and 4 (Ref. 6),

and by Westinghouse at Point Beach 1 (Ref.12) and Haddam Neck (Ref. 38).

2-48

-TABLE 2-22 ANNUAL IODINE NORMALIZED RELEASES FROM MAIN CONDENSER AIR EJECTOR EXHAUST

• Normalized Release Data Source (Ci/yr/pCi/g)

Turkey Point 3/4 (Ref. 6) 3.5 (+3)

Point Peach 1/2 (Ref. 7, 12) 6.1 (+2)

Haddam Neck (Ref. 38) 3.0 (+1)

Average 1.7 (+3)

• The normalized release rate represents the effective release rate for radiciodine. It is the combination of the steam flow to the main condenser, the partitioning of radiciodine between the main condenser and the air ejector exhaust where it is measured and the partition coefficient for radiciodine from water to steam in the ' steam generator.

N' l

2-49

v In a manner similar to the discussion of normalized releases for I

building ventilation releases in Section 2.2.4, the main condenser air ejector exhaust iodine releases are directly related to the secondary coolant iodine-131 concentration. Therefore', for the air ejector

- exhaust, the normalized iodine release, Rg , is determined using the following expression:

A N"CRW x PC where R = normalized effective release rate of iodine-131, Ci/yr/pCi/g N

R = measured (absolute) iodine-131 release rate, Ci/yr A

CRW = measured secondary coolant iodine-131 concentration, pCi/g PC = measured iodine partition coefficient from secondary coolant water to steam in the steam generator.

Data on normalized release rates from the main condenser air ejector exhaust are given in Table 2-22. To obtain the release in curies / year from the air ejector exhaust of a particular PWR, the normalized release

data in Table 2-22 are multiplied in the PWR-GALE Code by the iodine concentration in the secondary coolant water and the iodine partition coefficient from the water to steam for that particular PWR using the

, following expression:

RPWRi = RgxCPWRi x PCPWR where R PWRi = calculated annual release for particular PWR for iodine isotope i, C1/yr R = n rmalized annual release rate of iodine from Table 2-22, N

Ci/yr/pC1/g CPWRi = calculated secondary coolant concentration for particular I PWR for iodine isotope i, pCi/g 1

PCPWR = 1 dine partition coefficient from water to steam in the steam generator for the particular PWR (see Table 2-6) ,

i 2-50

As discussed in references 6 and 7, most of the iodine in the secondary system is not available for release to the main condenser air ejector exhaust due to iodine bypassing the condenser hotwell in the moisture separator / reheater drains and extraction steam, and possibly due to iodine plating out in the moisture separator / reheater, turbine and main condenser. As a result, the iodine release from the main condenser air ejector exhaust is small compared to the building ventilation releases.

2.2.9 CONTAINMENT PURGE FREQUENCY 2.2.9.1 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 2 purges per year at cold shutdown with the large containment purge lines. The continuous ventilation rate used in the evaluation is based on the applicant's design.

For older plants (those under review for operating licenses or those for which the construction permit SER was issued prior to 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 use of large containment purge lines and of 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 in the containment) to reduce the frequency of containment purging will be considered on a case-by-case basis.

2.2.9.2 Bases 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 2-23) has indicated 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 (Ref.17). For the above reasons, new plant designs are to include the capability to purge the containment continuously through small-diameter purge lines (about 8 inches in diameter) and only use the large containment purge lines at cold shutdowns and refueling outages (Ref.18). On this basis, source term calculations for new plants should consider a continuous ventilation rate based on the applicant's containment design, along with the two cold shutdown purges per year with the large containment purge lines, unless special provisions are made to eliminate or reduce the need for continuous ventilation flow.

2-51 l

TABLE 2-23 PWR CONTAINMENT PURGING AND VENTING EXPERIENCE (REF.17) i

. Yankee Rowe y Purge and vent only after cooldown following shutdown Reasons: Routinely pressurize containment for leak detection system checks and bring activity down Duration: 2 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

' Maine Yankee Purge once per. quarter Reason: Bring activity down Duration: 2 to 3 days.each quarter Indian Point 2 Vent 2 times each day Reason: Pressure balance control Duration: Approximately 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Purge once every 2 weeks. (duration not stated)

Three Mile Island 1 Purge approximately once per week during operation, always purge prior to shutdown Reason: Improve temperature and humidity conditions Duration: Approximately 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Connecticut-Yankee Purge .Cannot purge during operation, only during shutdown Reason: Primarily to remove activity Duration: 1 to 2 days-San Onofre'1 Purge each cooldown approximately 4 times per year, no purging during power operation Purge .for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, ventilate during entire shutdown period.

Oconee 1 Continuous purge from startup through 7/1/74 Purged twice since 7/1/74, once on 7/8/74 for several days and again on 8/22/74 for 1 to 2 days Reason: Reduction of gaseous activity for maintenance, etc.

I 2-52

7 -

TABLE 2-23 (continued)

PWR CONTAINMENT PURGING AND VENTING EXPERIENCE (REF. 17)

Oconee 2 s

Continuous purge since startup, lowest purge rate approximately-20,000 ft 3/ min

' Reason: Reduction of gaseous activity for maintenance, etc.

Robinson 2 Purge approx'imately 20 times per year for 2 minutes each purge for testing of purge valves. In addition, purge approximately 10 times per year for~ an average of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> each purge for-personnel comfort reasons.

Vent about 75 times per year for about four hours each. Venting occurs to control containment pressure and to bring containment pressure to zero gauge prior to purging as noted above.

Turkey Point 3 For period 1/1/74 to 7/1/74 Total purges 14 Total time 502 hours0.00581 days <br />0.139 hours <br />8.300265e-4 weeks <br />1.91011e-4 months <br />

• Maximum duration (1 purge) 253 hours0.00293 days <br />0.0703 hours <br />4.183201e-4 weeks <br />9.62665e-5 months <br /> Minimum duration (1 purge) 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> Infrequent purges or vents of 10 minutes for pressure control.

Turkey Point 4-For period 1/1/74 to 7/1/74-Total purges 5 Total time 984 hours0.0114 days <br />0.273 hours <br />0.00163 weeks <br />3.74412e-4 months <br />

• Maximum time (1 purge) 742 hours0.00859 days <br />0.206 hours <br />0.00123 weeks <br />2.82331e-4 months <br /> Minimum time (1 purge) 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Surry 1 and 2 Containnent operates at negative pressure. Discharge from vacuum pumps through filters to stack. During cold shutdown, there is continuous purging of containment.

Prairie Island L Frequency: Once per week for about 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Reason: To . relieve pressure buildup due to instrument air leakage to containment 2-53

TABLE 2-23 (continued)

PHR CONTAINMENT PURGING AND VENTING ~ EXPERIENCE (REF.17)

Kewaunee Frequency:- 5 times in 60 ' days usually for -less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, longer :

if for personnel entry. I Reason: Pressure control. During the 60-day period, purging l occurred for personnel entry.

Point Beach Continuous venting through a monitoring line at about 10 ft /3 min fl ow. Gas filtered on way to stack.

Palidades One- per week for about 10 minutes duration (planned upon resumption ofpoweroperation)

' Reason: To control pressure buildups Zion Venting for ressure buildup about twice per week depending on outside tt perature. Ranges from twice per day to once every two weeks.

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

Duration: 3/4 hour'on venting; 3-4 hours on purging.

Fort Calhoun**-

For periods from 1/1/76 to 6/31/76 and 5/5/77 to 12/31/77.

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

Millstone 2**

For period from 7/1/75 through 12/31/77.

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

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

t L ** Data for these plants was obtained from the Semi-annual Release Reports

! for the plants for the period indicated.

I l

2-54

For older plants (those under review for operating licenses or those for which the construction permit SER was issued prior to July 1,1975)

(Ref.18) not equipped with small diameter purge lines, frequent periodic purges or vents will be used to control the above parameters (Ref.18).

. 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.

2.2.10 CONTAINMENT INTERNAL CLEANUP SYSTEM 2.2.10.1 . Parameter Assume the internal cleanup system will operate for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> prior to purging, that-it provides a DF for radioiodine removal on charcoal

.adsorbers corresponding to the values _in Table 1-5, and a DF of 100 for particulate removal on HEPA filters and that there is a containment air mixing efficiency of 70%.

2.2.10.2 Bases Internal cleanup systems may be used to reduce airborne iodine concentrations in the containment air prior to purging. Such systems normally recirculate containment air through HEPA filters and charcoal

adsorbers to effect-iodine and particulare removal. For source term calculations, it is assumed that the cleanup systems are operated for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> prior to purging. . It is considered that charcoal adsorbers provide a DF for iodine corresponding to the values in Table 1-5, that HEPA filters provide a DF.of 100 for particulates, and that the containment air mixing efficiency is 70%. The system operation time of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> 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 pre-entry preparations.

A 70% mixing efficiency, based on data from the Ginna Station containment- building atmosphere . test conducted in 1971 (Ref.19), is used in evaluations. Data from Reference 19 are Parameter Symbol _

Value Length of test- run T 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

-8 Initial iodine activity A g 1.2 x 10 Ci/cc Final iodine activity A 1.2 x 10-9 pCi/cc 6 3

Containment volume V 10 ft Internal' recirculation F 6.1 x 10 5 ft3/hr i system flow rate

!~

2 - _ _

.The efficiency of iodine removal, E, can be estimated from

= exp (FET)

Substituting Ginna data.into the equation '

• 5 6 1 x 10_9 = exp [(6.1 x 10 )(E)(6)/(10 )3 10 = exp (3.7E), therefore E = 0.63.

The iodine removal efficiency E is a function of filter efficiency, E,,

and mixing efficiency, E,.

E = E,E,= 0.63 In calculating E, we used the assumed DF of 10 for charcoal derived from Table 1-5, (90% removal). Using E, equal to 0.9, E,is calculated to be 70%.

E,= E/Ea= 0.63/0.9 = 0.7 2.2.11 RADIOI0 DINE REMOVAL EFFICIENCIES FOR CHARC0AL ADSORBERS AND

-PARTICULATE REMOVAL EFFICIENCIES FOR HEPA FILTERS 2.2.11.1. Parameter Use a removal efficiency of 99% for particulate removal by HEPA filtration. For charcoal adsorbers, which satisfy the guideline of Reg. Guide 1.140 (Rev. 2), removal efficiencies for all forms of radiciodine are as follows:

Removal Efficiencies Activated Carbon Bed Depth a .For Radioiodine(%)

2 inches. Air filtration system designed 90

! to operate inside primary containment.

2 inches. Air filtration system designed 70 to operate outside the primary containment and relative humidity is controlled to 70%.

n a

Multiple beds, e.g., two 2-inch beds in series, should be treated as

! single bed of aggregate depth of 4 inches.

l 2-56

1 I

1 Removal Efficiencies l Activated Carbon Bed Deptha For Radiciodine(%) I t

0 4 inches. Air filtration system designed 90 to operate outside the primary containment and relative humidity is controlled to 70%.

6 inches. Air filtration system designed 99 to operate outside the primary .containnent

-and relative humidity is controlled to 70%.

2.2.11.2. Bases The removal efficiencies assigned to HEPA filters for particulate removal and charcoal adsorbers for radiciodine removal are based on the design, testing and maintenance-criteria recommended in Regulatory Guide 1.140, " Design,'. Testing and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants" (Ref. 2).

2.2.12 WASTE GAS SYSTEM INPUT FLOW TO PRESSURIZED STORAGE TANKS 2.2.12.1 Parameter The input flow rate to the pressurized storage tanks is variable depending on the system design as can be seen from Table 2-24 and 2-25.

'Therefore each applicant should supply the value of F, the waste gas system input flow to_ the' pressurized storage tanks. If detailed design information is not available, the data given in Tables 2-24 and 2-25 may be used. These data show that the average waste gas input flow is 170 ft 3/ day (STP) per reactor for PWR's without recombiners and 30 ft3 / day (STP) per reactor for PWR's with recombiners.

i- 2.2.12.2 Bases As 'can be seen from Tables 2-24 and 2-25 there is variation among

~

PWR system designs for the waste gas system input flow.

l

[ A review of the waste. gas processing systems proposed for a number of PWR's as given in the respective PSAR's and FSAR's has yielded the design flow rates ~ shown in Tables 2-24 and 2-25. -Table 2-24 indicates that for . reactors designed without recombiners to treat the gas prior to holdup in pressurized storage tanks, the average expected flow is approxi- .

mately 170 ft 3/ day (STP) per reactor. Table 2-25 indicates that for

? reactors designed with recombiners to remove hydrogen prior to holdup in pressurized storage tanks, the average expected flow is approximately 30 ft /3 day (STP) per reactor.

2-57

-- -._.-,...-,4

i TABLE 2 WASTE GAS SYSTEM INPUT FLOW TO PRESSURIZED STORAGE TANKS FOR PWR's WITHOUT RECOMBINERS Net Flow per Reactor R. : tor ft /3 day (STP)

San Onofre 2/3 57 Waterford 3- 1 71 Pilgrim 2 69 St. Lucie 1/2 139 Millstone 2 49

, Arkansas.1/2 68

~' Byron 1/2 173 Sequoyah 1/2 173 Marble Hill 1/2 173 Diablo Canyon 1/2 343 Trojan 225 Oconee 1/2/3 180 Davis-Besse 1 144-Bellefonte 1/2 163 1 .

~

. Average Net Flow for PWR's without- recombiners = 170 ft / day (STP) per' reactor-

+

k i

['

! ~

2-58 L

4 TABLE 2-25 WASTE GAS SYSTEM INPUT FLOW TO PRESSURIZED STORAGE TANKS l FOR PWR's WITH RECOMBINERS Net Flow'per Reactor-Reactor ft / day (STP)

-WPPSS 1 96 1

Farley 1/2 3 McGuire 1/2 18 Average Net Flow for PWR's with recombiners = 30 ft / day (STP) per reactor

• Net flow rate is determined downstream of any recombiner (which is assumed 100% effective in removing hydrogen).

6 2-59

{

(

2.2.13 HOLOUP TIMES FOR CHARC0AL DELAY SYSTEMS 2.2.13.1 Pa rameter

T = 0.011 MK/F l

l where. )

T is the holdup time, in days; and K is the dynamic adsorption coefficient, in cm /g, (see chart i

below);

3 M( is the mass of charcoal adsorber, in 10 lbs; t

3 F is the system flow rate, in ft / min; 0.011 is the factor to convert from (10 3 lb. cm3 /g)/(ft /3min) to days.

Dynamic adsorption coefficients, K, (in cm3 /g) are as follows:

Operating 77'F Operating 77*F Operating 77'F Operating 0*F

! Dew Point 45'F Dew Point O'F Dew Point -40'F Dew Point -20'F

! Kr 18.5 25 70 105 l

l Xe 330 440 1160 2410 2.2.13.2 Bases Charcoal delay systems are evaluated using the above equation and dynamic adsorption coefficients. T = MK/F is a standard equation for the calculation of delay times in charcoal adsorption systems (Ref. 20). The dynamic adsorption coefficients (K values) for Xe and Kr are dependent on operating temperature and moisture content (Ref. 21 and 22) in the i charcoal, as indicated by the values in the above parameter. The K  ;

values represent a composite of data from operating reactor charcoal ,

delay systems (Ref. 23 and 24) and reports concerning charcoal adsorption I systems (Ref. 20-22,24-27).

The factors influencing the selection of K values are:

1.- Operational data from KRB (Ref. 23) and from KWL (Ref. 24), and I from Vermont Yankee (Ref. 28).

2-60

2. The effect of temperature on the dynamic adsorption coefficients, indicated in Figure 2-3 (Ref. 21).

3. The'effect of moisture on the dynamic adsorption coefficients, shown in Figure 2-4. The affinity of charcoal for moisture, shown in Figure 2-5.

4. The variation in K values between researchers and between the types of charcoal used in these systems (Refs. 21 and 27).

Because of the variation in K values based on different types of charcoal and the data reported, average values taken from

.KRB and KWL data shown in Figure 2-3 are used.

-The coefficient 0.011 adjusts the units and was calculated as follows:

M(10 3 lbs) K(cm 3 g)(454 g/lb) 3.53 x 10-5 ft 3/cm3)

T(days)=

F(ft / min)(1440 min / day)  !

1 T = 0.011 f 2.2.14 LIQUID WASTE INPUTS 2.2.14.1 Parameter The flow rates listed in Table' 2-26 are used as inputs to .the liquid

.radwaste treatment system. Flows'that cannot be standardized are added to those listed in Table 2-26 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 applicant's proposed method of processing.

'2.2.14.2 Bases The flow rates used represent average values for a plant operating at steady-state conditions. The values are derived from values proposed by the ANS 55.6 Working. Group in proposed American National Standard,

" Liquid Radioactive Waste Processing System for Light Water Reactor Plants," (Ref. 29) from operating and design data, and from information furnished by applicants in response to source term questions. Data from Zion (Ref.' 5) indicate that the values for fraction of primary coolant activity given in Table 2-26 provide reasonable estimates of plant operating experience.

-2.2.15 DETERGENT WASTE 2.2.15.1 Parameters For plants with an onsite laundry, use the radionuclide distribution

_given in Table 2-27 for untreated detergent wastes. The quantities shown 2-61

a

+-^

10 5 Z

104 _-

_ 40 4

5 4 0

• o 103 _

i 2 8 -

z o _

t

= _

8 o

4 2 g 10 ._,

o Z e -

4 _

y _

y DATA FROM UNDERHILL 10 '. _- O DATA FROM BROWNING

+100 +32 0 -40 "F

200 l I I I I I O 0.001 0.002 0.003 0.004 0.005 0.006 0.007 RECIPROCAL TEMPERATURE (1/'K)

FIGURE 2-3 KRYPTON AND XENON K VALUES AS A FUNCTION OF RECIPROCAL TEMPERATURE 2-62 0

• - ~ , - , . ,

sg.

, .?

k ..

- L 60 -

r .

( l

\  ! I

'~ 5 L l

40 4' -

- uY 30 N N -  ;  ;

ir 1 20 COLUMBIA-G CHARCOAL 814 MESH KRYPTON 85 WITH OXYGEN SWE EP GAS (25'C) 10 0

0 1. 2 3 4 5 6 MOISTURE CONTENT OF CHARCOAL (wt %)

FIGURE 2-4 EFFECT OF MOISTURE CONTENT ON THE DYNAMIC ADSORPTION COE F FICIENT t

T i

N e

(

.k 2-63 1

l ur - - - - - -

l yt-l 1

m 45

- 40 -- ,,-

s F ADSORPTION 35 -

/

b-

,, G 30 --

E 7 WATER ADSORPTION

, . . - @f -

. DESORPTION / AND DESORPTION P " m . 25 / ISOTHERMS ON MACAR

-5 'f G210 AFTER REPEATED

,j 20 - ADSORPTION AND

' ' l_

-< DESORPTION g ,, _

_i (EOUILIBRIUM) n Lf

'[

10 - I DYNAhilC DETE RMINATION '

+ 1 ' 15-40*C R ANGE #'

S -

j) A MACAR G210

%,, 'O ~"~ ' I~ l' l ' 'I l I I s 0: '10 20 30 40 ' SO . 60 70 .- 80 ' 90 100 110 120 AIR RELATIVE HUMIDITY h .

FIGURE 2-5 CHARCOAL MOISTURE AS A FUNCTION OF

- RELATIVE HUMIDITY f ,

+'Y 4 5

g.-'

i 2

2-64

O $ T

.f

^ ~ *

.W"

' TABLE 2-26' -

PWR LIQUID WASTES.

~ EXPECTED DAILY AVERAGE INPUT FLOW RATE .(in Gal / day) _

- Type of treatment of blowdown recycled to secondary . '

. system (U-tube steam generator plants) or type of-treatment of condensate,(once-through steam .

generator plants). Plant with

. . blowdown treat-4 Deep-bed cond. ment. Product Deep-bed cond, demineralizers 'notirecycled to

. demineralizers without condenser or FRACTION OF t

with ultrasonic- ultrasonic. _ _ Filter- secondary coolant; PRIMARY COOLANT SOURCE resin cleaner resin cleaner demineralizer system ACTIVITY (PCA)' ,

i 1. REACTOR CONTAINMENT y-

$ a. Primary coolcnt pump'. seal . 20 20 20 20. 0.1

. leakage

~

$ b. Primary coolant leakage,- 10 ~ 10 10 10. 1.67*

l' miscellaneous sources

c. Primary coolant equipment 500 500 500 500. .0.001

, 2. PRIMARY COOLANT SYSTEMS (0UTSIDE OF CONTAllG4ENT) 4-

a. Primary coolant system 80 80 80- 80 .l.0 equipment drains

b. Spent fuel pit liner drains 700 700 700 . 700 0.001

c. Primary coolant sampling 200 200 200 200- ' 0.05 -

system drains d.' Auxiliary building floor ~200 200 200 200 0.1 drains i

, w - + - + - -- ___ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . - _ _._ - . _ . _

7 TABLE:2-26 (Continued)

3. SECONDARY'C00LANT' SYSTEMS

a. Secondary coolant sampling 1400 ~ .1400 - 1400 1400 10-4 system drains ~

b. Condensate demineralizer .3000 12000 .

- 10-8 rinse and transfer solutions '

c. Condensate demineralizer 850 3400- - - Calculated in regenerant solutions GALE Code

d. Ultrasonic resin cleaner 15000 - - -

10-6 solutions

e. Condensate filter- - -

8100 -

2 x 10-6 demineralizer backwash y f. Steam generator blowdown - - - Plant. dependent ii Plant dependent ii E g. Turbine building floor '7200 7200 7200 '7200 Calculated in-drains GALE Code 4 DETERGENT AND DECONTAMINATION SYSTEMS 300' 300. 300 **

a. On-site laundry facility 300

b. Hot showers Negligible- Negligible Negligible Negligible -

200 **

c. Hand wash sink drains 200 200 200 Equipment'and area 40 40 40 . **

d. 4 0..

decontamination TOTALS 29,700 26,300 19,000 10,000

• • GALE Code uses release data given in Table 2-27 to calculate releases from this source.

tt Input parameter

• About 40% of the. leakage flashes, resulting in PCA fraction of the leakage greater than 1.0.

TABLE 2-27

' CALCULATED. ANNUAL RELEASE OF.RADI0 ACTIVE MATERIALS IN UNTREATED DETERGENT WASTE Nuclide C1/yr/ reactor P-32 1.8(-4)

Cr-51 4.7(-3)

Mn-54 3.8(-3) 7.2(-3)

Fe-55 Fe-59 2.2(-3)

Co-58 7.9(-3)

Co-60 1.4(-2)

Ni-63 1.7(-3)

Sr-89 8.8(-5)

-Sr-90 1.3(-5)

Y-91 8.4(-5)

.Zr-95 1.1(-3)

Nb-95 1.9(-3)

Mo-99 6(-5)

Ru-103 2.9(-4)

Ru-106 8.9(-3)

Ag-110m 1.2(-3)

Sb-124 4.3(-4)

I-1 31 1.6(-3)

Cs-134 1.1(-2)

Cs-136 ~ 3.7(-4)

Cs-137 1.6(-2)

- Ba-140 9.1(-4)

Ce-141 2.3(-4)

Ce-144 3.9(-3)

TOTAL 0.09 Ci 2-67

s P

in Table 2-27 should be addedsto the adjusted liquid source term.

Detergent waste releases should be~ reduced. using appropriate decontami -

nation-factors from this report if treatment is provided.

2.2.15.2 ' Bases In the evaluation of liquid radwaste treatment systems, it is assumed that detergent. wastes (laundry and personnel drains) will have.the radio-

- nuclide distribution given in . Table 2-27. The radionuclide distribution

-is- based on measurements at four nuclear power plants, which are given

- in Table 2-28.

2.2.16 CHEMICAL WASTES FROM REGENERATION OF CONDENSATE DEMINERALIZERS 2.2.16.1 . Pa rameter

1. - . Liquid flows to demineralizer at main steam activity.

2.- All nuclides removed from the secondary coolant by the demineralizers are removed from~ the' resins during regeneration.

3. Use a regeneration cycle of 1.2 days times the number of demineralizers for deep bed condensate system without ultrasonic resin' cleaner (URC); for systems using URC, use a regeneration cycle lof 8 days times the number of demineralizers.

2.2.16.2 Bases Operating ' data (Ref. 30, 31) 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 from Reference 29.

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 allcmaterial 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.

- 2.2.17 TRITIUM RELEASES 2.2.17.1 Parameter

.The' tritium releases through the combined liquid and vapor pathways are 0.4 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 1.0 pCi/ml up to a maximum of 90% 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 2-68

y_; , m TABLE 2-28

'RADIONUCLIDE DISTRIBUTION OF-DETERGENT WASTE (millicuries / month)

Oyster Fort-

~

Creek Ginna -Zion

• Calhoun

, f (1971-1973) (1972-1973) (1977) (1977)

(Ref. 4)

Nuclide (Ref. 41) (Ref. 8) (Ref. 5)

P-32: 1.5(-2) NA NA NA C r-51 2.3(-1) NA- 9.4(-1) NA

-Mn'-54 1.3 1.2(-1) 1. 6(-1 ) '1.9(-2)

Fe-55 -3.5(-1) NA 1.9 1.6(-1)

Fe-59 2.9(-1) NA 2.6(-1) NA 2.4 00-58  ; 3.5 (-1 ) 4.1(-1) 1.5(-1)-~

Co-60 3.8 9(-1) 9.8(-1) 3(-2)

Ni-63 NA~ NA 3.5(-1) 7.1(-2)

Sr-89 2.1(-2) NA 7(-3) 1.4(-3) .

i~ _Sr-90 2.5( 3) NA 7.6(-4) NA Y-91 NA NA 1.4 -2). NA Zr-95 -8.3(-2) 1.6(-1) 1.4 -1 ) NA

. Nb-95 1.6(-1) 2(-1) 2.7 -1) NA Mo-.99 NA- 5(-3) NA NA-Ru-103: 1.3(-2) 3.2(-2) 5.2(-2) NA Ru-106 NA .7.4(-1) NA NA -

'Ag-110m. NA 1(-1) NA NA-Sb-124 6.1(-2) NA 4.7(-2) NA I-1 31 4.3(-1) 5.5(-2) 1.7(-1) 1.7(-2)

, Cs-134 1.7(-1)- 1.4 1.5 1.4 Cs-136 . _ NA NA - 6.2(-2) . NA-Cs-137 -2.9(-1) 2.5 2.1 1.7 NA Ba-140 7.6(-2) NA NA

-Ce-141 3.3(-2) 5(-3) NA NA Ce-144 7.3(-2) 5.8(-1) NA NA TOTAL: 7.7 7.2 11.4 3.5 Note: NA = radionuclides were not analyzed.

• For two units.

1

'2-69

ventilation exhaust systems. About eighty percent of the tritium in the gaseous effluents is released from the auxiliary building ventilation system, including the refueling area, and the remaining 20% of the tritium in gaseous effluents is released from the containment building ventilation system.

l 2.2.17.2 Bases 1

The release rate of 0.4 Ci/yr/MWt is based on a review of the tritium I release rates at a number of PWR's and on data from specific measurements of tritium inventory and tritium releases at the Ginna plant (Ref. 8).

The measurements at 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 1410 C1. For the same period, 910,000 mwd of thermal power were generated. Using these data, considering an 80% plant capacity factor and considering tritium decay, the annual average tritium release is (0.8)(365 days /yr) e-0.6930 )/12.3 = 0.43 Ci/yr per MWt 91 0 Wd Table 2-29 gives the reported liquid and gaseous tritium releases for 1972-1978 for thirty-five operating PWR's that use zircaloy clad fuel and started commercial operation before 1978. Table 2-29 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 2-30 illustrates the fact that the tritium release rate is approaching a steady state value of approximately 0.4 Ci/yr/MWt which is the value obtained from the Ginna measurements. At steady state, the release rate from a plant is approxi-mately equal to the amount entering the primary coolant since only about 5% per year of the plant tritium inventory will decay. Based on the data from Ginna and the data in Table 2-30 we will use a release rate of 0.4 Ci/yr/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 that is released in the nonrecyclable waste streams for the boron recovery, clean waste, and dirty waste systems. i The concentration of tritium in wastes originating from primary coolant is assumed to be 1 pCi/ml, consistent with the N237 source term. 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 prior to processing are used to calculate the tritium concentration in the waste streams.

Data in Table 2-31 indicate that tritium released in liquid effluents can make up a large fraction of the total tritium produced. Therefore we have considered that the tritium calculated to be released in liquid effluents is up to a maximum of 90% of the total quantity of tritium calculated to be available for release.

2-70

f' TABLE 2-29 TRITIUM RELEASE' DATA FROM OPERATING PWR's WITH ZIRCALOY-CLAD FUELS

• Nuclear Thermal Output per unit Power 6 10 MWDt per unit Startup Reactor Name MWt Date 1972- 1973 1974 1975 1976 1977 R. E. Ginna 1520 1969 0.32' O.45 0.28 0.40 0.29 0.46 H. B. Robinson- 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 0.87 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 1973/74/74 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 0.29 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 2560 1975 0.63- 0.59 Trojan 3411 1975 0.31 0.88 St. Lucie 1 2560 1976 0.73

. Beaver Valley 1 2652 1976 0.42 Salem 1 3338 1976 0.28 o Data from semiannual reports of reactors listed.

2-71

I L

TABLE 2-29 (continued) i TRITIUM RELEASE DATA FROM OPERATING PWR's WITH ZIRCALOY-CLAD FUELS

• Power Tritium Released (Ci/Yr) Per Site per unit Startup Gaseous Reactor Name- MWt Date 1972 1973 1974 1975 1976 1977

'R. E. Ginna 1520 1969 0.01 1.1 0.36 5.8 23.6 50 H..B._ Robinson 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 ** ** ** 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 1973/74/74 0.75 1600 502 62.6 Zion 1/2 3250 1973/73 ** ** **

iFort Calhoun 1420 1973 2.4 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 2560 1975 21.3 47 Trojan 3411 1975 1.5 2.9 St. Lucie 1 2560 1976 320 Beaver Valley 1 2652 1976 21 3 Salem 1 3338 1976 51

• Data from semiannual reports of reactors listed.

• • No reported data. ,

l i

2-72 1

TABLE 2-29 (continued) l TRITIUM RELEASE DATA FROM OPERATING PWR's WITH ZIRCALOY-CLAD FUELS

• Power Tritium Released (Ci/Yr) Per Site l per unit Startup Liquid 1 Reactor Nare MWt Date 1972 1973 1974 1975 1976 1977 R. E. Ginna 1520 1969 120 286 195 261 242 119 H. B. Robinson 2200 1970 410 431 475 624 980 685 Point Beach 1/2 1518 1970/72 560 556 832 886 694 1000 Palisades 2530 1971- 21 0 185 8.3 41.3 9.6 56 Maine Yankee 2440 1972 21 9 177 368 153 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 1973/74/74 124 3550 2192 1918 9 9 9 Zion 1/2 3250 1973/73 39.4 1.1 727 Fort Calhoun 1420 1973 111 122 157 Prairie Island 1/2 1650 1973/74 763 1925 1349 Kewaunee- 1650 1974 277 21 3 .295 Three Mile Island 1 2535 1974 463 189 ,192 ii Rancho Seco 2772 1974 132 0.0 0.09 ti' Arkansas 1 2568 1974 460 21 2 245 Calvert Cliffs 1/2 2700 1974/76 263 274 575 Cook 1 3250 1975 192 285 Millstone 2- 2560 1975 277 211 Trojan 3411 1975 36 311

-St. Lucie 1 2560 1976 242 Beaver Valley 1 2652 1976 108 Salem 1 3338 1976 296

• Data from semiannual reports of reactors listed.

i No radioactive liquid wastes were discharged from Unit 2 during the entire year. Note: For 1975, there were no radioactive liquid wastes discharged from Unit 1 during the last 6 months.

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

2-73

TABLE 2-29 (continued)

TRITIUM RELEASE DATA FROM OPERATING PWR's WITH ZIRCALOY-CLAD FUELS *

' Pmve r Total Tritium Released Per Unit per unit Startup -(Ci/yr.-Mwt at 80% capactiy)

Reactor Name MWt Date 1972 1973 1974 1975 1976 1977 R. E. Ginna 1520 1969 0.11 0.19 -0.20 0.19 0.27 0.11 H. B. Robinson 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 1973/74/74 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 0.01-0.13 0.19

~

Arkansas 1- 2568 1974 0.21 Calvert Cliffs 1/2 2700 1974/76 0.13 0.11 0.16 Cook 1 3250 1975 0.06 0.13 Millstone 2 2560 1975 0.14 0.13 Trojan 3411 1975 0.04 0.10 St. Lucie 1 2560 1976 0.22 Beaver Valley 1 - 2652 1976 0.22 Salem 1 3338 1976 0.36

• Data from semiannual reports of reactors listed.

I i

.2-74

?

TABLE 2-30

-TRITIUM RELEASE RATE FROM OPERATING PWR's AS A FUNCTION OF NUMBER OF YEARS OF OPERATION

~ (Ci/yr.-MWt per unit. at 80% capacity) 1 2 ~3 4 5 6 7 Ginna 0.11 0.19 0.20 0.19 0.27 0.11 0.17 Robinson 0.19 0.25 0.39 0.42 0.50 0.37 -

Pt. Beach.1/2 0.I9 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 Pt. 2/3 0.04 0.17 0.19 0.08 -

- Surry 1/2 0.11 0.11 0.32 0.30 -

' Turkey Pt. 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 Ft. Calhoun 0.12 0.12 0.12 0.13 -

Prairie Is. 1/2. 0.24 0.66 0.41 0.25 Kewaunee 0.20 0.14 0.19 0.20 -

TMI' 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 'O.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 l

2-75 L

TABLE 2-31

l. TRITIUM RELEASE DATA FROM OPERATING PWR's PERCENT OF TOTAL TRITIUM RELEASED IN LIQUID EFFLUENTS Reactor 1972 1973 1974 1975 1976 1977 R. E. Ginna 100.0 99.6 99.8 97.8 91 .1 70.4 H. B. Robinson 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 83.8 Palisades 97.7 99.8 ** ** ** 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 99.6 Oconee 1/3 99.4 68.9 81.4 96.8 Zion 1/2 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 0.0 ii 0.43 it 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 Average

• 99.2 98.0 91.1 89.5 83.5 78.5

• Average weighted by nuclear thermal output per unit.

• • No reported data.

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

2-76 i

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 l building ventilation exhaust systems. Based on measurements taken in L 1975 through 1977 at Ginna, Calvert Cliffs and Three Mile Island (Ref. 7) and in 1976 and 1977 at' Zion 1/2 (Ref. 5), and in 1977 at Turkey Point (Ref. 6), in 1978-79 at Rancho Seco (Ref .43), and in 1980-1981 at Prairie Island 1/2 (Ref. 42), Table 2-32 provides the distribution of tritium released from various sources within the plant. Based on data in Table 2-32, approximately 32% of tritium in gaseous effluents is released from the auxiliary building, 50% from the refueling area, and 18% from the containment. Since the refuling area in a PWR generally vents to the same release point as the auxiliary building, we have included these two releases together in our parameter.

2.2.18 DECONTAMINATION FACTORS FOR DEMINERALIZERS 2.2.18.1 Parameter Other Anion Cs, Rb Nuclides Mixed bed purification 100 2 50 system (lib 03 )

Boron recycle system 10 2 10 Evaporator condensate (H+0H~) 5 1 10 2 2 Radwaste (H+ OH~) 10 (10) 2(10) 10 (10) 2 2 Steam Generator Blowdown 10 (10) 10(10) 10 (10)

Cation bed (H+) (any system) 1(1) 10(10) 10(10) 2 Anion bed (OH~) (any system) 10 (10) 1(1) 1(1)

Powdex (any system) 10(10) 2(10) 10(10)

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

The following operating conditions were considered for the evaluation of demineralizer performance:

1. The DF is dependent upon the inlet radioactivity and 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.

2-77

TABLE 2-32*

. DIS'TRIBUTION'0F TRITIUM RELEASE IN GASEOUS EFFLUENTS Source of Gaseous Tritium Release (% of Total)

Auxiliary Refueling Containment

P1 ant -

Building Area Building Ginna (Ref. 7) 31 69 -NM~

Calvert Cliffs 1 (Ref. 7) 38 46' 16

-'Three Mile Is. 1 (Ref. 7) 5 43 52 Zion 1/2 (Ref. 5) 79 WA 21 Turkey Point 3/4 (Ref. 6) 75 17 8

- Rancho Seco (Ref. 43) 92 WA 8 Prairie Island 1/2 (Ref. 43) 7.2 91.8 1.0 Average 32 50 18 NM - Not measured.

WA - Release from refueling area combined with auxiliary building release.

-

• The following method is used to determine the 3 H release in this table.

Containment Building operation average % of total release E

(16 + 52 + 21 + 8+8 + 1)% + (6) = 17.7% = 18%

Then the Refueling Area for Ginna is reduced by 18%, i .e. , (69-18)% = 51%

Now the operation average % of the total release for the Refueling Area is.

(51 +'46 + 43 + 17 + 91.8)% + (5) = 50%

Then use (79-50)% = 29% and (92-50)% = 42% into Zion and Rancho Seco auxiliary building's data,'respectively, to calculate the operational

-average of Auxi_11ary Building release which is equal to (31 + 38 + 5 + 29 + 75 + 42 + - 7.2)% + (7) = 32%

2-78

2. When two demineralizers are used in series, the first demin-f eralizer will have a higher DF than the second. However, the F data in Reference 32 indicate that Cs and Rb will be more I strongly exchanged in the second demineralizer in series than I the first as the concentration of preferentially exchanged competing nuclides is reduced.

4 3. As indicated in Reference 32, compounds of Y, Mo, and Tc form k 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 y of these nuclides to the degree stated above. An analysis

=

of effluent release data indicates that these nuclides, although i present in the primary coolant, are not found in the effluent

( streams.

2.2.18.2 Bases E

h The decontamination factors (DF's) 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 In-Plant Source Term Measurement Program J

(Refs. 4, 5,,6, 42, and 43); (2) the findings of a generic review in the nuclear industry by the Oak Ridge National Laboratory (0RNL) (Ref. 32);

and (3) measurements taken at Three Mile Island 1 (Ref. 40). The DF's for the remaining demineralizers are based on ORNL findings.

The ORNL generic review contains operating and theoretical data 3 which provides a basis for the numerical values assigned. The ORNL data f were projected to obtain a performance value expected over an extended period of operation. It is considered that attempts to extend the service (g life of the resin will reduce the DF's below those expected under

% controlled operating conditions.

?

$ Average DF's for Ft. Calhoun, Zion, Turkey Point, Rancho Seco, and

[ Prairie Island stations were obtained by dividing the average inlet b radionuclide concentration of samples by that of the average outlet g concentrating for each nuclide.

Based on the data in References 4, 5, 6, 32, 42, and 43, the DF s

used for the parameter was that considered to be representative of the data.

w

1. 2.2.19 DECONTAMINATION FACTORS FOR EVAPORATORS 7

5 2.2.19.1 Pa rameter Decontamination Factors All Nuclides

Except Iodine Iodine 3 2 Miscellaneous radwaste evaporators 10 10 I Boric acid evaporators 10 3

10 2

2 2 Separate evaporator for detergent wastes 10 10

{

b i

[ 2-79 i

k

W ,

[ ,

~

f2.2.19.2 ~ Bas'es-

-~ - LThe' decontamination ~ factors for ~ evaporators are.-based oni (1) source

-term measurements made at Fort Calhoun, . Zion, Turkey Point,l Prairie. Island, 1 .

and' Rancho Seco stations by'In-Plant Source Term Measurement Program N ,(Ref.14, 5, 6,:42, .and 43) and (2) the. findings of a generic review in )

the: nuclear industry by the Oak. Ridge National- Laboratory (Ref. 33).

I . - Average' DF's' for -Zion, Ft. _Calhoun,l 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

~

s

'radionuclide.J Based on the data given in References 4, 5, 6, 33, 42, and 43, the DF eused for the parameter- was that considered to be the most representative o f : the . data'.

-2.2.20 . DECONTAMINATION FACTORS FOR LIQUID RADWASTE FILTERS

- 2.2.20.l' Parameter.

[ A DF: of:1 for liquidJradwaste' filters is assigned for all radionuclides.. .

! 2~.2.20.2 T Bases R'eference 34 contains fin $1ngs ;of:a generic review by ORNL of liquid radwaste filters used in'the nuclear industry. Due to.various filter.

types:and filter media employed,~ reported values of-decontamination factors-vary widely, with no' discernible trend. The principal _. conclusion reached in the' 0RNL.. report: is that no credit should be assigned to l liquid radwaste.

9 filters (DF of;1) 'until a . larger data base is obtained.

[ .

Additional _ data. from_ Ft. Calhoun (Ref. 4),. Zion 1/2 (Ref. 5) and

. Turkey-Point: 3/4:(Ref. 6),: Rancho Seco (Ref. 43), and Prairie _ Island 'l/2 i l(Ref. 42)~ indicate _that' decontamination factors in~ liquid radwaste.

' filters vary widely from less. than 1 to greater than 50 (with a mean v'alue '

~

Lof1.3). Therefore 'a DF of 1 for liquid radwaste filters-is-used.

2.2.21-l DECONTAMINATION. FACTORS'FOR REVERSE'0SMOSIS c ,

' 2.2.21 .1 Parameter i Overall DF. of 30 for laundry _ wastes and DF of 10 for_ other. liquid '

radwastes.

L2.2.21' 2 Bases

' 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 vol'ume processed through-the membrane to the inlet batch

~ volume'is the percent recovery. The DF normally specified for the

[.

9

[

~

2-80 7

., - . .w . . _._,__._.;..__,.. ,.a.

process is the ratio of nuclide concentrations in the concentrated liquor stream to the concentrations in the effluent stream. This ratio is termed as the membrane DF (DF ). For source term calculations, the system DF (DF ) should be use@. The system DF is the ratio of the nuclide conce5trations in the feed stream to those in the effluent stream. The relationship between the system DF and the membrane DF is

' nonlinear and is a function of the percent recove:y. Tnis relationship can be expressed as follows:

F DF =

s j _ () ,p)l/DF, where DFm is the membrane DF; DF s

is the~ system DF; and F is the ratio of effluent volume to inlet volume (fractional recovery) .

Tables 2-33 through 2-36 give membrane DF's derived from operating data at Point Beach, Ginna and Robinson (Ref. 35) and laboratory data on simulated radwaste liauids (Ref. 36). These data indicate that the overall membrane DF is approximately 100. The percent recovery for liquid radwaste processes .using reverse osmosis is expected to be approximately 95%, i.e., 5% concentrated liquor. Using these values in the above equation, the system DF is approximately 30.

DF = 0.95 8

1 - (1 - 0.95)l/100 = 30 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 2-33, 2-35 and 2-36, the membrane DF for these isotopes is lower than the average membrane DF used in the evaluation for laundry waste.

2.2.22 GUIDELINES 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% volume change in the tank, based on the liquid waste flow rates from the inlet sources.

2-81 l

- *- i' ,

~

Y 4

. TABLE ~2 "

< [ REVERSE'0SMOSIS DECONTAMINATION FACTORS,'GINNA STATION c:

f Concentrate;

, Activity 2 Product Activity.

3- 3- '

Nuclide. -(uC1/cm ) (uCi/cm ) Membrane DF Ce-144 ~ 2.68 (-4) <2.2(-7) -1200-

_Co;58 8.55 (-5) '<3.4 (-8) 2500 Ru-103 5.83-(-5)' <5.5;(-8) 1100' Cs-137- ,4.09(-4) 6.6 (-6) '60 Cs-134D 12.02(-4) 3.2-(-6) 60-JNb-95' 5.35(-5)' <5.3 (-8) 1000 oy Zr-95 2.36(-5) <3.7(-8)_ 640 Mn-54 .8.82.(-5) .<3.4 (-8) -2600

- '  ;Co-60' .9.62(-4) <8.1 (-8) 12,000 Total-isotopic 2.15(-3) 9.8 (-6). 21 9-Gross. beta 1.63_(-3) 1.86(-5) 88

' TOTAlf 3.78(-3) 2.84(-5)

Average 133.

4 e

.T

~ '

2-82 i '

TABLE 2-34 REVERSE OSMOSIS DECONTAMINATION FACTORS, POINT BEACH Concentrate Activity Product Activity Date Time (uCi/ml) (uCi/ml) Membrane DF 6/14/71 0840 1.1 (-5) 6.8 (-7) -16 1225 6.3(-5) 4.2(-7) 150 1350. 6.8 (-5) 3.2(-7) 21 3 6/15/71 .1030 2.7 (-4) 3.1'(-6) 87 1315 1.0-(-4)- 1.7 (-6) 59 1440 1.3-(-4) 1.1 (-7) 1200 1510 1.6 (-4) 1.1 (-7) 1500 l 1530 1.8 (-4) 5.7 (-7) 316 TOTAL 9.8 (-4) 7.0 (-6)

Average 140 w

t 2-83

- - - ..++r,-. ...p-q. ..i<ww. , ..wr -y- y , -e yr -,e'r.-

4 TABLE 2-35 REVERSE OSMOSIS DECONTAMINATION FACTORS, H. B.-ROBINSON NO. 2 STATION'

.l

.\

l Co-60 Co-58 -I-1 31 264 12 9 14 382 --- 20 l

436 --- 39 107 229- 26 76 490 96 i 94 1 31 11 Average 227 220 34 i

~

)

I i

2-84

1 c.

1 TABLE 2 -

EXPECTED REVERSE OSMOSIS DECONTAMINATION FACTORS

'FOR SPECIFIC NUCLIDES ,

. 1 Concentrate-

. , Activity:: Product' Activity

Nuclide (pCi/ml) - (p Ci/ml ) . Membrane DF-

- Co-60 '2.5 (-4). 5 (-7) 500 Mo-99' 3.8 (-2)- 1 - (-3)' 40

- l-131,'132, l'.2(-1) 4(-3) 30-1133, 134, 135!

Cs-134,137 4.3 (-2) 2 (-4)

-200-

. l TOTAL .2(-1) 5(-3)

. AVERAGE:- 40

.w. 4

- $ 9 . _.

k.

L 2-85

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.

13. 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 an, the nominal tank volume.

-The calculated values.in 1. and the~ total of 2. and 3. are used as inputs to the computer PWR-GALE Code.

. 2.2.23 ADJUSTMENT TO LIQUID RADWASTE SOURCE TERMS FOR' ANTICIPATED-OPERATIONAL OCCURRENCES 2.2.23.1 Parameter

1. Increase the calculated source term by 0.16 Ci/yr per reactor using the same isotopic distribution as for the calculated 1 source term to account for anticipated-operational occurrences  !

such as operator errors that result in unplanned releases.

+ 2., Assume evaporators to be unavailable for two consecutive days per. week for maintenance. If a 2-day hold-up 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:

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

b. Chemical Waste - Discharged to the environment to the extent holdup capacity or an ' alternative evaporator is available.

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

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

b. Effective System DF - Should the reserve storage capacity be inadequate for waste holdup over a two-day evaporator outage and should an alternate evaporator be 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.

[

2-86 r_.__,_.____._ ___. _ _ _ _ , _ _ _ . _ .

For example, a DF of 10 5t was calculated for a radwaste demineralizer and radwaste evaporator in series. If an adjustment were required for the evaporator being out-of-service two days / week, with only one day holdup tank capacity, then the effective system DF can be calculated as follows:

1. For 6 days (7 - 2 + 1) out of 7 the system DF would be 105 ,

2. For the remaining one day, the system DF would be 102(only the demineralizer DF is considered). The effective DF is:

DF = [(f)(10-5) + ( )(10-2)3-1 = 7.0 x 102 2.2.23.2 Bases ~

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 due to operator errors, 26 due to component failures, 5 due to inadequate procedures or failure to follow procedures, and the remaining 8 due to miscellaneous causes such as design errors. Table 2-37 summarizes the findings of this evaluation. Based on the data provided in Table 2-37 it is estimated that 0.16 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%. Unavailability is attributed to scaling, fouling of surfaces, instrumentation failures, corrosion, and occasional upsets resulting in high carryovers requiring system cleaning. A value of two consecutive days 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 / week outage, no adjustments are required for the source term. If less capacity is available, the difference between the waste expected during two days of normal operation and the available holdup capacity is assumed to follow an alternative route for processing. Since processing through an alternative route implies mixing of wastes having different purities and different dispositions after treatment, it is assumed that the fraction 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.

i 103 (Evap.) x 102 (demin) = 105 is obtained using DF's from Section 2.2.19.1.

2-87

I

-TABLE 2-37' FREQUENCY'AND EXTENT OF UNPLANNED LIQUID RADWASTE RELEASES FROM OPERATING PLANTS

• Unplanned Liquid Releases Total number (unplanned releases) 62 Fraction due to personnel error 0.37 Fraction due to component failure 0.42 e

Fraction due to inadequate procedures or failure 0.08 to follow procedures Fraction due to other causes 0.13 Approximate activity (C1) 24 Fraction of cumulative occurrence per reactor 0.16 year (plants reporting releases <5 gals of liquid waste / reactor year).

Fraction of cumulative occurrences per reactor 0.28

. year (plants reporting activity . released

>0.01 Ci/ reactor year)

Activity per release (Ci/ release) 0.39 Activity released per reactor year (C1/ reactor year) 0.16 Volume of release per reactor year (gal / reactor year) 633.

• Values in this table are based on reported values in 1970-1977 Licensee Event Reports representing 154 reactor years of operation.

l 2-88

-2.2.24 ATMOSPHERIC STEAM DUMP-J 2.2.24.1 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.

2.2.24.2 Bases.

In the evaluation, consideration has been given to the quantity of Lnoble gases and radiciodine 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% of the turbine building gaseous source term.

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

1. An average of two turbine trips annually;

2. -40% turbine bypass capacity to the main condenser;

3. Two-second rod insertion time required to scram the reactor following a turbine trip; and -

4. Twelve-second cycle time to recirculate one primary coolant volume through the reactor and steam generator.

The above parameters are based on a 3400-MWt RESAR-3 reactor. Using these parameters, it is postulated that steam will continue to be produced at a full-power rate during the time the contro1~ 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

= (1.5 x 10 7lb/hr)(60%)(14 sec)(2 trips / year)(454 g/lb)(30uu sec)

= 3 x 107g-steam /yr The iodine-131 concentration in the main steam for a U-tube steam generator _ is approximately 1.8 x 10-8 pCi/g-steam from Table 2-2.

Based on the steam release calculated above, the associated iodine-1311 release is approximately 6.0 x 10-7 Ci/yr.

7g t' )(1.8x10-8 pC1/g-steam)(10-6 1-131/yr = (3.2 x 10 r

= 5.8 x 10-7 Ci/yr 2-89

._________________.________.______j

l Releases due to low-power physics tasting 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% 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 prior to low-power '

physics testing. The iodine-131 releases are calculated to be approximately 4.6 x 10-6 C1/yr using the following equation:

I-131/yr = (1.5 x 107 lb/hr steam)(0.05)(454 g/lb)(10 hr/yr)

(1.8 x 10-8 Ci/g-steam) exp [-(0. 9 d s)] 10-0 Ci/pCi 0 ay I-131/yr = 4.6 x 10-6 Ci/yr 2.2.25 CARBON-14 RELEASES 2.2.25.1 Parameter The annual quantity of carbon-14 released from a pressurized water reactor is 7.3 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.

2.2.25.2 Bases The annual release of 7.5 Ci of carbon-14 is based on measurements at ten operating PWR's presented in Table 2-38. Kunz et al. (Ref. 37) found that the carbon-14 reacts to form volatile compounds (principally CH4 ,

C26 H , and C02 ) 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 from Refs. 4, 5, 6, 42, and 43 also indicate 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 1 atmospheres, l As shown in Table 2-39, an average of measurements, made at Turkey Point 3 and 4, Zion 1 and 2, Fort Calhoun, Prairie Island 1 and 2, and l Rancho Seco indicates that the release of carbon-14 breaks down to 22.6%

f rom the containment building, 61.0% from the auxiliary building vents and 16.4% from the waste gas processing system. Therefore on this basis, it is assumed that 1.6 Ci/yr of carbon-14 is released from the containment building, 4.5 Ci/yr of carbon-14 is released from the auxiliary building vents and 1.2 Ci/yr of carbon-14 is released from the waste gas processing system.

2-90

I TABLE 2-38 CARBON-14 RELEASE DATA FROM OPERATING PWR's Annual Average Plant

• 1975 1976 1977 1978 C1/yr-unit Conn. Yankee. 44 40 30 70 46 Yankee Rowe -1.6- 0.13 0.24 0.33 0.58 Annual Release

-Plant ** Area Ci/yr-unit Turkey Point 3/4 Aux Bldg. 2.4 Containment 0.075 i

WGPS 0.82 Spent Fuel Area 0.38 Total 3.7 Fort Calhoun Fuel Pool and Aux. Bldg. 0.30 WGPS 0.81 Containment Bldg. 0.78 Total 1.9 Zion 1/2, Cont. Bldg. 1.8 Fuel Handling and Aux. Bldg. 1.4 WGPS 0.062 Total 3.3 Prairie Island 1/2 Cont. Building 0.01 6 Fuel Handling and Aux. Bldg. 3.3 WGPS 0.25 Total 3.6 Rancho Seco Cont. Building 0.9 Fuel Handling and Aux. Bldg. 1.85 WGPS 0.85 Total 3.6 Average 7.3

• Based on semi-annual release reports.

i Based on In-Plant Source Term Measurements.

Waste gas processing system.

2-91

'f TABLE 2-39 DISTRIBUTION OF CARBON-14 RELEASED IN GASE0US EFFLUENTS-Plant Areas: Aux. Bldg. and Plant- Containment Fuel Handling WGPS 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%

j.

2-92

2.2.26 ARGON-41 RELEASES 2.2.26.1 Parameter The annual quantity of argon-41 released from a pressurized water reactor is 34 Ci/yr. The argon-41 -is released to the environment via the containment vent when the containment is vented or purged.

2.2.26.2 Bases 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 provides a summary of available data and gaseous argon-41 releases from operating PWR's. 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 34 curies per year.

2-93

TABLE 2-40

SUMMARY

OF ARGON-41 RELEASES FOR OPERATING PWR's FOR 1973-1978 (Ci/yr per reactor)

Reactor Name Year Release

Yankee Rowe 1974 '0.85 1975 0.93 1976 0.3 1977 0.49 1978 (1/2 yr) 0.47 Haddam Neck 1973 0.044 1977 0.08 1978 (1/2 yr) 0.041 Ginna 1975 5.8 1976 0.19

. Point Beach 1/2 -1973 17.6 1974 16 1975 208 1976 31 1977 9.2 1978 (1/2 yr) 13.3 H. B. Robinson 1975 (1/2 yr) 16.2 1976- 15.4 1977 '23.1 1978 (1/2 yr) -46.2 Surry 1974 (1/2 yr) 15 1975 0.32 1976 9.15 1977 (1/2 yr) 16.5 D. C. Cook 1978 (1/2 yr) 19.7 Turkey Pt. 3/4 1974 26 1975 51.3 1976 39.4 1977 45 Oconee 1/2/3 1974 (1/2 yr) 59.5 1975 42 1976 118 1977 8.1 1978 (1/2 yr) 19.9 2-94

TABLE 2-40 (continued)

SUMMARY

OF ARGON-41 RELEASES FOR OPERATING PWR's FOR 1973-1978 (Ci/yr per reactor)

Reactor Name Year Release Fort ' Calhoun 1975' 8.2 1976 2.2 1977 2.3 1978 (1/2 yr) 0.27 Palisades 1978 (1/2 yr) 0.01 Zion 1/2 1978 (1/2 yr) 24.8 Prairie Island 1/2 1975 1.3 1976 21 1977 31 . 8 1978 13.5 Kewaunee 1976 (1/2 yr) 30 1978 (1/2 yr) 5.9 Three Mile Island 1 1975 (1/2 yr) 50 1976 12 1977 66 1978 (1/2 yr) 46.5 Calvert Cliffs 1976 1/2 yr) 2 1977 1/2 yr) 3.1 Rancho Seco. 1977 9.8 1978 (1/2 yr) 1.8

• All data provided by the semiannual effluent release reports and the annual operating reports for each PWR listed.

2-95

fn CHAPTER 3. INPUT FORMAT, SAMPLE PROBLEM, AND FORTRAN LISTING 0F THE PWR-GALE CODE

3.1 INTRODUCTION

This chapter contains additional information for using the PWR-GALE Code. Chapter.1 of this report described the entries required to be

- entered on input data cards. Section 3.2 of this chapter contains sample input. data and an explanation of the input to orient the user in making the entries described in Chapter 1.

Section 3.3 of this chapter contains a listing of the input data for the sample problem and the resultant output. Section 3.4 contains a discussion of the nuclear data ' library used and a FORTRAN listing of the PWR-GALE Code.

3.2- INPUT DATA This section contains (a) an explanation of the input used in the sample problem and (b) input coding sheets for the sample problem.

3.2.1 EXPLANATION OF THE INPUTS FOR THE SAMPLE PROBLEM Only the inputs for the GALE code runs for the sample problera that are not obvious are explained:

Condensate demineralizer regeneration time (days)-

Input - 8.4 days Put this input in card 10 in the appropriate field allotted for this input.

Basis The sample problem assumes eight condensate deep beds, one of which is spare in parallel with no ultrasonic resin cleaning. The regeneration time for a bed is therefore 7 x 1.2 days = 8.4 days.

The liquid waste inputs are based on assuming the following:

A. Waste Generation Rates and Effective PCA Fractions Waste Type Gal / day PCA Fraction Shim Bleed 1440 Code applies the CVCS DFs internally Equipment Drains Pump seal leakage 300 i.0 Pump seal leakage (Table 2-26) 20 0.1 3-1

t A. Waste Generation Rates and Effective PCA Fractions (Cont'd)

Waste Type Gal / day PCA Fraction Other primary coolant leakage from miscellaneous' sources inside the containment (Table 2-26) 10 '1.67 Total equipment drain wastes '330 0.97 effective Clean Wastes Primary coolant equipment drains (Outside containment) 80 1.0 Spent fuel pit liner drains 700 0.001 Primary coolant sampling system drains (segregated from secondary coolant samples) 200 0.05 Total Clean Wastes 980 0.093 effective Dirty Wastes Primary coolant equipment Reactor containment cooling system 500 0.0 01 Auxiliary building floor drains 200 0.1 Secondary coolant sampling system drains 1400 0.0001 Total Dirty Wastes 2100 0.01 effective Regenerant Wastes 3400 Code internally calculates the buildup on the beds Condensate demineralizer rinse and transfer solution (secondary systemwastes) 12000 10"8 B. Available Equipment for Liquid Wastes Processing Capacity Equipment Number (Each)

Recycle holup tank (To collect shim bleed and equipment drains) 2 50,000 gal Clean waste holdup tank - 2 7,000 gal Dirty waste holdup tank 2 10,000 gal Regenerant solution receiving tank 2 20,000 gal 3-2

B. ' Available Equipment for Liquid Wastes Processing (Cont'd)

Capacity Equipment Number (Each)

Resin and transfer solution receiving tank.

.(To collect secondary ' system condensate

demineralizer -resin and transfer solution) 2 20,000 gal

. Clean waste monitor tank (For processed shim bleed, equipment drains and clean wastes) 1 10,000 gal Dirty waste monitor tank _

2 10,000 gal Secondary waste monitor tank -

.(For. processed regenerant wastes and secondary-system condensate demineralizer resinandtransfersolution) 2 10,000 gal Recycle-feed demineralizer-

'(To process shim bleed and equipment drains and located upstream of the recycle holdup 1 50 GPM tank )

Recycle ev'aporator condensate demineralizer 1 50 GPM Evaporator condensate demineralizer A 1 50 GPM (For clean wastes)

Evaporator condensate demineralizer B (For dirty wastes) 1 50 GPM i Secondary waste evaporator condensate demineralizer (To process regenerant wastes) 1 50 GPM Secondary waste demineralizer -

(To process secondary system condensate demineralizer resin and transfer solution) 1 50 GPM Steam generator blowdown demineralizer

_(To process steam generator blowdown) 2 in 300 GPM series Recycle' evaporator (For processing shim _ bleed and equipment drains) 1 30 GPM Radwaste evaporator (For processing dirty wastes and clean wastes) 1 30 GPM

' Secondary waste evaporator (For processing regenerate wastes) 1 30 GPM C. Additional Notes about Liquid Wastes

1. The above list' includes only the processing equipment assumed for generating the liquid waste inputs for running the GALE code. For example, it does not consider such equipment as filters, evaporator condensate tank, reactor makeup water storage tank, etc.

3-3

2. Except the condensate deep bed deminera'izers in the secondary system, all other demineralizers are assumed to be mixed bed and non-regenerative.

3. - The' processed steam generator blowdown is assumed to be totally returned to the secondary' system. It is also assumed that the steam generator, blowdown is 75,000 pounds /hr (~150 GMP).

4. Secondary system condensate demineralizer rinse and transfer solution waste has not been included as input for the sample problem GALE code run for the following reasons:

a. This waste is. assumed to be collected in a collection system dedicated for this waste in the sample problem.

b. Even if 100 percent of this waste is released without treatment, the release from this-stream is expected to be < 0.15 percent

-of the total liquid effluent release. If, however, this waste is processed by the secondary waste demineralizer listed above, the release from this stream is expected to be < 0.012 percent of the total liquid effluent release. Furthermore, it is likely that this waste will be processed and a major fraction of this processed waste will be recycled to the condensate storage system for eventual reuse in the secondary plant.

Note that if assumption a is not satisfied in any design, then the inputs for this waste should be properly integrated with the appropriate subsystem inputs (for example, the dirty waste subsystem) and the effective inputs for the combined waste system should be included for the GALE code run for that design.

5. The detergent wastes are assumed to be released without any prior treatment.

6. All the liquid waste subsystems included in the GALE code run for the sample problem have at least .a two-day holdup capacity for

. holding up the wastes prior to processing them.

7. In vieu of what has been stated above, no additional run need be made to evaluate the liquid effluent releases; also no adjustments need be made to waste subsystem DFs for possible equipment downtime.

D. The gaseous waste inputs to the GALE code run for the sample problem are based on assuming the following:

1. There is neither continuous degassification of the full letdown flow to the gaseous radwaste system via a gas stripper nor continuous purging of the volume control tank.

2. Fill time and holdup time for gases stripped from the primary system are based on the following:

3-4

\

~

E r

Number of pressurized storage tanks ~- 4

' Volume ;of each tank at STP - 650 CF

. Design pressure for each tank - .150 psig No recombiners

-t 3. Containment has small diameter (8 inches) purge line and the Llow volume containment purge rate .is .1000 CFM.

4. . : Containment has no internal cleanup (kidney) system.

5. . Number of high volume-containment purges during power operation -

0.'

-6. Fuel, auxiliary and containment buildings have HEPA filters and-four inch charcoal _ adsorbers on their exhaust lines and these

, filter units satisfy the guidelines of Regulatory Guide 1.140.~ -

Containment building has these ' filters both on the low and high.

, volume purge exhaust lines.- Waste gas system has HEPA filters on its exhaust line which satisfies the guidelines of Regulatory Guide 1.140.c The iodine releases .via the main condenser air '

ejector removal system are assumed to be released without any treatment prior to their. releases.

-7. Steam generator blowdown flash tank exhaust is'not vented -

directly to the atmosphere.

3.2.2 INPUT CODING SHEETS Figure 3-1 shows -the input coding sheets used for the sample problem.

3.3 ~ SAMPLE PROBLEM - INPUT AND OUTPUT Figure 3-2-shows printouts of the input and output: for a sample

' problem using the PWR-GALE Code.

3.4 LISTING 0F PWR-GALE CODE 3.4.1 NUCLEAR DATA-LIBRARY Calculation of the releases of radioactive materials.in liquid effluents.using;the GALE Code requires ~a library of nuclear-data available

.from'the Division of ADP Support, USNRC (301) 492-7713. For convenience, the tape consists of five files, written in card image form. The contents of.the'five files are:

File 1:

l. A FORTRAN listing of the liquid effluent code.

2. File 2: Nuclear data library for corrosion and activation r products for use with the liquid effluent code.

3-5

FORTRAN CODING FORtl CODER lDATE ADDRESS PHONE PROBLEM TITLE PWR-GALE CODE PROGRAM NO SHEET or lCHGNO.

S Taft wtC Os2 tac *

• C hi 2* Tec' hvwete i;ggyisic ar,Cm 3 . a o=a o I = A, p= a , e* At*aa I ei2i3ie r S 6 P i 9 i 9,.C I t'2 t 3 f'4, S r' 6 i.? itt U9:2Q2t n2 2i2 3.24JSi26n2 'i2 0:29130 313?i3 3 343St%3'.30,39:444 e42n43i444Sn444 *i4449 SCIS ' iS2iS 3,*4r* S 465'd 469.60.6 e62t6 3r646M6,6'16 6i4 9'^e i'2 Ni'4s'* t'6i"i'4N cia R.D, li i E A M F2 . e im i A 4F i eniri iRir Asr Terk Ri i ef s i i i i i , , , i i i i i i e i e i i i i i i i

,). , i i i i , iT ViPiEi e r , i , iP aw iR C A R D. 2. i P,0,WiTeH. . , iTdt E RiMiAiti i Pi ni wi ri R, i t i ri vi ri e r i fiM i r .r:i A iW Ai ti Ti ti ), i i , i i i i i i i a e e i i i i i i i i e i fa r , i i i i) cia Ride 3e iPittVintti e e iMiArtiti i Di Fi i Pi pt ie m Ai n vi i n n a ti a ti T if iTi Hidi ttis Ai h ni iL iB iS t), r i iiie i r , e i i i e i e e fi,,e i i )

C,A R D, 4, iL i E , T.Di Wr Ni e iP,ReliMrAipiYr i Si Yr Si T Er m iL iEiTiDi0eWiNi iniArTir, efiGiPiM,): i i e i i i i i e i i r i i e i e i e i > fi i i i i i iI Ci Ai Ri Di S i i t , B Fi t , R. . i eliE T D.0iu N. . C Ai ri fi ri n iD E MiliN.E R.A1 Li li 7, Fi n i re r_ iO Wi i fi C= P M li i i . . i , , , , i i , . , , i I f , , , i , i i_)

C,A R Di i 6. int 0iner,Ni e i iNiUiMiBiEtRt eniFi i tr Te ri Ai ne if:1ri u n a m 1 1 ne i e ie e i e i t t i e i a i e 't i , ,i e i e i ffie i i e e }

cia R,D, 7, Tio,S,T,Fifi e iT iO iT i A v f , r 9 s Tiri Ai wi i Fi r e ni 1.a if>Melil:L.ficeNe i Li Bi s /, M R li e i i , i i i e , i e i e i i i i i i , , 'fi i , i i , i)

CiAi R. D. 8, iW L,Ir i i , i Mi Ai % % ,0,F. ital,0,UnfiD, n ii Ni iriAira n i n Ti m a u iG E N EiRiAiTio,Ri i f. T Hi oi lli St A. M n sti R S 1 i (i , , , i e .)

Cr A Rini Qi iBitiWiDiWeNe r iBit:0iWrDi0iWiNi-,TiHiditn se itiBi/iHiRifi i e i , i 12 8: Lt On Wi D o W M iTiRiFiAiTiMiErNrTi if tN iP it!,Ti ini ,1 _i nt R12. , i ()

C,A,R D, i 1 On iR ,r ,t: ir im ,T i i iciorNiD EiNiSiAiT F1 iD,EiMil NiE R Aitifi7 Fi % in ir ie ir im i r i n i n ita ri ni o, iTifiMiFi if in iA iY iS ) , e i i i i i fu i i i r i e)

Y c)

C air,De 1 li F F c e Me e t !C,OiN,DiFiNis AiTert iDiE,M,IiNi Ee Ri At tr li Zi Ei n iF it ikliwi iF iR iAiC rif f iOiNr i e i i i t i e e i e i i e r i fie i t i i r)

C,A Ri Di 1 1 2, t , , , e i i r iSiHifiMi i R e t t Fi re ni i Ri Ai Ti re i e iie i r i r efi e i i , e ili iGiPiDi e i i i i , i i e i i i e i i i i i i i e i i i Ci iA rid, 1 3, t , r i e i , e iDiFifierf, i i , , r i),D F CiS,=ifi , , e i i)iDiFiDe i=,fi i i . , i) i i i i i i i , , i i , i , i i i i , , . , i i

.Cr A r R, Di 1 a, , e e , i , e r iC o d d iF ir iT iY iniN i if, i i i i )r iDiAiYisi iPipineriri t % i fi i e i il i iD iA iY iS i eFiRiAiriT. iDiliSiCiHi' if i i i il i i i cia,R Di 1 1 S. , , , e , i i i E 4 ,U il iP MiE iN iT i ,DiRi Ai li N: Si , Ii M P t1 Ti i , , ,f i i , e i i ils iG,Pini i f, i i e i) i prai,,iii.r i i ,i , ,

C,A R.D, l'6 , , , , , , i ,DiF ,lia ,(i i , , , , ,)iDiF .C Si=rf, , i i i ,)iDiF,0r i=i f e e i i , .) , , , , , i e i , i i i i i I

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1 CARD .1-NAME- NAME'0F REACTOR ~ SAMPLE PWR REV 1 TYPE = PWR

CARD '2 POWTH THERMAL' POWER LEVEL (MEGAWATTS) 3400. ,

CARD 13.PCVOL fMASS 0F PRIMARY COOLANT-(THOUSAND LBS) 550.

CARD f4 LETDWN ' PRIMARY. SYSTEM' LETDOWN RATE (GPM) 75. .

CARDE5:CBFLR LETDOWN CATION DEMINERALIZER FLOW (GPM) 7.5 CARD 6 NOGEN- NUMBER OF STEAM GENERATORS 4.

CARD 17 TOSTFL . TOTAL = STEAM FLOW (MILLION LBS/HR) 15.

~ CARD ;8 WLI MASS OF: LIQUID IN EACH STEAM GENERATOR (THOUSAND LBS) . 112.5 .

CARD '9 BLWDWN BLOWDOWN-THOUS LB/HR 75.0 BLOWDOWN TREATMENT-INPUT 0,1, OR 2 0

CARD 10. REGENT' CONDENSATE DEMINERALIZER REGENERATION TIME (DAYS) 8.4 CARD 11.FFCDM CONDENSATE DEMINERALIZER FLOW FRACTION 0.65 CARD 12- SHIM BLEED RATE 1440. GPD

-CARD 13 DFI= 5.0E03DFCS= -2.0E03DF0 = 1.0E05

' CARD'14 . COLLECTION -22.6 DAYS PROCESS 0.93 DAYS FRACT DISCH .1 CARD'15; -EQUIPMENT DRAINS INPUT 330.0 GPD AT 0.97 PCA

CARD 16 DFI= 5.0E03DFCS= 2.0E03DF0 = 1.0E05

-CARD 17' COLLECTION: 22.6 DAYS'-PROCESS 0.93 DAYS FRACT DISCH .1

. CARD 18 CLEAN WASTE INPUT 980. GPD AT .093 PCA CARD 19. DFI='. 5.0E020FCS= 1.0E03DF0 = 1.0E04 CARD 20 COLLECTION' 5.7 DAYS PROCESS 0.13 DAYS FRACT DISCH 0.1:

CARD 21 DIRTY WASTES 2100. GPD AT 0.01 PCA

. CARD 22 DFI=' 5.0E02DFCS= '1.0E03DF0'= 1.0E04

-CARD 23: COLLECTION -3.8 DAYS PROCESS 0.19 DAYS FRACT DISCH 1.0 CARD 24 BLOWDOWN- FRACTION PROCESSED l.

CARD 25- -DFI= 1.00E03DFCS= 1.00E02DF0 = 1.00E03 CARD 26 COLLECTION 0.0 DAYS PROCESS 0.0 DAYS FRACT DISCH 0.0

~ CARD 27 REGENERANT FLOW RATE'(GPD) 3400.

t CARD 28 -DFI= 5.0E020FCS= 1.0E03DF0 = 1.0E04 CARD 29  : COLLECTION 4.7 DAYS PROCESS 0.37 DAYS FRACT DISCH 0.1 CARD 30 IS THERE CONTINUOUS STRIPPING OF FULL LETDOWN' FLOW 7 0,1,0R'2 0 ,

-CARD 31 TAU 1 ' HOLDUP TIME FOR XENON (DAYS) 60.

CARD 32 TAU 2 HOLDUP TIME'FOR' KRYPTON (DAYS) 60.

CARD 33 TAU 3 FILL TIME OF DECAY TANKS FOR THE GAS STRIPPER (DAYS)' 30. .

CARD 34 GAS-WASTE SYSTEM HEPA?99.

( -CARD 35~ FUEL HANDLG BLDG CHARC0AL?90. HEPA?99.

CARD 36 AUXILIARY BLDG .CHARC0AL790. HEPA?99. . ..

l CARD.37 CONV0L1 CONTAINMENT VOLUME (MILLION FT3) 2.45- '

CARD 38 CNTMT ATM. CLEANUP CHARC0AL?0.0 HEPA?0.0 RATE (1000CFM)

i. -CARD 39- CNTMT-HIGH VOL PURGE-CHARC0AL790. HEPA?99. NUMBER / YEAR 0.0 L . CARD'40 LCNTMT' LOW.VOL PURGE CHARC0AL79C. HEPA?99. RATE (CFM) 1000.

l' -CARD 41 FVN ' FRACTION IODINE RELEASED FROM BLOWDOWN TANK VENT- 0.0 l CARD-42 FEJ PERCENT OF'ICDINE REMOVED FROM AIR EJECTOR RELEASE 0.0

' CARD 43 PFLAUN DETERGENT WASTE PF 1.  ;

l l

3-7a i

WAMPLE PwR rey I pua IMERMAL PosEy LEVEL (MEGAwATTsp 3400.0000 PLANT CAPACI!? FACTnR 0.R003 MAb5 OF PRIMARY COOLANT ITHOUSAND Leb: 550.0a00 FMIMARY SySI$N LSTDomN HATE ESPN) T4.0800 1 LLTDowN CATION 06MINERALITER FLOW lefM) 7.5e00 i NUMRER nF ST$AM DENEMaiORS A.0000 TOIAL STEAM (Lou (MILLION L95/HR3 1%.0000 Mass or L30010 IN EACM STEAN GENERATUR (THOUSAND LW53  !!?.5000 MA>5 0F WATEy th STFAN bENERATOR4 (IODUSAND Ls53 45s.0000 WLbwDOwN RATE tiUOUSAND LRS/HR3 T5.0300 PHIMARY 70 $$C0hDAHy LEAR RATE tL85/ YAY 3 T%.P900 Co8*0ENSAIF pgMIh$RALITER REGE NE R A TIDM TIME EDAYS5 R.4000 fib 5 ION PRODUCI CANRY.0VER FRACTION 0050 mal 0eEN CARRT 0VER FRACTION 0300 LUhDENSAIE USM!hbRALIZLR FLOW FRACT!WM 6500 LIguiu WASTE jNPUTS FSACTION FMACTION COL LECTION DECAT STREAM FLod RAT) 0F PCA DISC?ARSED TIME TIME DECONIAMINAIION FACTnRS (GAL /0AT4 IDAYS) (DAv51 3 Cs cTHERS

$ NIM WLEED RATE I.4AE.0J 1.0000 1000 33.6000 9300 5.0it 03 2.00E+03 1.00E.05 EDulPMENT nRAIN$ 3,3gE.04 9700 ,8000 34.4000 .?300 5.OnEe03 2.00F+03 3.00E*0%

CLLAN wASIE IMPt" 9. ROE.04 0930 1000 9.7000 3300 5.OnEe02 1.00E+03 1.00E+04 DIRTY WASTE 5 2.30F+0J .0100 1.0000 4.0000 5000 5.0gE+02 1.00Ee00 1.00E+04 Stowo0*N 2.16E+0% 0 000 0 000 0 000 1 0aE*03 1.0GF*03 3.00E.01 Id UNTRFAIED SLOwOOuN 0. 1 000 0.000 0 000 1.00E+00 1.00Ee00 1.00Ee00 33 RE6ENLRANT SnL5 3.40E 03 .300 4.T00 .370 5.00E+02 1.00E+00 1.00E+0A SA5E005 WASTE INPUT $

IMLRE IS NOT CONI! NOUS STRIPPING OF [ULL LETOWN FLOW MULOUP TIME t0R BEHON DAYS 3 60.0000 MULDUP TIME EnR RRYPTON (Days: 60.0000 FILL TIME OF DECAT TANKS FOR THE GAS SINIPPER DAv$3 30.0000 UAh WASTE SY9 FEM PARTICULATE YELEfsE FRACTION 0100 Aus]LgARy stue 10 PINE RELEASE FMACTION .IS00 PAMTICULATE WELEMSE FR.'.CTION 0300 CONTAINMENT WOLUBE (MILLION ET3D P.4500 FHLQUENCY OF Chini gLOU HIGN WOL PURWE (TIMES /TRS 7.0000 LNIMT.HIGH WPL PWRGEIODINE RELEAbt FWACTION .3000 PANTICULATE YELEast FhACTION 0100 LNIMT Low vv4 Pua6C HalEECFNp IO0s.0000 CNINT Low VUL PURGE 200!Nr RELEASE FWACTION .3000 PAATICULATE RELEASE FRACTION .0100 STEAM LFAK TU TuaBINL RLDo (LBS/Het 1795.0400 FNACTION IODINL SELEASED FROM BLUw00'N TANK WENT a.0000 PENCENT OF IUDIhE REMOVED FROM AIR E9ECfDN HFLEASE 1.0000 FIGURE 3-2 PRINT 0UT OF INPUT AND OUTPUT FOR THE SAMPLE PROBLEM

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=>

me gg > se as as N N sees M e $$ em e me e - N e4 se ame me e se se $$ Af e N e as se N e e e N se N Af me ao e M se e se e se N N

e. ueeeeeeeeeeeee++ JD e e e e e e e e e e e e e e e e e e e e e e o e e e e e o e e e e o e e e ==

ese an e 0 e eeeee 8 e se e o eee 0 eee 0 +0 e0 e 8 e e 6 eo e0 e ee 0 ee I e ee eo e I i

af M hpbWWb e > g r m e P e e.infePww#4# b g hf W 4ts W W bi las ind W W b W W 15>he 3eAmD.ecemeyit4ep 3 b W h.l 440 Sh# W has 4AJ he led b tad tad bl Ga.d 4.ad bs W W Ihlled b tad g W*= les be P OS e 9 t # $ e r p O=c e > e e e w ese em em Ee5Ne.8*eepmepMeDm eo e o e o e e e e gO g. ee eoeeOeNoe e em pe 3e e oe e e eo @ e e O ee. e m **e as e N o p e se e ee.e >= e M eme e peNo ee Ne M *M

• e*M

• um

.3 O e .e se e= N M P e P= me M N 9 N 2 m === e M e * .e oM N N M M M M N N 4 M e se e M P e M ** N == == ** ** == M ** *e A8 == W E

' t L 3 1-W me e sk me e e e e E

me e e

, e

== m e e e 5 ** m.e . e e e e e e .e mMM N m e e en e = = o e me m m e e C .J W

, - e i

J AsMWedWW.em.e**e .OM

N M e @ P O S S @ @ @ @ .e e e .e .eeeme ** ao sese N

Ass=N== M.. e n as.e .M 8 asM.eMseMseM- -7* sem see== e se e eenese e ==

4u e J e= st e

tap tus O O o s & E I m m u u =z= N2 m I . m ma u a. E S 2 > > > 3NI E u d

a z ID I D e 1, ma eee>>> iad.ae on > u uutusuuliafff=3tas es se == mo m se in 4 4 e as tas ime u

3-9 e + - * - . w - S - - - . - . -v w--------_.w. --er-,, . - , - m - - . , - - - - - , - , y--- w ww- -. m -,w +-wy- -., ----

T..1 TIUM RELECSE 200 CURIES PEH YEAR NOTE: .0000s If*DICATES THAT THE VALUE 15 LESS IHAN 1 0E-5.

Y

'bAltPLE FWH RLv i PWR Int.I'M4L PowEh LLbCL g rts.02b AT TF ) 3000*.00000 PLAtsi CEPACIly PCCT084 0,E0 MAb% of op tanPV L ouL A81T (TH005Af8" L64) 550.00000 MHINAsty SyslLH t.bfon* eta Hair. ii.PM) T5*.00c00 LLinuuN CATIuu ottittetHALIytp rlow gbrH3. T j 5 * *.10 horHto et sTpA> or.hEMAIons 4,.00000 f ul AL SIFA'S F LUa (MILLIOt8 LHS/MH) 15.00000 Aas% of LIDUa.0 AN E ALH STE AM 6LNh4A rou (THOUS AND LilS) 112.50000

.sLowbOW'8 f?Aik (TDOtisAtm LnS/HR) T5*.00000 Lu 40EfIS ATE 18bMI**tH ALIZE4 ptGLNtNaTI* J d TIME (DAYS) 0.40000 Lusa0E45* TE )tMidR ALl7ti' FLOW F l4*'C T i d'4 65000 LlutsIO wA$]F INPUI$

FH ACT Iuts FR AC T lute COLLECTIOf4 DF.C AY SIHLAN FLOW hAlp uF pCA oIgCet ApuEls T14E TIME DECONTAMINAtlpN FACTORS inAL/nAvl (DAYSt (DAVsl i CS OTHERS SWIM hLtt u palt 1.44r+0J 1.0000 1000 24.6000 .V300 5.0sE+03 2.00E+0S 1.00E+05 Eus#IPHL*'T npa1845 4.30F.od 1700 1000 22.6000 .V300 5.00E+03 2.00E*0) 1.00E+05 CLF At* WASIF !.ePUT o.ADF.

• 0 4 09JO 1000 h.T000 1300 5.0AE+02 1.00E+49 1.00Ee04-O!NTY 4A51F% 2.10E+0J .0100 1.0000 J.tt000 1900 5.00E+02 1.00E*00 1.00E*04 HLO dDuhte 2.16F+0h 0.000 0.000 0 000 1.00E+03 1.00E+04 1.00E+03 UNTRFAIED HLoadOH's c. 1 000 0.000 0 000 1 00E+00 1.00E*09 1.00E+00 NEGENLHANI SOL % :4. A GE + 04 . 300 4.T00 3T0  %.05E+02 1.00E*00 1.00E*04 casEoub WA%rF Inputs 1stLDE Is .#8T COMTI4ti"O*3 STHIPPIN's OF FULL LEinwN FLOW PLuw n4TE Ts:M0ubH esas $1pgeppa ('ipH) 1*.2291T Ld tit 8 Loup TIwE P,nd AE. ante (04ys 60.00000 la .toLoup TI=E P0H KHYPTO'4 (PA(Si 60.00000 t!LL TI'tE of DECAY TA4MS FUp TPE GAS STRIPPER (OAyS) 30,'00000 Pf41Nagy C.:Ot. TNT LtAK Tu AsseILIAMT llLd3 (LH/UAy) 00000 bas WASTF SVbTEM PARTICOLAIE M(LtaSE FRACTION 360,01000 tesLL HAraDLG ULOG 10DI'8E pt LL A3E FN ACTION hl0000 P Adi t ctiL A lt HELLASL FRACTION 01000 auxILIAny HLUG 190Iht .aLLLASE FWACTION '.10000 P Ad T I cut. Ait hELLMSE FRACTIuN 01000 CONTA1HHEN{ WOLut!E (MILLId4iT33 2'.A5000 FHLOUFNCY OF PHINAdy ConLgny DEuASSiqG (TIMES /YR) 2 00500 phi"ARY To ShC0hpAlty LEAK HATE ELR/00VI T5)00000

IHLHE 19 HUT 4 MIDI 4FY FILTER FHaCTidd topf t#L hfPASSING CohnENSA1E DEMINEHALIZER 35000 luulhE PANTil10m FACTOU t ri AS/L I WG I D) IN STEAN 6ENERATOR *01000 FPLuyFNCY OF ChlMT 9LDd litGH VOL PUHuEtTIMES/VH3 2.00000 Clel.ti Hisin VUL Pu. stir t udlNT PLLEAbt FM4CTION *.10000 PAHIICesLATL MFLEcSL FHACTION '.01000 L'Jll'T Low vcL POWat etAtt f Ct e ) a 1000.00000 CHitif LOP VUL PUH'il 100 t'8E Ht.t.E ASF FHACTION *10000 PA'iflCUL A TE i4 ELL ASE FHACTION 01000 STLAM II A, t u Iudnit.t. blaso gLIN/*t) IF00,.00000 t it a C 11 +)48 101.)HE HELF 8 ht e Fhu*e tiLowDU*N TANK Vf t4T 0,00000

PLNCLHI tv IdolhL ltFHOWt p FH JH All LJECION HELtASF  !.00000 l

l l

l l

l 0 8 9  % . __ _ _ _ _ - _ _ _ _ _ _ _ _ . _ _ _ . . _ .

.SCMPLF Pop I?Fv 1 SASEGUS DELEASE RATE . CURIES PER YEAR Pn!MAHY SLCONUAHf Hu!LDING VENTILAft0M . .

CootaNI Co0LANI .. ... ---.----- .. ........- ..-...==..=.6 DLOWOOWN AIR EJECTOR TOTAL tu!ChutI/GHit"lCROSI/6") FHLL ManuLG REACToa auxgt3ARy 1peetaE , VFNT OFFSAS EMMAUST I&til 4.SUOL-02 1.J97E.U6 5. 4. 04 3.7E-03 i.4E 02 9. 8 4. }.pfEOS IL11J 1.400L-01 1.fi7E-Db 1.*E 03 7.9F-03 4.5E.02 1 9E=04 8 8. 5.55 03 TuyAL H.s pELEAStp Via 045EOUS PATHWAY a 1100 CI/YR C.l* HELtASED vlA 9AS9005 PATHWAY a 7.3 C1/yR AM.43 RELEASED v!A CONTAINMENT

• VENT e 34 C3/yp LiJ B

a N

6 0 0 6 0 se M se ee M es M e e e M 8 0

t .eo .e*

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t

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ta=

9 me e oe ee e ee eo e e e o On e == 8 9 e e o e 4 0 0 W et se e ee ** * > d e e e 0

> 0 t 4 0 0 0 4 8 0 8

8 e e o e e e e e e

!4 U 0 e * * * *

• 4 e o E Ina > 0 taf W One taa Int taf W Inp ana $

SW 8 ee e ee e e 4 e e e 6 and 3 0 e e o e e o e e e 0 4 0 Elf 4 ** M k e N ** O e wo 1 E3 8 9 ee at 4 0 4 tad 0 0 0 0 E O 8 4 0 '0 Inf W 4 0

> 4 0 t se G 0 0

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• e e O

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e* taf 9 0  %

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> 0S 8 e e e is e e e e e e e 8 e D

SE 8 9 e d(' D 0 s e= e t .

Inf 2 5 0 t m W O 4 0 0 4 4i se 0 9 e e e o se e es e en e 0 g Ini > 0> 0 ee e e o e e e e e e 0

.J 4 9E O * *

• e o e e o -e e tal hJ .J 84 8 inJ InJ tal tal taJ tal tal in, la? Inf I J S == 0 ** I e o e e an e in e e e 8 e e- t .3 0 e e e e e e e e e e e e o M 2

• 9 e= *B -F e l'. C t == le= @ M t oe oc M S 2 3 W G4 0 4 O > 83 0 0 E taf 84 0 t 3 04 9 5 0 0 in.

4 2 0 0 0 W ** 9 0 ** N == ee Pt N PS e N O e E Q 0 0 e e e o e o e e e o 0 >

.A 5E O e e e e e * * * *

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e= 80 0 W Ina W tas la. LaJ tat La; saJ enJ t e*

3 8 t=

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9 e

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ift e

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o e ee e e e 8 Q e e t 84 4 4 9 == @ =* **

• f'l e e N 8 O t ena 0 0 e E I O ==

0 e t 4 2 8 0 4 0 0 N == 0 Z 4e e o o e e-t3 0 e e 8 0C 0 La; 6 8

• t 3 A e' t023 8 9 e e e -a e o e o e o e O

0 ast E S se a=

0 O e O O N o O C o e O O tai 8 t= 0 0 .A 1 0 3 6 9 "A 00 0 4 A

== 0U e I o=e er 8 0 4 I t= 0 e fb em 0 tn3 Et* 0 9 o o e M I- 07 0 e e 4 4 i of 9? I 4 laJ 0 W l ( 0C 0 == == 8 .A l C ea 8 e e e e e e e e e o e t 4

> 0 *= 0 C' N C, O ** tt O O O o O 8 T .

8 0 8 -

0 .2 8 9 A tr 5 0 W D 0 p=

== 0 2 C e s > C > 3 > & C 0 4

& 4 4 3 3 0 3 e 3 3 O 2 S I J e=a > -l* 0 g g g 3 g g g g g g g g -e T *= % 0 Las .A 4 ina taJ ud 3 af d Z ee 3 T *J 8

3 o o 3

a o 3' nas a 3 c

a o

La#

o taJ o

eel o

naJ c a 0

0 2 l 0 3 3 3 3 S 0 ==

i 1 2 .A '3 0 4 3* O @ @ @ 4 h 2 == .r. 9 03E 6 e e e e e e o e e o e 4 tas t  ? El C U 0 I" E f* t == == tf ft e A M I J

! #1 laJ W == 0 0 T a er 3 e e c 0 0 >

r La,? == 8

==,

.= == et e s A 0

.J E I C o o o o o o o o o o 0 .J j 1 C 9 0 8 8 0 0 t e t t t 8 m 8 I i L >*=% 0 43 as of .J J aa .8 .nJ 4 4 .a .J 0 ==

• ( T 2 == 9 O 3 o 3 a 3 o G S c C /* I

' 'A 4 t *J 0 C D D t% C C C D S *= C < 0  ?

T .0 3 6 4* M an G 881 o C ** *D t *st T t ==

== I e o e o e o e e o e o tr CCt. 3, e I I == 4 == N > > n e it ta w k 8 e I llL U =* 8 .J 0 .'.

3 0 J t e=e

~ 8 C 8 2 8 T f T 2 0 4 8 2 == m M A A A m 6 .nJ 9 A A e= S *l 'l 8'l =4 at a4 t's .# 9 1 0 6 K t E e e e. e e e e en 0 *L 4 5 0 0 0 8 0 0 0 9 0 0 t- 4 4 0 E Z E 2 .aA nad .A mJ aJ .aJ W O I I E M an. E x x x x x x w *= 8 C 8 0 e 0 0 0 3-13

aAtiPLI add R$v 1 AIHbOHNE PARTICULATE RELEASE RATE.EURI:s PER YEAR ansTE GA5 BUILDING VENTILATION Nutt. lWE SfsTEM REACTOR AUAILIARY FUEL HANGL6 total CP651 1.4F-0F 9.2E-05 3.2E.06 1.6E-06 l

9.7E 05 MdL%4 2.lE.08

! 5.3E.05 7.9E.67 3.0E.06 5.7E-05 l COL 57 C. 8.2E.06 0. O. 9.2E.06 CO.%9 B.fr-08 2.bE.04 1.9E.65 2.lE.04 4.8E-04 C0464 1.4E.07 2.6E.05 5.!E 4e .8.2E.05 1.lE-04 FEL%o i

1.8E-08 2.iE-05 5.0E-57 0 2,8E.05 SHL39 4.4E-pi 1.1E-04 7.5E-56 2 !E.05 1 6E 04 SR.94 1.7E 0! 5.2E-05 2.9E.56 8.UE.06 6.3E.05 2R69%

• .dr-08 p. 1.0E.55 3.6E-08 1.0E-05 NHk9%

w 3.7F-08 1.8E.05 3.0E.57 2.4E-05 4.7E.05

2. Hubin3 3.2E 08

.p. 1.6E-05 2.JE.57 3.8E.07 1,7E.05 Hubin6 2.7r.08 0. 6.0E-59 6.9E.07 7.4E-07 SH.125 O. c. 3.9E-58 5.TE-07 6.lE-07 CS-114 3.3F.01 2.5E-ss 5.4E-56 1.7E.05 4.SE-05 Csello i.3r.0d J.2E.05 4.8E-57 0. 3.3E.05 CSel17 7.7 E 07 5.5F-05 7.20-66 2. TE 05 9.0E.05 BAbl40 2.30-Oi 0. A.0E-56 0. 4.2E.06 CE.141 2.2r.0$ 1.3E.05 2.AE-67 4.4E-09 1.3E-05

......... .....--......... 4......

i l

-3. File 3: Nuclear data library for fuel materials and their transmutation products for use with the liquid effluent code.

4. File 4: Nuclear data library for fission products for use with'the liquidreffluent code.

5. File 5: A FORTRAN listing of the gaseous effluent code.

The tape is written in the following format:

DCB = (RECFM = FB, LRECL = 80, BLKSIZE = 3200)

Use of the tape requires two data cards in addition to those described in Chapter 1 containing the plant parameters. For a low enrichment uranium-235 oxide-fueled light water reactor, these cards should always contain the following data:

Card Column Input Data 1 1-72 Title 1 75 The value 2 2 1-10 The value 0.632 2 11-20 The value 0.333 2 21-30 The value 2.0 2 31 -4 0 The value 1.0E-25 2 41-46 The date (month, day, year) of the calculation 2 48 The value 1

2. 50 The valu 0 2 52 The value 0 A description of the taformation contained in the nuclear data library can be found in the report OKNL-4628, "0RIGEN - The ORNL Isotope Generation and Depletion Code," dated May 1973.

3.4.2 FORTRAN PROGRAM LISTING Figure 3-3 and 3-4 provides the program listings for the PWR-GALE Code gaseous and liquid determinations.

3-15

FIGURE 3-3

~

• DECK PuALEGS PROGRAM LISTING FOR GASE0US DETERMINATION 000260 C . GALE CODE F OR CALCUL3Tlhb GASEOUS EFFLUENTS (ROM PwHS. MODIFIE1 C' AUG. 1979 70 IMPLEMENI APPENDIX 1 TO 10 CPR 50 HEACTON C WATER CONCLNTRATIONS bALCULATED USING METt10DS OF DHAFT STANDARD On0300 C' ANS,237 "HADI0 ACTIVE MAltkIAlb IN PRINCIPAL FLUID STHEAMS OF 000310' C LIGHT WATER COOLED NukLEAN POWER PLANTS". DRACT DATED MAY'20, 1974000320 C

C -THE FOLLOWING FIRST STATEMENT IS SPECIFIC FOR.TNE CDC USEks.

~

C FOR THE IbH USERS, DELETE THIS STATEMENT. l t

C PROGRAM PGALEGS(INPUTS 0VTPUT, TAPE 5= INPUT,TAPEb=00TPUT)

HEAL NUCLIU(13)

HEAL PPART(18)

DIMENSI ON ACONT(13),CUCP(13),CBSP(13),ASHIMC(13),ASHIMS(13)

DIMENSION CHPP(13)

DIMENSION A5HIH(13),CMNCP(13),CONCS(13),DECON(13),FHHL(13)

DIMENSION DECOH(13)eNAME(W),EJT(13)iT B L(13),C9L(13),AX8L(13)

OIHENSION bV0G(13), TOT (13),X2(13),X3(13),X4(13),x5(13)

DIMENSION CTPR0(13) XCl(13),xP2(11),W RD(14),WAHD(5),WUND(4)

O D1HENSION PCBL(18),PAABL(18),PC8P(18),PAXpP(10),PGWS(la),PTOTP(18)

DIMENSION POWL(18),PCYCP(18),PCBSPg18),PA90NT(18)

' OIMENSION PFH8L (18) ,PtH8P (18) 01MENSION hh AX (2) ,RN AXS (2) ,RNF H t 2) ,RNFHS (2) , RNT (2) ,RNTS (2) ,RNS (2)

DATA htlCL10/" KR-85M"s" KR-85"," KR-87",",KR-?8"," XE-131M",

1" XE-133M"," XE-133"," XE*135d","

xE-135" " XE-137",

2" xE-138"," I-131"," l-133"/

~ DATA PpART/" CR-51"," MN-54"," CO-57"," CO 58"," CO-60"," FE-59".

1" SR-69"," dR-90"," ZH 95"," HB-95"," RU-103"," RU-106"," SB-125",

p" CS-134"," CS-136"," CS-137"," RA-140"," CE-141"/

C C XP1 AND Av2 AHE THE PHIMANY COOLANT AND SLCONVARY COOLANT

-C CONCEhTRATIONS, RESPLETIVELY (MICROCI/GM).

C DATA Xp)/1 6E-1,4.3E-1,1.SE-is2.8E 1,7.3E-1,T.0E-2,2.6E*0',1 3E-1 8 1.bE-1,3.4E-2,1.2E-1,4,5E-2,1.4E-1/

DATA AP2/3.4E-8,8.9E-9,3.0E-8,5.iE 8,1.5E-7,1 5E 8,5.4E-7,2.7E.8.1 1 6E-7,7.1E-9,2.5E-8/

C C DECAY CONSTANTS FOR TUE CORRESPONDING NUCLID (1/SEC). ,

C DATA DECON/4.38E-5,2.03E-Y,1.52E 4,6.8 E 5,6.#0E 7,3.55E 6,1.52E 6 8 )

1,7.41E 4,2.09E-5 2.96E 3,8.14E-4,9.97E-7,9.17E-6/

i C .

C NORMALIZED I^91NE ANHUAL RELEASE (CI/. Y R/MICRDSI/GH).

C DATA RNS/0.32,0.32/

_ DATA RNAX/0 68,0 68/

DATA RNAX5/2.5,2.5/

DATA RNFH/0.038,0.038/. 1 UATA RNFHS/0.093 0.094/

DATA RNT/3.8E3,3.8E3/

OATA RNTS/4.2E2,4.2E2/  !

i: C C PARTICULATE ANNUAL RELEASb RATE (CI/YH) i: C DATA PCRP/9.2E 3.5.3E'3,8.2E 4,2.5E-2,2.6t 3,c.7E-3,1.3E 2,5.2E-1, 10.0E 0,1.aE-3,1 6E-3,0.0E+0,0 0E+0.2.5E 3,3.2E 3,5.5E-3,0 0E+0.1.3 L 2L-3/

DATA PAXdP/3,2E 4,7,89 5,0.0E+0,1.9L 3,5.1E 4,5.0E-5,7.5E-4,2.9E 4 1,1.0E-1,3.0E-5,2.3E-5,6 0E-6,3.9E-6,5.4E-4,4.?E-5,7.2E-4,4.0E-4,2 pot-S/

DATA.PFHdP/1.8E 4,3.0E 4,0.0E+0,2.1E 2,8.eE 3,0,0E+0,2.1E-1,8.0E 4

- 3-16 I

1,3.6E-6,2.,*L-3,3.8F. 5,6.9L-5,5.7E-5,1 7E-J,0.UE.0,2.7E 3,0,0E.0,4 24E T/-

DATA PGWS/1.4E-5 2.1E'6,0 0E.0,8.7E-6,1.4L 5,1.8E-6,4.4E S.I.7E-5, 14.8E 6,3.7E-6,3.2E-6,d.7L-6,0 0E.0,3.3E 5,5.3E-6,7.7E-5,2 3E-5,2.2

- 2E-6/.

C C BUILT-IN PAHAMETERS C

OPFRAs6.80 j AUXLRs160.

ems 2.0-GENLs75..

CLFNGun.03 CLFIa6.0E-6 PURTIMs16.

TBLn=1700.

C AFPTEG=0.0 3 0618R0 READ (5,1000)NAME. TYPE WRITE (6,1440)

. WRITE (6,1000)NAME, TYPE HEAD (5,1010) WORD,POWTM WRITE (6,1010) WORD,P0d!H WHITE (6 1020)

READ (5,1010) WORD.PRIVUL WHITE (6,1010) WORD PRIVOL HEAD (5,1010) word DEM1FL WRITE (6,1010) word DEH1FL READ (5.101u)wCRD.CHFLR WHITE (6,1010lWOHD,C8FLR HEAD (5,1010) WORD, GEN WHITE (6,1010) WORD, GEN READ (5,1010)WURD,TOSTtL WHITE (6,1010) WORD,TOSTFL HEAD (5,1010) WORD,WLI WRITE (A,1010) WORD,wl!

wLI=GENedLI 054600

! HEAD (5,1040)TBD,KFNRT WHITE (6,1050)TRD IF(KFhRT.EG.0)FNRTSC=0.99 IF(KFNRT.EQ.1) FNRTSCfD.9

• IF (KFNRT.Eu.2) FNRTSS=1 0 READ (5,1010) WORD.REGE!1T

WHITE (6,1010) WORD, REG @NT .

i C 004740 l C HEAD DATA FOR LIQUID .1,NF0HMATION

! C 004760

( HEAD (5,1010) WORD FFCDM r

WHITE (6,1010) WORD,FFCDM HEAD (5,1060) WARD,58 LOR cwa =1.6 004820

HEAD (5,1070)DFICW,DFC)Cw,DFCW l NEAD(5,1080)TC,TSTORC CWFD WRITE (6,1090)

WRITE (6,1100)

. WRITE (6,1110) WARD,SBLUR, cwa,CWFD,TC,TSTORC DF1.Cw,DFCSCW,DFCW ftEAD(5,1120) WARD.EDFLR, EDA

HEAD (5.1010)DFIED,DFCbED DFED

. HE AD (5,1080) TE.T4,EDF u WHITE (6,111Q) WARD,EDFLR. EDA,EDFD,TE,TS.DFIED,uFCSED,0 FED HEAD (5,1120) WARD,0WFLM,0We HEAD (5,1070)DFID*,DFChDe,UFDW 3-17 l-

HEAD (5,108Q)TD,TSTORDoDeFW WHITE (6,1110) WARD,DWFLR,0wA,0wFD,TD,TSTORp,DFA.Dw,DFCSDw,DFDw READ (5,1120) WARD,0wFL2,04E HEAD (5,1070)DFIDP,DFC4D2 pF02 READ (5,)o80)T2,TSTOR2sDeF2 WRITE (6,1110) WARD,0WFha,DW2,0WF2,T 2 ,TSTORd,DF102,DFCSD2,DFD2'

-RLAD(5,1130)RDTFR READ (5,1070)DFICH,DFC)CM,DFCM READ (5,1080)TCH,TSTORg,CMfD READ (5,1130)ROWFR READ (3,1070)DFIRG,DFCyRG,BFRG READ (5,1080)TRG,TSTORN,RGFD IF(TBD'EQ.0.0) GD TO $0 BDFR=TBD*1 0E3*BDTFR/0.34T6 055070 WHITE (6,1140)BDFR,CHFU,TCo,TSTORR,DFICH,DFCSCp,DFCM BufR=TRD*1.UE3* ABS (1 6BDTLR)/0.3476 005090-WRITE (6,1150)SDFR IF (FFCDM.EQ.0 03 GO Tu 50 ,

30 IF(REGENT.Eu.0.0) GO TO 40 WRITE (6,1160)RGWFR.RGtD6THG.TSTORR,DFIRG,pFCS7G,DFRG GO.To go +

40 RGwFR=n.0 WRITE (6 1160)RGWFR RGtD,THG.TSTORR.DFIRG,UFCSOG.DFRG C .

005100 C READ DATA FOR GAS INFU,RMATION C 005210 50 wkITE(6.1170)

HLA0(5,1160)KGTRWT IF (KGTRwT.EO.0) GO TO 70 UTRW=(DEHIFL.SBLDR/1440.1/DEH1FL 055250 IF(KGTRwi.EO.2) GO TO 60 mRITE(6,1190) 90 TO 80 60_ GTRw=0.pboGTRW aHITE(6,12003 GO TO An

- 70 u T R a = 0 '. 0 WHITE (6,1210)

~80 SRH=GTRW*0EMIFL+(S8 LDH.EDLLR)/1440 WHITE (6,1220) SRB.

READ (5,1010) WORD. TAU 1 j

WRITE (6,1010) WORD, TAU 1 1 READ (5,1010) WORD, TAU 2 WRITE (6,1010) word, TAU 4 HEAD (5,1ul0) WORD, TAU 3 WHITE (6,1010lWORD,TAUJ WRITE (6,1230)

OWPRF=1.0 AXIRF=i.0 \

AAPRFai,0

, CHIRFai.0 CHPRFs).0 CLIRF91.0 CLPRF=1.0 F.HI RF = 1. 0 FHPRFai.0 CAIRF=j.0 CAPRF=1.0 NEAD(5,1250)wuRD.GWHRL IF(GWbRE.uT.0.0)GwPRFql.0-GWHHE/100.

WRITE (6,1260)wuRD,GwPNF HEAD (5,1210) WARD.FHCH5E,FUHRE IF (F HCH AE .GT . 0.0 ) FHIRt = 1.0-FHCHHE/10 0.

3-18

IF(FHHRE.hT.0 0)FHPRFgl.0-FHHHE/100.

wRI TE (6,12dD ) W ARD ,FHINF,F HPRF NEAD(5,1270) WARD,AxCHNEeAAHRE IF(AXCHRE.GT.0.0)AXIRLul.0-AXCHRE/100.

IF(AXDRE.uT.O.0)AXPRFT1 0-AXHHE/100.

WRITE (6,1280) WARD,AXINF,AAPRF MEAD (5,1010) WORD.CONVul -

WRITE (6,1010) WORD,CONVOL )

WRITE (6,1290) '

NEAD(5,1370) WARD CACHHE.C8HRE6CFM IF(CACHRE.GT.0.0)CAIRt:1 0-CACHRE/100.

IF(CANRE.GT.O.0)CAPRFyl.0-CAHRE/100.

IF(CFF.EG.O.0) GO TO YO , .

KID =1 005710 WHITE (6,1300)CFM,PURT4M GO TO_iOO 90 KID =0 wkITE(6,1310) 100 IF(FFCDM.uT.O.0) GO Tu 110 mRITE(6,1320)

GO TO i20 110 FISCD=i.0-FFCDM WRITE (6,1330)FIBCD 120 IF(TBD,E0.0 0) GO TO 130 .

CON =0 01 Un5840

.. WRITE (6,1340) CON

,GO TO i40 130 C0n=1 0 wkITE(6,13401 CON 140 READ (5,3350)dARD.CHCMHE CUHHE,EtiP IF (CHCHRE.GT.0.01 CHIRL:1 0-CHCHHE/300.

1F(CHHoE.uT.O.0)CHPRFs .0-CHHHE/100.

EN=2.0.ENP wkITE(6,1JbO)EN 4HITE(6,1200) WARD,CHIHF,CNPRF READ (5.1310) WARD.CLCHME,CLHRE,PNOV1 IF(CLCHRE.6T.O.01CLIHt=1.0-CLCHH6/100.

IF(CLHRE.6T.O.0)CLPRF51.0 CLH1E/100.

IF(PNOV1.LT.1 0) Go To 150 WRITE (6,1380) WARD,PNDV1,hbRD.CLIRF.CLPRF GO TO 160 150 WHITE (6,1390) 160 WRITE (6,1400)TRLK READ (5,1010) WORD,FVN I

WRITE (6,1010) WORD,FVN HEAD (5,1010) WORD,FEJP

,FEJP=1.0-FEJP/100

\ WRITE (6,1010) WORD,FEJP HEAD (5,1010) WORD,PFLAUN IF(PFLAUN.LE.0.0) WR Il E (6 ',14 30 )

C On6210 C CONVESSION OF UNITS 006220 C On6230 T05TFL=TOSTFL*i000000, 0g6240 wL1=WLI.1000 On6260 CONVOL=CONVOL*1000000s 066270 CFH=CFHo1000 0662A0 TbD=TBDe163 066290 PHIVOL=Pd1VOL*1E3 On6300 OtHIFL=DEMIFL*500.53 066310 SbLDR=sPLuHe.3476 056370 EDFLR=F0FLR*.3476 066310 3-19

DwFLH=DwFLH+.3476 Oh6340 DwFL2=DwFL2*.3476 006350 CBFLR=CRFLH'500.53 On6360 C

C H3COPW IS THE PWR TRIi!UM PRIHANY COOLANT CONSENTRATION IN UCI/G9 On63A0 C

H3COPW=1.0 056390

'H3PRPW=0.4*POWTH TPLRPWsSOLDR* CWA *CWPD?EDFLR* EDA *EDFDeUWFLH*DWM*DWFD+DWFL2*DW2*DWF2006400 H3RLPW=TPLRPWeH3COPW'J.??T 006410 IF(H3RLPW.GT.0.9*H3PRCW)M4RLPWs0.9*H3PRPW H3RLG=H3PRPw-H3RLPW _

006430 DIV=10.**(INT (ALOG10(U3RL93)-1) Oh6440 IDIV=pIV 006450 IH3RLG= INT (H3RLG/DIv+9.5)*IDIV 056460 IF(TAU 3.Eu.0.) TAU 3=.01 066470-SR88SRR*500.53 006440 PE=365'./ tau 3 066490 T1=3.1557ET/EN#0PFRA 066500 T3=3.145TE*07/PE 006510 T4= Taut *86400. 056520 T5= TAU 2 *R6400. 056530 00 190 !=1'13 190 OECOH(I)=bLCON(I)*360V.

00 200 Isle 13 200 CONCP(I)=AP1(I)

IF (PO=TH.LT.3 0 0 0. .UR.f 0m TM.bT.3800. ) GO Tu 21u It(PR!v0L.LT.5.0E5.OR,PRIVOL.GT.6,0E5) GO TO 410 IF(UEFjFL.LT.3.2E4.OR, DEN 1FL.uT.4.2E4) G0 To d10 IF(50 Lop.LT.250. 0R.5pLOR.GT.1000.) GO TO 210 IF (CBFL P.GT.7500 1 GO TO 210 IF(AGTRwT.Gl.0) GO TO 210 60 TO 240 210 AFPTEG=1.0 HNG2=(SRLDR*DEH1FL* GTRw)/PRIVOL 006660 HHAL2=(DEH1FL*0.99+0.Y1*SULDR)/PHIVOL HK2G=161.16*POWTH/PRIVOL 00 230 I 1,13 IF(I.QT.11) GO TO 220 CONCP(!)=CuNCP(I)*RK29*(.0009+DECOHf!))/(8NG2*DECOH(I))

00 TO 230 220 CONCP(I)=CONCP(!)*RK2ue(0 06T+DECOH(I))/(NHALd+DECOHt!))

230 CONTINUE 2'40 IF ( T BP'. E4. 0. 0 ) GO TO $80 C

C PWTYPE=1.0 IS FOR PWR5 wITH U-TUBE STEAH YENEHATORS C

PwTYPE=1.0 0667A0 00 250 I=1,11 250 CONCS(I)=xP2(I)

CONCS(}2)=1.8E-6 CONCS(13)=4 8E-6 IF(AFPTEG.EG.1.0) GO TO 300 IF(WLI'.LT.4.0E5.0R.WLA.GT.5.0ES) GO TO 300 IF(TOSTFL.LT.1.3E7.OR.TOSTFL.GT.1 7ET) GO TO J00 IF (TBP' LT.S.0E4.OR.TBO.GT.1 0ES) GO TO 300 IF ( F F C DM . G T . O . 01 ) GO TO 300 60 TO 340 C

-C PwiYPE=2.0 15 FOR PWRy w1TH ONCE-THROUGH STEAN GENERATOHS C

280 PmTYPE=2.0 00 290 I=1,11 3-?0

290 CONCS(I)=&P2(I)

CONCS(12)=5.2E-8 CONCS(13)=1 6E-7 IF(AFPTEG.Eu.1.03 GO TO 300 l IF (TOS TFL.LT .1.3E7.0R e TOSTFL.hT.1.7ET) GO TO J00 l IF(FFCDH.LT.0.55.0R.FtCDH.GT.0.75) GO TO J00 1 60 TO 340 300 CONTINUE IF.(FFCDH.GT.0.01.'ANDeFFCWH.LT.1.0) FFCDH=0.2 RHAL3=(TBD*FNRTSC. 9'GON'TOSTFL*FFCDH)/WLI DO 330 I=1 13 j IF(I.QT.11) GO TO 310 i CONCS(I)=CONCS(!)*1.5E7/TUSTFL*(CONCP(I)/XP1(..)) l 007140 GO TO 330 310 IF(PWTYPE.EG.2.0) GO 10 330 .

CONCS (I) =CONCS (I) + (4.hE5/ uLI) + (0.17 +DECON (I) ) / g RH AL3 *DEC0H g I) )

• 1 (CONCP '(I ) /XP 1 (I ) ) Oh7190 00 TO 330 320 CONCS (I) =CONCS (I) + (1.UE5/mLI) * (27 0 +0ECOH (I) ) / (HN AL3.DECOH (I) )

• 1(CONCP(I)/XPI(!))

330 CONTIhuE 340 PNOV=PNOV1/CONVOL*60.

C 067240 C THIS PART OF PROGRAM IS fur NOBLE GASES On7P50 C

J:0 DO 370 I=1,13 .

,X2(I)=(DECON(I)+PNOV/J600.)*T1 On7?A0 IF(X2(II.bT.30.) X2(I'=30 X3(I) = OECON(I)

• T3 0n7100 IF(X3(I).6T 30.) X 3 ( ! ). = 3 0.

X4(1) s.0ECUN(I)

• T4 007330 IF(X4(I).GT.30.) X4t!)=30.

X5(I) = UECON(I)

• TS 067360 XUK=X5(I) on7370 IF (X5 (I) .GT.30.) XDK=30.

IF (1.QT.11 ) GO TO 350 IF(I.QT.4) X0K=X4(It CTPR0(I)=(CONCP(I)*PRIVOL*CLFNG)/(DECOH(I).PNuv)*1.892E-5 057470 ACONT (I) =CT PR0 (I) + (1.iEXP (-X2 (I) ) ) 0n7480 l_ ASHI H (I) = (CUNCP (I ) *SRD) /DECOH (I)

• 4.54E-4 * (1.-$XP (-X3 (I) ) ) 007490 l- AXBL(I)=CONCP(I)*AUXLH+.1657o0PFHA 007500 CuCP(I)= EN

• PNOV * (CTPR0 (I) *T1/3600 +CTP80 (I) * (E XP (-X2 (I) )-1.10 n 7510 1/(DEC0HtI)+PNOV)) 0n7520 CBSP(I)=EN*MCONT(I) 007530 CBL(I)=CHCP(!).CBSPtIl 0n7540 ASHIMC(I)=PE*ASHIH(I)*EXP(-XDK)*0PFRA ASHIMS(I)=EH+CONCP(IltPRIVOL*4.54E-4*EXP(-XDK) 067570 EJT(I)=CONCS(I)*TOSTFF)e3.Y77*0PFRA T6L(I)=CONCS(I)*T8LK* .977*0PFRA 0575R0 FHBL(I)=0.0 BV0G(I)=0.0 Og7590-TEST =1.0 0n7600 IF(CHL(!).LT.TESTICBLII)=0.0 On7610 IF ( ASHIMS(I) .LT. TEST) ASHIMS (I)=0.0 Oh7670 -

IF ( ASNIMC (!) .LT. TEST) $5HINC (I) =0.0 007630 IF(EJT(I).LT.TESTIEJT(I)=0.0 Oh7640 IF (TBL (I) .LT.TESTI THL 8 I) = 0.0 007650 IF(Ax8Lf!).LT. TEST)4X8L(I)=0.0 067660

.00 TO 370 C uh7680 C' THIS PART UF PROGRAM IS FOR IUDINE 067690 3-21

J 350 CTPR0(I)=(CONCP(I)*PRIVOL*CLFI)/(DEcon(I)*PNDW)*1.892E-5

-ACONT(I)=CTPR0gI)+(1 6EXP(-A2LII)) 007720 JmJ+1 . .

.AX8L(1)=(RNAX(J)+RNAAh(J))eCONCP(I)*AAIRF 1 FHbL(1)=(RNFH(J)+RNFHb(J))*CONCP(I)*FHIRF ~

ASHIMC(!)=0 0 007740 ASHIMS(!)=0.0 057750 GWCP(I)= Eh

• PNOV * (CTPR0 (1) *T1/3600, *CTP8W (I) + (EXP (-X2 (!) ) 1. ) C 07760

.1/(DECOH(1)+PNOV))*CLIHF.

JCbPP(I)=ENP*ACONT(ItohMIRR I

CBSP(!)=RNS(J)*CONCP(4)eCUIRF CHL (I) =CBCP (I) +CBSP t II.CBPP (I)-

EJT(I)=1 7E3*CONCS(I)*C0h?FEJP 4- TSL(I)=(HNT(J)+RNTStJA)*CONCS(I)* CON bV0Gt !)=CONCS(I)*TBD*tVN#J.97T*0PFRA 067830-IF(KID.EQ.0) GO TO 360:

'DLAK=(CFM*60.*CACHRE*0.01*0.7/CONVOL)+DECUHtI)

EXX2=DLAK*PURTIM' 057860 IF(EXX2.GT.30.) EXX2=f0, EAPF=ExP(-EAX2) 007880 E APC = 1 *. -E APF. Oh7890

• LLSS =CHIHF *CONCP (! ) *PH I VOL*CLF I+1.892E-5/DL A6*E APC

. col (I) =CHPP (!) *EXPF +ELSS epNP.CBCP (I) + (1.-FURI Au/ (8760.*0PFRA/FN) ) +

32* CHI?F*0.lboCONCP(I)*PUHIIM/(24.*32.5)*EAPC/(DLAK+PURTIM)

U 360.'p+(CBSP(I)-2.*CHIRFe0.16*LONCP(I)*P

.T ES T = 0'. n 0 01 NTIM/(24.*f2.5))

'If(CRL(I).LT. TEST)CBL(I)=0.0 007930 IF (EJT (I) .LT. TEST)EJT(I) =0.0

~

-0n7940 IF(R V0G(1).LT. TEST)BVuG(I)=0.0 067950

.IF(T8L(I).LT. TEST)TBLII)=0.0 . 007460

'IF ( AXBL(I).LT. TEST)AX?L(I)=0.0 0n7470 IF (FH8L (I) LT. TEST)FHdL (!) =0.0 370 CONTIhuE' Nh1G=1 06H0PO HSIG=13 CALL;5IGF2(CBL,HSIG,NS G) 008040 CALL SIGF2(ASHIMS,MSly,h51G)._ OhA050 CALL- SIGF2 ( ASHIMC,MSlu,NSIG) 008060 CALL'SIGF2(EJT,MSIG,NSIG) 058070 CALL SIGF2(bv0G,MSIG,NSIG) 0680RO.

CALL'SIGF2(TRL,MSIG,NblG) 05A090 CALL SIGF2(AX8L,MSIG,NSIG)' O h A100 .

CALL SIGF2(FHOL,MSIG,NSIG1 00 380 I=1,13 TOT (I) =CBL (!) +EJT (I t + } BL (1) + AXBL (I) +FHBL (I) + 0XOG (I) + ASHIHC (I) + ASNI 1MS(I) '!

380 CONTINUE CALL SIGF2(TOT,HSIG,NSI6) 008140

. WHITE (6,1440)

WHITE (6,1450)NAME.

' WRITE (6,1460) dkITE(6,1560)

WRITE (6,1480)

DO 385 !=12,13

.Wh1TE(6,1495)NUCLIDt!1,CONCP(l),CONCS(1),tHBL(I),

'ICbL (I) . AXBL (!) , TBL (Il e qV0?(I) ,EJT (I) , TOT (1) l385 CONTINUE' LWHITE(6,1*BU)

WRITE (6,1510) IH3RLG-WHITE (6 1440)

WRITE (6,1450)NAME wkITE(6,1460) j 3-22

W H I T E ( 6,14.7 0 )

WRITE (6,1460)

GASTOI=0.0

'00 390 I=1,11 WHI TE (6,1490) NUCLID t II,CONCP (I) ,CONCS (!) , ASHIMS (I) , ASHIMC (I) ,

ICdL (I) , AXHL (I) ,TBL (11 e 8V0b (I))EJT (I) , T OT (1)

GAST0T=GASTUT+ TOT (It 390 CONTINUE

' DIV810'.* * (INT ( ALOG10 (3 AST QT) )-1) 068360 GAST0T=AINT(GAST0T/DIV,+0.5)*DIV 008370 WHITE (6,1500) GAST0T WRITE (6,1480)

WHITE (6,1520) wkITE(6,1440)

WRITE (6,1450)NAME kRITE(6,1530)

WRITE (6,1540) nRITE(6,14801-QN=8760.*UPFRA/EN 008560 C

C THIS PART OF PROGRAM ,1,5 fur THE PARTICULATES

.C 00 430 I=1,18 PRC04T(I)=PLHP(I)/(8700.*0PFRA)

IF (PNov GT.O.0) GO TO 41u ,

PCi1CP(I)=0.0 Un8600 PCUSP(I)= Lit *PHCONT(I)*0H*LHPRF bu.TO 4?O 410 PLoCP ( I ) = ( EN* (QH*PRC01T

' '! ( I ) -PRCO,NT.( I ) /PNOV * ( 1.-EXP,(-PNOV eQH ) ) ) )

1*CLPRF PCBSP (I) = (EN'(PRCONT,(13 /PNOV* (1 0-EXP (-PNOV*4Q) ).) ) *CHPpF 420 PCHL(I)=PCnCP(I)+PCB5P(I)

PAXdL(I)=PAARP(I)*AXPnF PFHBL(I)=PFHHP(I)*FHPdF Powl(I)= POWS (I)*GWPRF IF(KID.EO.0) GO TO 439 PDLAK=CFM*o0.*CAHRE*0,01*0.7/CONVOL PLXX2=PDLAK*PURTIM 068710 IF (PEXx?.bT.30.) PEXX2=30. ,

PEXPF=EXP(-PEXX2) On8730 PEXPC91.-PEAPF On8740 PELSS=PRCONT(I)/PDLAK!PEXPC*CHPHF PC8L (I) =PCBbP (I) *PE XPF +PELSS*EN +PCBCP (I) * (1.-tURT IM/ (8760.*0PFR A/E 0 6 8760 INJ) 008770 430 CONTIh UE M51G=2 008740 NSIG=18 CALL SIGF2(PCBL,MSIG,NSIG) 008810 ,

00 440 !=1,18 PTOTP(I)=PCdL(I)+PAXB,h(I)tPOWL(I)+PFHBL(I) 440 CONTIPUE

. CALL SIGF2(PTOTP,MSIGsNSIG) Oh8850 DO *SO !=1,18 WRI TE (6,1550 ) PPART (!) e PGwL (I) ,PCBL (I) ,P AXbt (I) ,PFHBL (I) ,PTOTP (I) 450 CONT IrvuE WHITE (6.14HO)

STOP C

C FORMATS FORMAT) FORMATS F0kNATS C

1000 F0HM AT '( 32 A ,8 A4,1?X , A4) 1005 FORMAT (16A,"RLOWD0hN IS PdOCESSED THROUGH CONDENSATE DLHINn) 3-23 L_

. -. = .- .-

1007 FORMAT (16X, RLown0WN 1010 FORMAT (16x ,"13 A4, A2,F10,5)IS NOT PHOCESSED THRduGH COND. HEMIN.")

1020 F0HMAT(16A,"PLAHT CAPACITY FACTOH",T74,"0 80")

1030 FORMAT (16A." PERCENT FUEL *ITH CLADDING DEFECTf",T74,F7.53 1040 FORMAT (36A,f8.4,35X,IA) 1050 FORMAT (16X,"RLOWDOWN HATE (THOUSAND LbS/HM)",44X,F9.5) 1060 FORMAT (15X,4A4,A2,8xete.0) 1070 FORMAT (20X,F8.0,2(5X,f.8.0))

1080 FORMAT (27X,F6.2.14x,F3.2,18X,F6.2) 1090 FORMAT (/," LIQUID WAyTE ANPUTS")

1100 FORMAT (30X,"FRACTAON FRACTION COLLEC TION DECAY"/8X'," STREAM 1 ELOW RATE OF PCA DISCHARGED TIME TIME",5X," DECONTAM PINATION FACTORS"/20Xees(GAL / day)"i23X,"(DAYS) (DAYS)",7X, 3"I",8X,"CS",8X,"0THER)")

1110 FORMAT (2X,4A4,A2,1PE982,1A,4(OPF8.4,2X),3(IPE9.2,1XI)

'1120 FORMAT'(15X,4A4,A2,8x,F8.0'p7X,F6 4) 1130 FORMATg70X,F10.5) 1140 FORMAT (2X,"8 LOWDOWN",10A,1PE9.2,11x,F8.3,2x,2(F8.3,2x),

13 (IPE9.2 1X) )

1150 FOHMAT(2X," UNTREATED AL0=90WN",1PE9.2 11Xt" 1 000 0 000 10 000 1.00E+00 1.00C+0U i.00E+00")

1160- F0HMAT(2A ""EGENEHANT SOL 4 ",1PE9.2 14X ,0PFb. 3,2x ,2 (F8.J ,2X ) ,

13(IPE9.2,1X1) 1170 FORMAT (/e" GASEOUS wtSIE INPUTS")

1180 FORMAT (79A,11) 1190 F0HMAT(16A,"THERE IS SMNTINUOUS STRIPPING OF LULL LEToowN FLOW")

1200 _ FORMAT (16A "THERE IS 90hTINU005 low VOL PURGE OF VOL. CONTROL TK")

1210 FOHMAT (16A,"THERE IS NOT C U NTINddUS SIRIP(INb 0F FULL LETUwN FLO*H

1) . . ~

1220 FORMAT (16X,'" FLOW RATE' THRO U GH' GAS ~ STRIPPER(6fM)",19X,F9.5).

1230 . FORMAT (16A,"PRIHAHy CUOLANT LEAK..TO AUXI,LIARY BLDG (LR/UAY)",T72, 1"160.08n00"I-1250 FORMAT (16 A ,4 A4,6X ,F3.V) 1260 FUdMAT(16A,4A4,4X,"PANTICULATE HELEASE FHaCT!uN",6X,F10.5) 1270 FOHMAT(16x,5A4,10x,F3pi,6X,F3.0) 1280 1FuHMAT(16X,bA4," IODIN 9 RELE ASL FRACTION",11X et10.5/36X,"PARTICUL AT 1E RELEASE FHACTINe.,6A,F10.5) 1290 FORMAT'(164,"FRE0HENCY OF PRIMARY COOLANT DEGA? SING (TIMLS/YR) 1,"2 008h0"/16X,"PHIMA5Y TO SECONDARY LEAK RATE (L8/ DAY)",T72,".T14 l

2" 75.0 0000") '

, 1300 . FORMAT (16x,"THERE IS A KIDNEY FILTER"/20X,"CDHTAINMENT ATMOSPHERE 1 CLEANUP RATE (THOUSAND CFM)",T71,F1 0.5 20A,"PJRGE

/ TIME OF CONTAINM I

2ENT (NOURS)" T71.F10.4) l 1310 FuHMAT(16X,"THERE IS NOT 8 KIONEY FILTER")

1320 FORMAT (16X,"THERE IS NOT e CONDENSATE DEMINER$LIZER") l 1330 FORMAT (16X," FRACTION ADDINE BYPASSING CONPENSATE DEMINERALIZERes, 17 X,T74,F9.5) i 1340 FORMAT (16X," IODINE PANTITION FACTOR (G AS/LIQU,1,0 ) IN STEAM GENERATO l

'1R ",F7.5)

{

1350 FORMAT (16X,5A4,10X,F3,0,6X,F3 0,lix,F3.0) 1360 F0HMAT(16X," FREQUENCY OF CNTMT 8LDG HIGH VOL PURGE (TIMES /YR)s', j I T 73, F 8 '. 51 s 1370 F0HMAT (16X ,5 A4,10x ,F3,0,6 A,F3 0 14X ,F8.2) I 13RO.-FORMATjl6x,5A4," RATE (bFP)",24X,F11.5/16X,5A4," IODINE kELEASE FRACT l IION",11X,F10.5/36x,"PtRTICULATE RELEASE FNACTAON",6x,F10.53 1390 F0dMAT (16A "THERE IS MOT A CNTHT RLOG LOW VOLudE PURGE")

1400 FORMAI(16A," STEAM LEAS TO TURBINE HLDG (LoS/MO)",19x,F10.b) 1 I 1430 FORMAT ("0 -'sX,"THERE IS NOT AN UN SITE LAUNDyy") j 1440 FORMAT (1H1:  ;

1450' FuRMAi(16A,ba4) i 1460 F0HMAT(IHO,b7X," GASEOUS HLLEAbE RATE - CU"IES PER YEAR")

1470 FuHMAT(1HO,11x," PRIMA $Y",4X," SECONDARY",7X,"60S STRIPPING",11X, 1"dOILDING VLNTILATION"/12A,"CuGLANT",5X,"900LONT",5X 21("-"),

l 3-24

74X,30(n "),5x nBLOWD0pN AIR EJLCTOR TuTAL"/10X,"(HICROCI/GM)(M 31 CROCI /GM) SHUTDOWN CONTINUOUS REACTUR AUXILIARY TURBI*E 4 VEb7 0FFuAS EXHAUSTH) 1480 FORMAT (1HO,130(" "))

1490 FORMAT ("O ",A8,2(2X,1(E10 3),8(3X,1PE8.1,1x))

149$ FORM AT (0 ", AB,2 (2X,10E10 3) ,12X,T (3X,1PEv.1,1X) )

1500 FORMAT (1HO," TOTAL NuBLE GASESu,101XelPEO.1) 1510 FORMAT (1HO,30X," TOTAL H-3 RELEASED VIA GASEOUb PATHWAY = es,14,n CI 1/YR"//31X,"C-14 RELEASED WIA GASEQUS PATHWAY ? T.3 CI/YR"7/31X, 2HAR-41 RELEASED VIA CUNTAINHENT VENT = 34 CZ/YRn) 1526 FORMAT (IHO,"0 0 APPEARING IN THE TABLE INDICAlES RELEASE IS LESS 1THAN 1".0 CI/YR FOR N09LE 9AS, 0.0001 CI/Y$ F05 In) 1530 FORMAT (1HO,54X, HAIR 805NE PARTICULATE RELEASE 5 ATE-CURIES PER YEAQn n) 1540 FORMAT (1HO,16X," WASTE GAS",16X," BUILDING VENTI.LATION"/2X,"NUCLIDEH 1,28X,uSYSTEMu,14x,nREACTO$ AUXILIARY FUEL HANDLG",7X,nTOTAL")

1550 FORMAT'(1H0,AB,28X,1PEU.1,11X,1PE8 1,4X,1PE8.1,4X,1PE8.1,10X,1PE8.1 1) 1560 FORMAT (1H0 11X,HPRIH AHy'8,4X,s SECONDARYn,2bX,nuu!LDING VENT IL ATIO1" 5 "),6A," BLOWDOWN AIN EJECTuR 1/12X,"C00LANT",",5x,nCoulANT",1 2 TOTAL"/10X (MICROCI/GM) X,44(n (MICROCI/GM) n,12X," FUEL H ANDLG REAC10 3R AUXILIARY TURBANE VENT OFFGA) LXHAUST")

END 008930 eDECK SIGF2 008940 SUBROUTINE SIGF2(RLPTsMSIV,NSIG) 068950 01MENSION RLPT(NSIG) On8960 IF (MSIG.Eu.2) Go-TO JO 068970 uO 20 1=1*NsIG n58980 IF (RLPT (1) .EU.0.0 3 GO TO 20 '

0n8990 I F ( I . G T .11') GO TO 10 C- .THIS PART OF SUBROUTIHE 14 FOR NOBLE GASEh. 009010 DIV*10.**(IkT(ALOG10($LPT(I)))-1) ~ 009020 IF (RLPT(I).LT.10.1 DAvs1 00 069030 RLPT(I)=AINT(RLPT(It/91V+0.53*DIV 009040 i GO TO Pn 009050 C THIS PART OF SUBROUT!dE Ib FOR IODINE 069060 10 CONTINUE 0n9070 ISUB=2 0n9080 IF (RLPT (I) .GT.1 0 31SUP=1 069090 DIV=10.**(IHT(ALOG)D(HLPT(I)))-ISUB) 004100 RLPT(I)=AINT(RLP1(It/9IV+0.5)*DIV 069110 20 CONTINUE 069120 30 CONTINUE 009130 C THIS PART UF SUBROUTIdE Ib FOR PARTICULATts 069140 00 50 I=1,NSIG 069150 IF (RLPT (I) .EO.O. ) GO TO SO 069160 DIV=10.**(INT (ALOG10($LPT(I)))-2) 009170 RLPT(I)=AIN1(RLPT(I)/VIV+V.5)*DIV 069180 50 CONTINUE 069190 HETURN 069200 END 009210 3-25

,_ w-mma - - - - - - -

FIGURE 3-4 PROGRAM LISTING FOR LIQUID DETERMINATION eDECK PGALELO . 000760 C GALE CODE FOR CALCULOTINV LIuu!D EFFLUENis F50M PWRS. MODIFIED C AUG. IgT9 TO IMPLEMENT APPENDIX I TO 10 CFR P9RT 50. REALTOR C WATER CONCENTRATIONS hALCULATED USING METDODS OF DRAFT STANDARD 060300 C ANS 23T HRAul0 ACTIVE MATE 51AL) IN PRINCIP$L FLUID STREAMS OF 060310 C LIGHT WATER COOLED NUCLEAW POWER PLANTSn DRALT DATED MAY 20, 19T4000320 C H0DIFIED EDITION OF OHIGEN PROGRAM TO COMP U TE EFFLUENTS FROM 4WR 000330 C ANd PWR RADWASTE SYSTpHS Oh0340 C 000350 C STATENENTS

• PROGRAM PuALELQ* AND

• LEVEL 2* ARE FOR CDC USERS.

C FOR IBM USERS DELETE THtS9 STATEMENTS.

C PROGRAM PGALELQ (INPUT,0VTPUT, TAPES = INPUT, TAP 96=0UTPUT, 1 TAPEq)

REALLETDWN,NOGEN Oh03 0 REALLETDWA Ch0390 COMMON /MATHIX/A(2500)eLOC(2500),NON0(800),KD(9001 000470 LEVEL 7,A. LUC,NON0,KD 000380 COMMON /CONST/MMN. ERR,NZERU COMMON /EQ/XTEMP(R005,ANEn(10,800),0(800),P(8001 COMMON /FLUXN/ REGENT,DAS(800),lLITE.IALT IIOT COMMON /00T/kUCL(900t CUMM0h/CUNC/PCONC(R001,5CUN(800),RINV(8003 COMMON / COOL /REACTR.P0wl,SPLDR,BLWDWN,FPEF,HEF,EUFLR,DFIED DFCSED, 1 DFED,DFlow,DFCSDW,0FuW,EuA,0*A, CWA,DFCP,UFICM.DFCSCM, 2 DFCW,0 FICA DFCSCW,BDIFR,LDFD,DnFD.CWFD,CHFD,l S,TE,TO,TC,TCM, 1 TST08C,TSTURD,TSTORd,0=F L2,Dw2,DWF2,12,ThT odd,0F ID2,DFCSD2,DFD2, 4 PFLAUN,0wFLR COMM0h/APC00L/RGWFR,OLIRG,DFC5RG,DFRG,TRG,TSTVRR,RGFD 000660 COMMON /90TES/RFNRT On0670 COMMON /CONP/PWCONC(800),SLUTv(800),5 COT (800)

DIMENSION WORD 15(4),wuRDie(5), WORD 56(14),*0RD#(2),REACTR(7)

C 050A70 C READ h0CLEAH DATA AND C0hbTRUCT TRANSITION MAlRIX C 000890 CALL NUDATA(NLIBE)

C 000910 00 20 I 2,lTOT 060920 NON0(I)= NUN 0(I)+NON0(1-1) 000030 20 KD(I)=KD(II+NON0(I-1) 050940 C

C HUILT-IN PAHAMETERS C

PF=0.05 TBLKal700.

C HMN=0 MZER0=pl DO 30 J:1,800 PCONC(J)=0 0 0n1170 SCON(J)=0.0 Un1190 RINV(J)=0.0 0n1200 30 CONTIh0E C 0n17s0 C HEAD DESCdlPTION OF HEACTUR AND PADwASTE TREAIMENT PLANT 0n1760 C 0n1270 PHINT 9n26 0n1200 NEA0 9610,HEACTR.TYPL On1300 PHINT Qolo,HEACTR, TYPE On1310 REAU S h l l , 'a OR D S 6. P O W 1 Ohl370 PHINT 90ll m0R056.Poal ,

on1330

PHINT on27 001350 C

C HEAD DATA FOR LIQUID ANFUWHATION C

NEAD 9022, WORD 56,PCVOL On13A0 PRINT 9022, WORD 56 PCVWL 001390 NEAD 9612,w0RD56,LETDwN 001420 PRINT 9012, WORD 56,LETVWh 001430 NEAD 9612, WORD 56 CBFLU 001440 PRINT on12, WORD 56,CBFhR 001450 READ 9611, WORD 56,NOGEH 001460 PRINT 9011, WORD 56,N0 BEN 001470 READ 9622,wuRD56 STMFH 0n1480 PRINT 9022 WORD 56,STMER 001490 READ 9622,WURD56,WLI 0n1520 PRINT 9022,w0RD56,WLI 001530 SCv0L=NOGEN*WLI 0h1940

, PRINT 9029,bCVOL 001550 NEAu 9655,BLWDWN,KFNRT 0615A0 PRINT 9051,BLWDWN 001590 RFNRT=I.0 IF (KFNRT.EU.2) RFNRT90 0 .

PRINT 9041 0n1620 HEAD 9612, WORD 56, REGENT 061630 PRINT 9012,n0RD56, REGENT 051640 IF(8LnDWN.EQ.0.0) GO 10 40 FPEF=0'n05 MEF=0 61 061670 PRINT 9030,FPEF,HEF 001680 GD TD go 40 F P EF = 1',0 nEF=1.o 0n1710 -

PRINT on30,FPEF,HEF On1720 50 HEAD 9520, WORD 56,FFCDN PRINT 9020, WORD 56,FFCuM U01750 IF(FFCDM.LT.0 001) GO TO 60 DFCHm16.0 On1770 DFCBCS=2.0 on1780 GO TO 7n 60 WFCR=1.0 DFCBCS=1.0 061810 70_ READ 9656, WORD 18,S9LD6

'CWAul.6 001860-NEAD 9614,UFICw,DFCSch,DFCW 061870 READ 9615,TC,TSTORC,CwFD 051880 PRINT 9045 0n1890 i

PRINT 9016 0n1900 PRINT 9017, WORD 18,SBLVR, CWA,CWF0,TC,TSTORC,DF(CW,DFCSCW,DFCW On1910 HEAD 9h13, WORD 18,EDFLU,WOBD8, EDA 0n1920 READ 9n14,DFIED.DFCSEy,DFLD 061930 HEAD 9615,TE,TS,EDFD 001940 PRINT'9017, WORD 18,EDFLR,EUA,EOFD,TE TS,DFAED,pFCSED,0 FED On1950 READ 9613,WURD18 DwFLY,w0HD8,DWA Un1960 READ 9614,0FIDW,DFCS0W,UFDW 001970 NEAD 9615,TD.TSTORD,UWFD on1980 PRINT 9017, WORD 18,DWFLR,DWA,DWF0,TD,TSTORD,DFADW,DFCSDW,DFDW 0n1990 NEAU 9n13,duRD18 DwFLf,WOMD8,pW2 On2000 NEAD 9614,DFID2,DFCSO4,UFW2 002010 READ 9015,T2,TSTOR2,DeF2 .

0n2070 PRINT 9017,>0RD18,0WFL2ech2,0wF2 T2,TSTOR4,DF(02,DFCS02,DF02 OhPn30 HEAU 9637,dDTFR 062040

-HEAD 9616,UFIC't,DFCSCM,DFCM nn2050

' READ 9615,1CM,TSTORe,CMFD 052060 3-27

1 l

l HEAD 9537,H6WFR .

052070 READ 9016,DFIRG,DFCSRu,0FHG Oh20R0 HEAD 9615,TNG,TSTORR,5GFD 002090' !

IF(WLWDWN.EQ 0.0) GO 10 75 l BUFR=dLWUWH+1E3*RDTFR/0 3476 .

On2110 PRINT on34,bDFR,CMFD,TCP,TSTORB,DFICH,DFChCH,WFCH 062120 bufR=BLWDWN#1.0E3 EARS (1.-DDTFH)/0 3476 ,052130

' PRINT 9035,bDFR 002140 IF (FFCDM.EG.O.6) GO TU 90 T5 IF(REGENT.Eu.0.0) GO TO 80 PRINT 9038,RGWFR RGFDsTRG6TSTORR,DFIRG,DFCSRG,DFRG 0521R0 GO TO en 80 HGWFR=6.0 PRINT 9038,RGWFR.RGFDsTRG,TSTORR,DFIRG,DFCSRS,DFRG 052210 90 IF (KFNRT.Eu.2) GO TO 100 FNRT50=1.0-1.0/(DFCM*WFCB) 002240 FNRTSI:1 0-1.0/(DFICM!DFCW) OhP250 FNRTSC=1 0-1.0/(DFCSCNe0FL8CS) 052260 GO'TO ilo 100. FNRTSO=1.0 FNRTSI=1.0 052290 FNRTSC=1.0 062300

'C -

052320 C READ DATA FOR GAS INFMRNATION C 002340 110 PRINT 9646 HEAD 95P1,KGTRWT 062360 IF (KGTRWT.EO.0) PRINT 9043 002370 IF (KQTRWT.t0 1) PRINI 9052 0023R0 IF (KGTRai.EQ.2) PRINT 9075 On2390 READ 9512.h0RD56, TAU 1 On2400 PRINT 9012, WORD 56,TAUL 052410

. READ.9612, WORD 56, TAU 2 0n2420 PHINT gn12, WORD 56, TAUR 002430 READ 9612, WORD 56 TAU 3 Oh2440 PRINT 9012, WORD 56,TAUJ on2450 ,

GWPHFai.0 AXIRFai.0 AXPRF=1.0 CHIHF=1.0 CHPRF=1.0 CLIRF=j.0 CLPRF=1.0 FHIRF=1.0 FHPRF=1 0 CAIRF=1.0 CAPRF=1.0 NEAD 9665, WORD 15,GWHRE IF(GWHRE.uT.0.0) GWPRF31.0-GWHRE/100.

PRINT'9066, WORD 15,GWPNF HEA0 9567, WORD 18.FHCH5E,~.'bHRE l

IF (FHCHRE.GT.O.0) FHIHF=1 0-FHCHaE/100, j IF(FHHRE.GT.O.0) FHPRt=1.0-FHHRE/100 2 READ 9667, WORD 18.AXCH5E,AAHRE  !

IF(AXCHRE.GT.O.0) AXI5F=1 0-AACHHE/100 IF(AXHRE.uT.0.0) AXPRL=1.0-AXHHE/100.

PRINT 9n68,w0RD18,AXI5F,AXPHF HLAD 9522, WORD 56,CONvuL 052600 PRINT 9022,h0RD56, CON?,0L 002610 HEAD 9669. WORD 18,CACHUE,CAHRE,CFM t IF(CACHpE.GT.O.0) CAINF81 0-CACHHE/100 IF(CAHRE.GT.n.0) CAPRt=1.0-CAtJRE/100 3-28

I HEAD 9671, WORD 18,CHCHHE.COHRE IF (CHCHRE.GT.0.0) CHIHF=1 0-CNCHRE/100 ,

IF(CHHRE.bT.0 0) CHPRLs1 0-CHHRE/100. I i

EN=2.0 PHINT 9072,EN 002680 PHINT 9068, WORD 18,CH1HF,CbPHF READ 9669, WORD 18.CLCH0E,CLHRE,PNOV1 IF(CLCHRE.uT.O.0) CLIRF=1 0-CLCHRE/100 IF(CLHRE.GT.0.0) CLPR6=1.0.CLHRE/100 IF(PNOV1.LT.1.0) GO TH 120 PRINT 9070,w0RD18,PNOV1,WQRD18,CLIRF,CLPRF 60 TO i30 120 PRINT 9073 130 PRINT 9064,TBLK READ 9620. WORD 56,FVN 062800 PRINT 9020,w0RD56,FVN 052Bf0 NEAD 9620, WORD 56,FEJ 002820 (FEJ=1.6.FEJ/100.

PRINT 9020, WORD 56,FEJ 002830 NEAD 9620, WORD 56,PFLAUN 002900 IF(PFLAUN.LE.0.0) PRINT 9048 002910 PRINT gn26 002920 C 002930 C CONVERSION OF UNITS 002940 C 062950 EUFLR*EDFLR*48.8 On2960 OWFLRzDWFLR*48.8 0n2970 UWFL2=DWFL2*48.8 062980 C On40R0 C -CALCULATE PHIHARY COOLANT CONCENTRATIONS C 064100 AFPTES=0.0 064120 00 140 I=1,ITUT -

140 PCONC(I)=PWCONC(I) 054150 POWA= POW 1 PCVOA*PCVOL*1E3 054160 LETDWA=LETUNN#500.53 004170 SBLDA=SBLDR*.3476 004180 CdFLA=CBFLR*500.53 064190 C

C CHECK TO SEE IF PRIHAHy PLANT PARAHETERS ARE WITHIN SPECIFIED 054?00 C HANGES 004210 C

IF (PON A.LT.3000. .OR.PMW A.WT.3 8 00. ) GO TO 150 IF(PCVOA.LT.5.0ES.0R.FCV04.GT.6.0ES) 00 Tu 150

! IF(LETDWA.LT.3.2E4.0R,LETWWA.GT.4.2E4) GO TO 150 IF (SBLD A .LT .25 0. .OR.5?LD A e GT.10 0 0. ) GO TO 150 IF(CBfLA.0T.7500.) GO TO 150 00-T0 i90 C

C CALCULATE PHIMARY COULANT ADJUSTMENT FACTURS C.

150 AF PTES=1 0 RHAL28 tLETowA*0.og.0.uleSdLDA)/PCVOA RCSRB2=(LETDWA*0.5+0.betSULDA+C0FLA*0 9))/PCVuA 064310 HCFP28(LLTUWA*0.48 0.V2*(SBLDA.CBFLA*u.9))/PCVOA HK2=16i.7b'POWA/PCVOA 00 180 Jul ITUT IF(PCONC(J).EO.0.0) GM TO 180 ,

144 = NijCL (J ) /10 0 0 0 004360 DL=01S(J)*3on0 004370 IF (NZ.EO.53.OR.N7.EQ.45) 90 TO 160 IF (NZ .EO.3 7.0A.NZ .EO.D5) uo TO 170 3-29

l PCONC(J)=PCONC(J)*RK2*(0.066.0L)/(RCFP2 UL) 60 TO 180 160 PCONC(J)=PCONC(J)*RK2*(0.067+0L)/(RHAL2+0L) 60 10-180 170 PCONC(J)=PCUNC(J)*RK2*(0.037eDL)/(RCSRB2+0L) 180 CONTINeJE 190 SBLDR9SBLUR*48.8 PCVOLapCVOL*1000.*0.T/62.6 054470 C

C CALCULATE SECONDARY Cu0LANT CONCENTRATIONS C

SCy0A=SCv0L*1E3 054490 ULWDWe=BLWDhN*iE3 064500 STHFA*STMfR*1E6 054510 EFCDA*FFCDM 054520

~

C C' CHECK TO SEE IF SECONDARY PLANT PARAMETERS ABb WITHIN SPECIFIED 064530 C HANGES 054540 C -

IF(8Lv0WN.EQ.0.0) GO TO 290 C

C PWTYPE=1.0 IS FOR pwR) WITH U-TURE STEAH bENEMATORS

-C PWTYPE=1.0 00 200 I=leITOT

-200' SCON(I)=SCUTV(I)

IF(AFPTES.EO.1.0) GO TO 250

>IF(Scy0A.LT.4.0E5.OR.)CVUA.GT.5.0ES) GO Tu 250 IF(STMFA.LT.1.3E7.0R.4TMF$.GT.1 7E7) GO TO 250 IF(BLWDWA.LT.5.0E4.OR,8LWUwA.GT.1 0ES) GO TO 250 IF(FFCDA.6T.O.01) GO TO 250 IF(FNRT SC.LT.0.89996 20 TO 370 00 TO 390 C

C PWTYPE=2.0 IS FOR PWR) WITH ONCE-THROUGH STEAN GENERATORS C

230 'PhTYPE=2.0 00 240 Isle 1 TOT 240 SCON(I)= SCOT (I) .

IF(AFPTES.EO.1.0) GO TO 2S0 IF(STMFA.LT.1.3E7.0R.)TMFA.GT.1.7E7) GO Tu 250 I IF(FFCDA.LT.O.55.OR.FLCDA.GT.0.75) GO TO 250 GO TO 390 C

C CALCULATE. SECONDARY COOLANT ADJUSTHENT FALTORb C

250 IF(FFCDA.GT.O.01.AND.[FC04.LT.1.0) FFCOA=0.2 RHAL3=gBLwDWA*FNRTSI+U.9*HEFeSTMFA*FFCDA)/SCvuA-IF(FFCDA.GT.0.01.AND.[FCDA.LT.I.0) FFCDA=0.1 j HCSRB3=(HLNDWA*FNRTSC*0.5'FPEFeSTHFAetFCDA)/SkVOA 054910

. N C FP3 = (RL w D W A *FNR T SO + V . 9 e FPEF *S T MF A *FF CD A ) /SC V.0 A 054920 IF (PWTYPE.EG.2.0) GO IO 330 RK3=4 5E5/SCVOA 064040 00 320 I=1,1 TOT IF (SCON(I) .EO.0.0) GO TO D20 NZ=NUClill/10000 055.120 plau1Sg!)*3000 005130

' IF(NZ.EO.53.OR.N7.EG.J5) ko TO 300 IF(NZ.EO.3T.0R.N7 EG.b5) 00 TO 310 SCON(I)=SCON(I)*RK 3 * ( U .1 T

• DL ) / ( RCFP3.UL ) * ( PC Ql1C ( I I /PWCONC ( I I )

GO TO 320 360 5 CON (I)=SCUN(I)*RK3*(0.1T*DL)/(RHAL3.DL)*(PC0hC(I)/PwCONC(II)-

3-30

F I

GO TO 3F4 310 SCON(I)mSCON(I)*RK3etV.15*DL)/(kCSRB3+0L)+(PQuNC(II/PwCONC(I))

320 CONTIbuE GO TO 390 330 NK3=1.0E5/SCVOA DO 360 !=lelTOT IF(SCON(I).EQ.0.0) GO TO 360 065260 NZ=NUCL(1)/10000 DL= DIS ~g ! ) *3600, 065270 IF (NZ.EQ.53.OR.NZ.EG.35) 90 TO 340 IF(NZ.EO.55.OR.NZ.EO.)T).90 TO 350 SCON(13sSCON(IleRK3*(14.0*DL)/(RCFP3 0L)*(PCONC(1)/PWCONC(I))

60 TO 360 340 SCON(I)=SCON(IleRK3e(?7.0*DL)/(HHAL3.DLle(PCQNC(I)/PWCONC(I))

~40 TO 360 350 SCON(I)=SCON(IleRK3*(7.g.ul)/(RCSRB3 0L)*(PCONC(I)/PdCONC(I))

360 CONTINUE GO TO 390

.370 RCSRB3=(BLwDWAeFNRTSC*0 5eFPEFesTMFAeFFCD61/SEV0A 055340 HK3=4 5E5/SCVOA D0.380 I=!,ITOT IF(SCON(I).LQ.0.0) GO TO 380 055410 NZmNUCL(1)/10000 IF (NZ.NE.3T. AND.NZ.NE ,55) GO TG 380 DL= DIS 005430 f I)*3600.

NFC=0 15 065460 SCON(I)mSCON(I)eRK3e(HSC+PL)/(RCSRR3+0L)+(PCQNC(I)/PWCONC(I))

380 CONTINUE 390 BLWDWN=BLvDWNo1E3/500,53 SCVOL=SCVOLe1000./62.! 055490 STMFR=STMFR*2000. 005500 DO 400-I=leIT0T If(PCONC(l).EG.0.0) GW TO 400 PCONC (I) =PCONC (II / (DI? (I) *1 6283E13) 095530 SCON(I)sSCON(I)/(DIS (Alel.6283E13) 0o5540 400 CONTINUE C- 005560 C COMPUTE REMOVAL CONSTANT POR CONDENSATE DLMINERALIZER C

IF(FFCDM.GT.O.01.AND.[FCDM.LT.1.0) FFCDM=0.1 CIXRC=(0.9eBLWDWNeRFNHT/0hCM.0.9eSTMFReFPEFeFLCDH)/(SCVOL*7.48 60.

1) 055620 CIXRCS=(0.5eBLWDWNeRFhRT/DFCSCM+0.5eSTMFR*FPEteFFCDM)/(SCVOL*7.4be005630 1 60.)

~

065640 IF(FFCDM.GT.0.01.AND.tFC06.LT.1.0) FFCDM=0.2 CIXRIO (0.9e8LWDWNeRFURT/DFICM.0.9eSTMFReHEFetFCDM)/(SCVOLe7.48e60065650 1.) 065660 00 410 I=lelTOT 005680 NZmNUCL(I)/10000 PR=CIXRC IF (NZ.EO.3T.OR.NZ.EQ.55) PR=C1XRCS IF (NZ.EO.53.OR.N7.EQ.45) HRzCIXHIB XZHJ=SCON(IlePR*SCVOL*0.02832 B(1)=5ZMJ 005730 410 CONTINUE C .

C CALCULATE HAD10 ISOTOPE INVENTORIES ON CONVEN5dTE RESINS C

CALL SOLVE 065790 00 420 I=le1 TOT 420 MINV(I)=ATEMp(I)

CALLS EFFTAW 005820 STOP 3-31

l C _. 00'6490 C FORMATS '

FORMATS F0HMATS 006500

'C- .

. 006510 9010' FORMAT (32A,7A4,16X A44. '0h6620

-9011 FORMAT (16X,13A4,43,F9p4) 056630 9012 FORMAT (16X,14A4,F8.4 Oh6640-

.9013 FORMAT (15X,4A4,A2,AX,t8.0,1X,A4,A2,F6.4) 9014 FORMAT (20XtF8.0,2(5X,F.e.01) .

006660-9015 FORMAT (27X,F6.2,14x,Ft.2 18X,F6.2) 0h6670 9016 FORMAT ("0",30X,"FRACT40N FRACTION. COLLECTIQM DECAYu/8X," STREAM 0666R0 1 ELOW RATE- 0F PSA PISCHARGED TIMs. TIME",10X,"DECONTA006690 2MINATION FACTORS"/20Xa"(GAL / DAY)"23X,"(DAYS) (DAYS)",7X,- 006700 3"I",83,"CS",6X,"OTHER5") 006710 9017 FORMAT (2X.4A4,A2,1PE9e2,15,4(OPF8 4',2X),3(IPEY.2,1X))

9020 FORMAT (16X,14A4,F8.4) 066730 9021 FORMAT (79X,II)' 056740.

. 9022 FORMATi16X,14A4,F8.4) 066750

-9026 FORMAT (Idl) . .

066870 902T, FORMAT (16X," PLANT CAPACITY: FACTOR",T75,"9 8000") . _

9029l FORMAT'(16X,"

• MASS OF'WOTER IN STEAM GENERATORS (THOUSAND LBS)",T73,006900-

!. 1F8.*) On6910 9030 FORMAT (16X,"F1SSION PH00 ULT CARRY-OVER FRfCTIPN",T75,F6.4/16X, 006970 i 1" HALOGEN CARRY.0VER FMACT10N",T75,F6.4) On6930 9034 FORMAT (2X,"bt0WDOWN",10X,lPE9 2,14X,0FF5.3,24,2(F8.3.2X),

13(1PE9'.2 1X)) 066960. .

.9035'F0HMAT(7X," UNTREATED'9 LOWDOWN",1PE9.2e11X," 1 000 0 000 10.000 1.00E+00 1.00E+00 1 00E+00")

9037 FORMAT (72X,F B.2) 057000 9038 FORMAT (2X,"HEGENERANT OLb ",1PE9.2 14X,0PFD.3,2X,2(F8.3,2X), 007010 13(1PE9.2,1X)) 067020 9039 FORMAT (316,E21.14) ,

9040 FORMAT (16,E21.14) 9041 FORMATi16X," PRIMARY T9 SECONDARY LEAK. RATE (L95/ DAY)",T73, 1" 75.0560") .. .

19045'F0HMAT t/,"O LIQUID WASTE INPUTS") 007060 9046 FORMAT (/,"O GASE00) WASTE INPUTS")- .

067070 9048 FORMAT ("0",15X,"THERE IS NOT AN ON. SITE LAUNDTY") 0570A0 9051 FORMAT (16X," BLOWDOWN HATE (THOUSAND LBS/HY)",55X,F8.4) '067090

j. 9052 FORMAT (16X,"THERE IS 90NTINUOUS STRIPPING OF LULL LETDOWN FLOW")- 007100 9053 FORMAT (16X,"THERE IS NOT CONTIN 0US STRIRPINS OF FULL LETDWN FLOE"067110

~

1) 057120 59055 FORMATi36X,F8.4,35X,I&) 007140 9056 FORMAT (15X,*A4,A2,8X,l8 0) . .

067150 9064 FORMAT (16X," STEAM LEAh TO TURBINE BLOG . (Los/HN)",19X,F10.4) 9065 FORMAT (16X,4A4,6X,F3 0) 9066 F0dMAT(16X,4A4,4X,"PA0TICULATE RELEASE FR$CTIPN",6X,F10.4) +

.. 9067 F,0RMAT(16X,5A4,10X,F3,0,6X,F 3) 9068 FORMAT (16X,5A4,"IODINL RELEASE FRACTION",11X,[10.4/36X,"PARTICULAT

-1E RELEASE FRACTION",6A,F10.4) 9069 FORMAT (16X.5A4,10X,F3,0,6X,F3 0,14x,F8.2) 9070-F0HMAT(16X,5A4," RATE (CFM)",25X,F10.4/16X,bA4e" IODINE RELEASE FRACT I!ON",11X,F10.4/36X," PARTICULATE.RE LEASE FHACTdON",6X,F10.4) 9071 FORMAT (16X,5A4,10%,F3,0,6A,F3.0,19X,F3.0) .

9072 FORMAT (16X," FREQUENCY OF CNTMT RLOG MIGH VOL PURGE (TIMES /YR)", 057330 1 T 74, F 7'. 4 )

9073 FORMAT (16X,"THERE IS NOT.A CNTMT BLDG L0m V0h PURGE") 067350 9075 FORMAT (16X,"THERE IS 90NTINUOUS low VOL PURGE OF VOL. CONTROL TK")067360

-END_ 007370

• DECK EFFTAB 057380 SUBROUTINE EFFTAs 057300

-DIMENSION ISOTP(3,1001 DIMENSI ON HLACTR(7),NgME(J),CWCONC(600),0* CON 9(800),CMCONC(8003 p 3-32

l l.

( .. DIMENSI ON TURBDR(800),0wCON2(8001,EDCONC(#00),IOTHER(100)

COMM0h /FLUXN/ REGENT e D,1,S ( d D 01, lLI TE . I AC T ,I T OT COMMON /Our/NUCL(Boot

. COMMON /C00L/REACTR,P091,5bLDR',BLwDWN,FPEF,MEF,EDFLR,DFIEDoDFCSED, 1 DFED,DFIDw,DFCSDW,DtyW, EDA,0WA, CWA,DFCM.WFICM,DFCSCM, 2 DFCW,DFICW,DFCSCW,sD),FS,$DFD,0WFD CWFD,C6FD,15,TE.TD,TC TCM, 3 TST08C,TSTORD,TSTORBsDWFL2,DW2,0WF2,T2,ThTORd DFID2,DFCSD2,DF02, 4 PFLAUN,0WFLR

. COMMON /APC00L/RGWFR,0GIRG',DFCSRG,DFRG,TRG,TSTURR,0GFD 057630 COMMON /BDTES/RFNRT Oh7640 COMMON / CONC /PCONC (d .11,SCQN (800 ) ,RINV t 80 0)

COMMON /DET/ LAUNDRY (251,WL$UND(25; C

C H3COPW IS THE PWR TRITIUM PRIMARY COOLANT CONhENTRATION IN 007790 C UCI/GM 007800 C

H3PRPW=0.4* POW 1 M3COPw=1.0 067810 00 30 Jul,ITOT _

CwCONC(J)=0.0 On7920 EDCONC'(J)=0 0 067930 DwCONC(J)=0 0 007940 DwCON2(J)=0 0 0n7950 CMCONC(J)=0 0 007960 NZ=NUCL(J)/10060 On7970 IF(NZ.EO.36.OR.NZ.EQ.9*) VO TO 30 0n79A0 CWCONC(J)=PCONC(J)*CWe 0n7990 EDCONC(J)=PLONC(J)* EDA OnA000 4

DwCONC(J)=PCONC(J)eDwd 008010 UwCON2(J)=PCONC(J)eDW4 058020 CMCONC(J)=SCON(J). 068050 DFCVCS=50. _

IF(NZ.EQ.1)UFCVCS=1.0 008070 IF(NZ.EO.35.OR.N7.Eo.b33 DFCvCS=100.

IF(NZ.EO.37.0R.N7.Eo.55) DFCyCS=2. 068080 CwCONC(J)=CwCONC(J)/DLCVCS 00A090 30 CONTINUE 068100 C' 008110 C CALCULATE RADI0 ACTIVITY AfTER COLLECTION AT A CONSTANT RATE 008120

!. -C On8130 CALL. COLLECT (TC#86400,,CwCONC,ITOT)

! CALL COLLECT (TE*86400,,EDCONC,ITOT)

C ALL COLLECT (TD*8640 0, ,DWCONC',ITOT )

CALL COLLECT (T2*86400,,0wCON2,ITOT)

. CALL COLLECT (TCM*8640Y.,CMCONC,ITOT)

IF(REGENT.LE.O.0) GO TO 50 CALL STORAG(TRG'86400,,RINV,ITOT)

- 50 DO 100 !=1,ITOT 068210 NZ=NUCL(I)/10000 058220 TURBDR ( 1 ) = 19 91, e5. *SC,u,N ( 1 ) 008230 IF(NZ.EO.1) GO TO 100 068240 IF(NZ.EO.35.OR.NZ.EO.53) 90 TO 60 008250 IF(NZ.gQ.37.0R.N7.EO.45) ?O TU 70 008P60 C Oh8270 C CHEMICAL THLATHENT Fog OTUER CATIONS 0082A0 C' 008290 CwCONC(I)=CWCDHC(I)/DFCW Oh8300 EUCONC(I)=EDCONC(I)/0tE0 068310 uwCONC(I)=0wCONC(I)/Dt0w 008320 DhCONc(I)=DnCON2(I)/DtD2 058330 CMCONC(!)=CMCONC(I)+(4.0-oDTFR*(1 0-CHfD/ufCM)) 068340 i C-C TO THEAT FWH TURRINE pu!LDING FLOOR ORAINb TMkOUGH DIRTY WASTE 058350 3-33

C SYS, TEN, DELETE C FOR bOMNLNT ON CARDS BELo w , uNTIL NEXT'MLSSAGE 068360 C-RINV' II)=RINV (I)/DLRG 058370 TURBDR(I)=1991.*S.eSCUN(I)*FPEF 068380 C- TURBDSi!)=1991.*5.eSCUN(IleFPEF/DFDW 008390 60 TO iOO 068400

.C 008410 C CHEMICAL TREATMENT FO$ ANIONS 068420 C .

008430 60 CWCONCiI)=CwCONC(I)/DbICW 008440 EDCONC(I)=EDCONC(I)/Dt.IED 008450 DWCONC(I)=DWCONC(I)/DbIDW 068460 DWCONd(I)=DWCON2(I)/DPIB2 068470 CMCONC(I)=CMCONC(I)e(J.0-UDTFR*(1 0-CMFD/PFICM)) On8480 SINV (I)=RINV '(I)/Db!RG 008490 TURBDRills1991.e5.eSCUN(I)*HEF 008500 C TURBDR ( I ) = 1991. e5.*SCUN (I) *HEF/DF IDW 058510 GO TO i00 058520 C Og8530 C CHEMICAL TREATMENT F0p RB AND CS On8540 C 0nR590 T0 CWCONC(I)=CWCONC(I)/DtCSCW 09 8560 EDCONC (I ) =EDCONC (!) /DLCSED 008570 DWCONC(I)=UwCONC(I)/DLCSDu 068580 DWCON2(I)=DWCON2(I)/DLCSU2 068590 CHCONC (I) =CNCONC II) * ( 4.0-UDTFR e (1.0-CHFD/PFCS$M) ) On8600 RINV (I)=RINV (1)/DLCSRW 008610 TURBDR(I)=1991.*5.*SCMN(IleFPEF 00G620 C- TURdDR ( I ) = 1991. e5.eSCUN (1) eFPEF/DFCSD W Cn8630 100 CONTINUE 008640 C Og8690 C COMPUTE RADIOACTIVE DLCAY DURING PROCESSING AND SAMPLING On8660 C 008670 CALL STORAG(TSTORCe86?00.eCWCUNC,ITOT)

CALL STORAG(TSe86400.aEDCONC.1 TUT)

CALL STORAG(TSTORD*86400.tDWCONC,ITOT) I CALL STORAGtTSTOR2 86400.tDWCON2,ITUT)

' CALL STORAG(TSTORBe86900.,CMCONC,ITOT)

CALL STO9 AG (TSTORRe86?00.',RINV,ITOT)

CALL STORAG(21600.,TUgBDR.ITOT) ,

DO 130 I"1,ITOT On8750 ABLOW:6.0 068770 i IF(REGENT.LT.O.001) G9 TO 110 008780  !

ABLOW=RINV(I)*292.4eRuFD/MEGENT 110 ABLOW= A8 LOW + RLWDWNo 19y1.* CMCONC (I ) e (1 0-RF NR T )

130 CMCONCil)=ABLOW

-CWFR=SRLDR*CWFDe0.02822 058860 EDFLR=EDFLReEDFDec.02932 008870 DWFR=DWFLR*DWFDe0.02832 00AR80 DWFR2=DWFL2eDWF2eo.02932 00A890 TPLRPW3CWFR* CWA *EDFLR? EDA *DWFReDWA,0WFR2*DW2 008900 H3RLPW=TPLRfW*H3COPW 008910 IF(H3R LPW.GT.O.9eH3PRPW) 03RLPW=0.9eH3PRPs RN3RLP=H3RLPW/10 05A930

, INTRIM=RHJHLP 098940 t

IH3RLP=INTRIM*10 008950 TOTAL =6.0 11=ILITE+1ACT+i ,

DU 140 I=lelTUT 009050

! NZ=NUCL(I)/In000 l

-IF (NZ.EO. 36.OR.NZ.EO.?4) b0 TO 140 01SI= DIS (1)*1.62A3E13 3-34 u 080

1 * -

.-q'..

' v t ~

.- : r 1 - ; - , - y .,

3 y . .7;;l W , ..(.

. .y. f: ' ., . . . _ -

.g.w,-, y .. y ,. ; .. .

7.: . <.j,.: ,.f,$

, g ,m,.

"'ei 48y ' s * *N5-6..cfM

v. p' -.. 4#'r

$_.y. li g " i < ?'

y'l,. .p

. 1.j n 4 ; , ,

.y '

? ~, i,. ,. [ .

y ~.; . .

.-; '4 4' D.7

? ., c ;y. _:

':f [ * . ; '-

,,,-% o.

_. t v n :,,: ,g ;..

9 CwCONC !)=Disle(CWCONL I ) *Cw FH + EDCONC ( I ) ot.0FLM ) 069090 -

On9100  ;,'; .:

.{*

T Uw CONC ((Il = (uwCONC ( I) *Dw(f R

• DWCON2(I)*0=FH2)eDi@I CMCONC(I)=CHCONC(I)eDis! 009110 .g,"' g':/

J f TURBDRg!)=TURBDR(I)*DISI 009120 .-w- ,

O IF (NUCL (1) .EQ.10030 t 90 10 140 JcA y

.. TOTAL = TOTAL +CWCONC(.I)*DWCONC(I)*CHCONC(1)*TURSDR(I) 009140 M J4[-

h';Ni ' -S iy 140 CONTINUE 009150 AOI=0 16 7 O^,p'

• p AOR=(AOI+ TOTAL)/ TOTAL 009170 '.'

Y:

.- SCNORM=0.0

<. . E SAPRIM=0.0 069190 f 1, ' ; ;,.

'Sf SSEC=0'.o 069200 .f.N4 .i ..

"h SCWAST=0.0 009210 On9220 @p~ -:'%

~ -

S0 WAST =0.0 - .

cc SABLO*=0.0 059230  %.1, '

57 ST8=0.6 ST0TAL =0.0 06924-0 009750

.{

(-g. . f1 Ob92A0 .J.

' 7 6 PAPRIMan.0 PSEC=0'.0 069290 si, n 1. ;.

ci ;- PCWAST=0.0 009300 ,,'y. .(, . :

009310 P a PDWAST=0 0

,( PABLO*=0.0 069320 lld J. PT8=0.6 On9330 <.J > -

if PTOTAL =0.0 059340 .i . 'i. .'- A

- 'f. PNORMz6.0 069370 t., ih gf TLAUND=0.0 009380 ,,.0 Q .

3

'Y CTOTAL=0.0 009390 ." .~-

~' PRINT 0001, REACTR 009400

, PRINT 0002 ,

, ;;"4 .

PHINT 9610 On9430 ..

].,9 r0UNTR=1 DU 180 I=1,ITOT 069440 f ,er .

.&-O.r

? ';f IF (I .EO. ll) PRINT 901) .[I> D1 7

..y' NZ=NUCL(I)/10000 IF (NZ.EO.36.0R.N7.EO.?4 ) 90 TO 180

f. .'

./f IF(NZ.EO.1) GO TO 180 -)fi' M}W DISI= DIS (1)*1.6283E*14 APRIH=PCONC(I)eDISI Y.-g;.' , y. :

..C ASEC=SCON(1)eDISI N -i ?

M CWASTE=CwCONC(I) -

[W' N .

5

}k (

Ow ASTE=DwCONC (I)

AOL0w=CMCONC(I)  ? ?. :

• e Tb=TUNRDR(1) f5[ ,'I .:

'i4 .- TOTAL =CWAbTE+DWASTE*APLOW*TB Oly9600 Pl - ..'[

. h i '

/: TOTALN=TOTALeAOR On9610 NUCLI=NUCL(I) '*

f.'E. h c.; y XLAUN0=0.0 009630 i

. - IF (I.GT.155.AND.I.LT,190) GO TO 152 'd ,, ' -

J IF (I.EO.225) GO TO 142 7 ' .. t "

i+o 00 150 L=1,25 U-IF (L AUNDRY (L) .EO.NUCL1,) XLAUND=WLAUND(L)*PFLAuN 150 CONTIhuE

.'.['- a3 O

1 Q ...

1.t 152 CONTINUE .

.Y E

h 155 TOTALG=TOTALN*XLAUND IF(TOTALG.LT.0.00001) GO 10 160 ISUB=2 0n9710 U

1<,, }. 1 4 2

?, 009T20 i .7 "'

-4~- ^

rg IF (TOTALG.uT.I.)ISUdf1 p* 01V=10'.**(INT (ALOG10(!OTALG))?ISUH) 069730 {, . ,. , :

+-. TOTALG=AINT(TOTALG/DIV*0.ci)*DIV 009740 -#4 J '

sL- 160 IF (NUCL ( 1 ) .LO.10 0 3 0 ) TOTALN= TOTAL t.J,",f.5 '

T- IF (NZ.EQ.1) GO TO 162 ~j

• - SAPRIMasAPRIH.APRIM .k.j ' ?*- t r . S$EC=SSEC+ASEC Oh9810 C.[3 r~

str. .

. . .e

. 3 5 '# -- ; ,

.;n, y,

$,s

.)j.'y ' %.;:' q.;p.._

J ; y %_,.y; : 3. :;:,1.y..;.q: , y .;t v ; p- .( ,. y 9 .;;.-

3. .g.y ;y.;.y;;.yy g .# -n:P: -

SABLO*=SABLOW+ABL0w 059820 SC' WAST =SCwAST+CWASTE On9830 SDWAST=SDwAST+DWASTE 069840 STB = STB +Td 009P50 ST0TAL=ST0TAL+ TOTAL 009860 SCNORN=SCNORM+TOTALN 069890 TLAUNP=TLAUND+XLAUND 009900 CTOTAL=CTOTAL+T0TALG Ob9910 162 IF(TOTALG.LT.0.00001) GQ TO ido 168 IF(MOP (KOUNTR,50).NE.9) 09 TO ITO PRINT 9000, REACTR 069940 RRINT 9002 ITO CALL NOAH (NUCL'(I),NAME)

THALF=8.0225E-6/DISTI!

PRINT 9003,NAME,TMALF APRIM,ASEC,CWASTE,0WASTb,ABLOW, ITB, TOTAL,TOTALN,XLAUND,TOTALG ,

KUUNTB=KOUNTR+1 010030 IF(NZ.EO.1) GO TO 180 PAPRIN=PAPRIM+APRIM OI0050 PSEC=PSEC+ASEC OIO060 PCWAST=PCwAST+CWASTE Oio0TO PDWAST=PDwAST+DWASTE OT00nD PABL0n=PAULOW+ABL0w 0i0000 PTB=PTB+Td OI0100 PTOTAL=PTUTAL+ TOTAL OIO110 PNORH=PNORH+TOTALN 010140 180 CONTINUE OT0150 PAPRIP=SAPRIM-PAPRIM 0i0160 PSEC=SSEC-PSEC 0101T0 PCWAST=SCWAST-PCWAST OI0180

PUwAST=SUwAbT-PDWAST 0i0190 PAHL0n=SAdLOW-PABLOW OIO200 PT8= STB PTB OiO210 PTOTAL=STUTAL-PTOTAL 010220 PNORM=SCNORM-PNORM 010250 ISUBCa2 0i0260 IF (CTOTAL.GT.I.11SUBy=1 OIO2TO DIV=10.**(INT (ALOG10(CTOTAL))?ISUBC) OIO280 CTOTAL=AINT(CTOTAL/DIy+0.S)*DIV 0i0290 i

IF PNORM.LT.O.00001t 90 To 190 ,

DIV=10.**(INT (ALOG1V(Ph0RM))-2) 010310 PNORMT=AINT(PNORM/DIV?e.5)*DIV OIO320 60 TO 200 190 PNORHT=PNORM 200 PRINT gnO4, PAPRIM,PSECtPCWAST,POWAST,PABLO.w.,PTB,PTOTAL PNORM, ,

1 PNORMT 010380 PRINT 9005, StPRIM,SSEC,SCWAST,SDWAST,SARLD=, STB.ST0TAL,SCNORM, 010340 1 TLAUND,hTOT6L 010400 PRINT 9n12, Ih3RLP 010410 PRINT 9013 RLTURN DIO420 9000 FORMAT (IM1,20X,TA4," hIoulD EFFLUENTS (CONTINYED)") 0{0480 9001 FORMAT (1H1,20X,T A4," hlQUID EFFLUENTS") 010490 9002 FORMAT (IHO,55X," ANNUAL RELEASES TO DIS C HARGE LANAL"/20X,"C00LANT CoiO500 10NCENTRATIONS",5T(" "1," ADJUSTED DETENGEN! TOTAL "/" NilCLID0io5io 2E HALF. LIFE PRIMARY SECONDARY BOMON MS HISC. WASTES SECON0iO520 30ARY Ti1Ru bLOG TOTtL LWb TOTAL WASTLS "/10X. DiO530 4"(DAYS) "2("(MICR0 E!/NL)"),1X,4("(CURILS) ") , " ( CilR I ES ) ", OiO540 5" (CI/YR) (CI/YR1 (CI/YR)") OIO550 9003 FORMAT (1X,A2.I3,A1,2X 1PLV.2.2(2X,E9.2,24),0P,T(1X,F9.5,1X),F10,5) 9004 FORMATjlX,"ALL'OTHERSH,9X,1PE9.2,4x,E9.2,0P,2A,6(1X,F9.5,1X).3X, 1 "0 00000",1X,F10.5) 3-36

9005 F0HMAT(n TOTALn/,n (EACEPT TRITIUM) n ,1 P E 9,5,4 X , t 9. 2,0 p *,2 X ,

1 7(1X,F9.5,1X),F10.5) 9010 FORMATgn C0HROSION AND ACTIVATION PRODUCThn) OIO740 9011 FORMATfu0 FISSION PRODUCThu) 010750 9012 FORMAT (1HO,1X,nTRITIUM ALLEASEu,12X,13," CURIES PER YEARu) 010760 9013 FORMAT l(1HO,1XenNOTES #00000 INDICATES THAT THE VALUE IS LESS THAN 11 0E-5.n) 9014 FORMAT (3X,10(2X,A2,13sA1)/3X,10(2X A2,13eA1))

END OIO770

• DECK BLKDAT 0i0780 SLOCK DATA BLKDAT OIO790 C

C PWCONC CONTAINS PRIMARY COOLANT CONCENTRATION) FOR PWRS. SCUTV C AND SCOT CONTAIN SECONDARY COOLANT CONCENTRATA0NS FOR PLANTS C WITH UATUbE STEAM GENbRATORS AND F OR PLANTS W1.TH ONCE-THROUGH C STEAM GENERATORS, RESPECTIVELY.

C COMMON /CONP/PWCONC(800),SCUTV(800),5 COT (800)

COMMON / DET/ LAUNDRY (251,WLAUND(25)

DATA PwCONC/36*0,4.7E62,67*0,3.1E-3,4*0,1 6E-4,5*0,1.2E-3,3*0.3.0E 1-4,0. 0,4. 6E-3,2'0,5. 3 E-4,17

• 0,5 1 E 4,10 2

• 0,2 s 3E-3,6 4

• 0,2. 2 E-3,68

• 0 2,1.6E-2,18*0,1.9E-1,4*0,1.4E 4,3*0,1,2E 5,5*0,9.6E-4,4.6E-4,5.2E-6 3,9*0,C.2E-3,11*0,3.9E'4,0 0,2 8E 4.15*0,6.4E-J,4.7E-3,16*0,7.5E 3, 415*0,9'.nE-2,0,0,20*0,1.3E 3,104*0,1.9 E 4,2,4E-2,12*0,1.5E-3,7.7E-3 5,4.5E-2,5*0,1.7E-3,2.1E-l',4*0,1.4E 1,$*0,3.4E-1,2*0,7.1E-8,2*0,2.6 6E-1,880,8.7E-4,3*0,9.?E 3,13*0,1.3E-2,2.5E 2,Ve0,1.5E-4,12*0,2.8E-73,3*0,3.9E-3,9280/

DATA SCUTV/36*0,1.5E-9,67*0,1 3E 7,4*0,6.4E-0,5*0,4.9E-8,3*0,1.2t-16,0.0 1.9E-7,2*0,2.2E68,17*0,2.1E-8,102*0,8.75-8,64*0 8 4E-8,68*u, 27.5E-0,18*0,5.3E-7,4*0,5.7E-9,3*0,4.99-10,5*U,2.8E-8,3.2E-9,2.1E-1 30,9*0,1.2E-7.11*0,1.6L 8,n.0,1.1E-8,15*0,2.5E-7,1,1L-7,16*0,3.1E-7 4,15*0,3.7E-6,0.0.20*0,5 30-8,104*0,7,8E-9,2.29 7,12*0,5.4L 8,2.9E-SS,1 8E-6,5*0,6.6E-8,3,1E 0,4*0,4.8E-6,5*0,2.49-6,2*0,3.3E-7,2*0,6 66E 6,$*0,4.0E-8,3*0,4,4E 7,13*0,5 2E 7,9.3E-7,5*0,6.1E-9,1200,1.0L 7 7,3*0.1.oE-7,92'0/

DATA SCOT /36*0,1.'0E47,67*0,6.9E-9,4*0,3.6L 9,3*0,2.7E-9,3*0,6,7E-1 10,0.0,~.0E-6,200,1.2E69,17*0,1.1E-9,102*0,5 1 6t-9,64*0,4.9E-9,68*0, 21.8E-9 3 18e ,6.0E-7,4*0,3.1E-10,3*0,2.(E-11,5*u.,2.1E-9,9.7E-10,1.2E 3-11,9'n ,9. 3E 9,11

• 0,8,7 E-10,0 0,6. 2E 10,15

• 0,1. 4E-8,1. 0 E-0,16

• 0,1.

47E-8,15*0,2.0E-7,0.0,$0*0i2.9E-9,104 *0,4.2E-10,5.1E-8,12#0,3.3E-9, 51 5E-8,5.2E-8,5*0,3.8E-9,2.4E-7,4*0,1 6E 7,5*u,3.8E-7,2*063.0E-8,2 6* 0,3. 0 E 7,8

• 0,3.6E 9, f

• 0," 9E-8,13

• 0,2.9E-8,5 6E-8,5

• 0,3.3 E-10.12 e 70,6 2E-9,3*0,8.7E-9,94*0/

C C LAUNc9y ARE THE RADIO 1.SOTUPES IN THE DETESGENI WASTES.

C WLAUNO ARE THE CORRESPONuiNG CONCENTRATIONS.

C DATA LAUNDRY /150320,240510,25U540,260550,2605?O,27 0 580,270600,2806 130,380890,380900 390910,400950,410950,420990,441030,441060,471101, 2511240,531310,551340,951360,551370.561400,581?10,581440/

DATA WLAUND/1.8E-4,4.(E-3,3.8 E 3,7.2E-3,2,2E-f,7.9E-3,1.4E-2,1.7E-13,8 8ti5,1.3E-5,8.4E-?,1.1E-3,1.9E 3,6.0E-5,2 9E 4,8.9E-3,1 2E-3.4 2.3E 4,i 6E-3,1,1E-2,3,7E 4,1.bE-2,9 1E 4,2.3E-4,3.9E-3/

END OI1230

• DECK SULVE Oil 240 SUBROUTINE SOLVE 0i1250 =

COMMON /Eu/ATEMP(800),XNEt(10,0003,H(800),u(800)

COMMON /FLUXN/ REGENT,D1,5(800),1 LITE 1ACT,IIOT

, 00 10 Is1.ITOT OI1350 D(I)=-DIS (I) 011360 10 XTEMP(I)=0.u 051370 UELT= REGENT *86400

CALL DECAY (1.UELT,ITOI) Di1300 3-37

l CALL. TERM (DELT 1,ITOT)

CALL EQUIL(1,ITOT) Di1410 l DO 30 Is!,ITOT~ Oi1420 )

30 XTEMP(IlsANEW(1,1) 011430 RETURh 011440 END Di1450

• DECK TERM OI1460

' SUBROUTINE TERM (T,M,ITOT)

C TERM ADDS ONE TERM To EACU ELEMENT OF THE SOLUTION VECTOR 051490 C CSUM(J) IS IHE CURRENT APPROXIMATION To XNEW(B J) 011500 C CIM0(J) IS THE VECTOR CONTAINING THE LAST TERM ADDED TO EACH OI1510 C ~ELEMEhT OF CSUM(J) Oil 520 C CIMN(J) IS THE VECTOR CONTAINING 1/ TON T'.NES THE NEW TERM TO BE Dil530 C ADDED TO CSUM(J) OI1540 C CIMN(J) IS GENERATED LROM CIM0(J) BY A RECURSION RELATIONI Di1556

'C CI MN(J)s SUM OVEd L QF (AP(JilleCIM0(L)) 011560

.C AP(I,J) IS THE REDUCEu TRANSITION MATMIX FOR IHE LONG-LIVED- OI1570 C NUCLIDES OI1580 C OI1590 LOGICALLONG , Oi1600 DIMENSION AP(2$00),CIMB(800),CIM0(800),CINN(800),CSUM(800)

DIMENSION QU8(50), LOC ((2500),NONP(800)

COMMON / SERIES / XP(8001,XP6R(8001,LONG(800) Oi1690 COMMON /CONST/MMN ERR,MZERQ COMMON /Eu/ATEMP(800),XNEw(10,800)',8(800)eD(8003 COMMON / MATRIX /A(2500)tLOC(2500),NON0(800),KD(600) OI1730 LEVEL 2,A, LOC NONO,KD 050540

, COMMON / TEHMu/DD(1001,0XP(100),0UEUE(50),NWU(50),NQUEUE(50),NQ(80010il750 NUL=0 011760 NN=0 011770

.C FIRST CONSTHUCT REDUCED TMANSITION MATRIX FOR LONG-LIVE 0 ISOTOPES 0i1780

, DO 220 L=1,1 TOT OI1790 IF ( .NOT.LUN6 (L) ) GO Tu 210 Ojl800

.NUM=NONn(L) 011810 IF(M.97.MMN.0R.M.EQ.MlERO) NUHzKD(L) OI1820 i

.CIMu(L):HtL) Di1830  !

IF(NUM.LE.NUL) GO TO 410 ~

OI1840 ,

NS=NN*i 011850 NJNUL 011860 NL=NUM.NUL 011870 t DO 200 N1=1,NL 0{1880 I .N=N+1 011890 J: LOC (N) OI1900 .

DJa-D(J) OI1910- j

[ C 051920 1 C THIS IS A TEST TO SE5 IF ONE OF THE ASSYnPTOIIC SOLUTIONS APPLIES 011930 C Oilo40 1 IF(.NOT.LONGtJ)) GO Tu, 10 Oi1950 I NN=NN+1 OI1960

-AP(NN)sA(N) 0i1970 l LOCP(NN)=J 011980 ,

60 TO 200 011990 '

1

,C 0i2000

, C. GOING RACK UP THE CHA,1,N Tu FIND A PARENT WHICQ IS NOT IN 0 2030 C Euu! LIBRIUM 0 2020 C

OI2030 10 "NSAVE=6 012040 uuE=A(N)/DJ 012050 DRB=1.6 Oi2060

. CIM8(L)=CIMH(L)+0UE*PlJ) Di2070 40(L)=6 3-38 oi2080 4

NQ(J)=L. .

OI2090 20 NUX=NONn(J) 012100 IF(M.GT.MMN.0R.M.EO.M4ERO) NUX=KD(J) Oi2110 NUF=0 . -0I2120 IF(J.9T.1) NUF=NON0(9-1) DI2130 NX=NU5 NUF OI2140 IF(NX.LT.13 GO TO 199 Di2150 K=NUF DI2160 00 180 KK=leNX OI2170 K=K .1 0{21R0 J 1 =L OC '( K ) 012190 DJa-D(J1). Di2200 KP=J .

OI2210 30 ~ IF(J1.EQ.NQ(KP3) GO TO 140 0[2220 KPgNQ(KP) 012230 IF(KP.NE.0) GO TO 30 OI2246 8KDJQs0UE*A(K)/DJ- DI2250 IFt.NOT.LON6(Ji)) GO TO 160 OI2260.

TRM=1.6-AP(J1) 0{2270 IF(TRM.LT.1.0E-6) GO TO 120 012280 NQ(J1)=J 0I2290 lal 0I2300 KP=J1 OI2310

..40 DD(I)=-D(KP) OI2320 DXP(I)=XP(KF) OI2330 KPaNQ(KP) Oi2340 IF(KP.EO.0 GO To 50 Oi2350

!=1+1 OI2360

~ 1F (I .LE.10 0) GO TO 40 . OI2370 C lIF uuEUE OF SHORT-LIVED huCLIOES EXCEEDS 100 1.S0 TOPES. . TERMINATE OI2380

'C- CHAIN AND WRITE HESSA9E OI2390 PRINT 9000, M.L,J1sJoAhDJQ OI2400 9000 FORMAT (n1TUD LONG A QUEUE HAS BEEN FORMED IN lERMn,415.E12.5) Di2410 GU TO 190 012420 50 U AT M4 0 '.E 0 012430-IM=I-1 Oi2440 00 110 I=2,IM Oi2450 DL=UD(I) Oi2460~

XPL=DXP(!) OI2470 SATE =0'.E0 OI24RO 11=1-1 OI2490 C 0 R VONDY FORM OF BATEMAN EQUATIONS - ORNL TU-361 0I2500 00 100 KB=1,Il CI2510 XPJ:DAP(K8) Oi2520 IF(XPL;XPJ.LT. ERR) GO TO 100 OI2530

, OK=DD(KB) Oi2540 PR00=(DL/DK-1.0) 012550 DKR=PM0D Oi2560

-IFt ASS (PROD).GT.1.E-9) 00 TO 60 OI2570 C- lUSE THIS FORM FOR Two NEARLY EQUAL HALF-LIVES UI2580-

' PROD =T*DK*XPJ'(1.0 0.h* (DL-DJ) *T) OI2590 60 TO 70 OI2600

60. RHOD =(XPJ-XPL)/ PROD OI2610

' PRO 1=XPJ/UKH Oj2620 70 PI=1.0 012630 Sl=2./(Dn*T) OI2640 00 90 JK=1,11 Di2650 IF(JK.EO.Kd) GO TO 90 OI2660 S=1 0-DK/UD(JK) OI2670 IF( ARS(S).GT.1.E-4t 90 IU 80 OI26A0 1F(ABS (DKH).GT.1.0E-41 PROD =PRol 0i2690 .

h=51 Oi2700 80' Pl=Ples Oi2710 3-39

. .___ _ _ _ I

l 1

IF(ABSgpl).GT.1.E25p 00 10 100 412720 90 CONTINUE 012730 BATE = BATE + PROD /PI OI2740 100 CONTINUE 012750 C IF SUMMATION IS NEGATIVE, SET EQUAL TO ZENO AND PRINT MESSAGE 012760

.IF(BAT E.LT.O.EO)PRINTY001,L,IM,8 ATE,8ATM 012770 9001 FOMMAT(n1HAIE IS NEGATIVE IN TERM. THERE eRE MORE THAN TWO SMopT-L012780

-11VED hUCLIDES IN A CU$IN WITH NEARLY EQUAL DI6 GONAL ELEMENTS"/ 012790 2n L,IM.8 ATE,BATM = ",?IS,1P2E12 5) Di2800 IF (R ATE.LT.O.EO) R ATE =9.E0 012810 BATMsWATM+ BATE OI2820' 110 CONTINUE .

Di2830 DRA=AKDJQ*DJ'(TRM-BATN)/T*M 012840 GO TO 130 012850 120 DRA=AKDJQ'AMAXi(DRB,0,0)*DJ Di2860 130 IF(NS.GT.NN) GO TO 159 OI2876 DO 140 LJmNS,NN Oi2880 IF(LOCp(LJ).NE.J1) GH TO 140 OI2890 AP(LJ)=AP(LJ)+DRA OI2900 GO TO i80 0I2910 140 . CONTINUE Oi2920 150 NN=NN+i 012930 AP(NN)=DRA 012940 LOCP (hN)=J1 OI2950 Ou TO iB0 0'12960 160 IF(AKPJ0.LE.1.0E-06) GO TO ioO 012970 l IF(NSAVE.GE.50) GO TO 100 ci2980 '

170 NSAVE=NSAVE+1 OI2990 HQUEUE(NSAVE)=J1 OI3000 uuEUE(NSAVL)=AKDJQ OI3010 NQu(NbAVd)=9 OI3020 QUd(NSAVE)=DRB-1./(DJfT) OI3030 180 CONTINUE OI3040

.190 IF (NSAVE.LE.0) GO TO 200 Oi3050 J:NQUEUE(NSAVE) OI3060 QUE= QUEUE (NSAVE) Oi3070 NQ(J)=NQULNSAVE) DI3080 DRBsQUB(NSAVE) Oi3090 ,

l CIMB(L)=CIMb(L)+QUE*8(J)*eMAX1(DH8,0.0) Di3100 I NSAVE4NSAVE.1 013110 60 TO Po 013120 L 200 CONTINUE 013130 210 NUL=NON0(L) .053140 l l NONP(L)=NN 013150 l l

220 CONTINUE OI3160 C FIND hORM OF MATRIX AND ESTIMATE ERROR AS DESCRIBED IN LAPIDUS DI3170 C AND LUUS, OPTIMAL CONTROL OF ENGINEERING HROGESSES BLAISDELL 1967 OI3180 C FIN 7 THE MINIMUM OF TUE NAXIMUM' ROW SUM AND TUE MAXIMUM COLUMN SUH0i3190 ASUM =0 0 013200 ASUMJ=0.0 013210 NUL=1 OI3220 i DO 250 I:1,ITOT 013230 IF(.NOT.LONb(I)) GO TR 250 013240 DI=-D(I)*T Oi3250 8J=DI- 053260 NUM=NONp(I) 013270 IF (NUL .GT . HUM ) GO TO f40 Oi3280 00 230 N=NUL NUM OI3290 230' AJmAJ+AP!N) OI3300 240 Al=DI+n! Oj3310 IF ( AI .GT. ASUM ) ASUM =AI 013320 IF ( AJ.GT. ASUMJ) ASUMJ=AJ Oi3330 t 3-40

250 NUL=NONP (14 +1 OI3340 IF(AsyyJ.LT.ASUM) -ASUM=AbuMJ 013350 C USE ASUM TU DECIDE MOW MAHY TERMS ARE REQUIREY AND ESTIMATE ERROH OI3340 NLARGE=3.5+ASUM +5. Oj3370 XLARGE=NLARbE 013380 EHR1=EXP(ASUM )*(ASUM +2.T1828/XLAR GE)**NLARbb/SQRT(6.2832*XLARGE30i3390 IF (ERRi .GT.1.E-3 ) PRANT-9002, ERR 1,ASUM ,NLARGE Di3400 9002 FORMAT'("0 MAXIMUM ERRO$ GT 0 001, anF10.6.". TOACE = uF10.4 0I3410 1

" NLARGE = nI6) OI3420 C NEXT 6ENERATE MATRIX EXPONENTIAL SOLUTION OI3430 DO 260 !=1,ITOT OI3440 CSUM(I)=XTEMP(I) OI3450 CIMN(!)=xTEMp(I) 013460 260 CONTINUE 013470 EkR3=0'.001* ERR OI3480 00 310 NT=1,NLARGE Oi3496 UO 270 !=1,1 TOT 013500 CIM0(I)=CIMN(I) Di3510 270 CONTIhuE Oi3520 TON =T/NT Oi3530 NUL=1 OI3540 00 300 I:1,ITOT OI3550 IF ( .NOT.LONG (1) ) GO TO 300 Oi3560 NUM=NONP(I) Oi3570

'CIMNIs6.0 OI3580 IF(NT.EO.1) CIMNI=CIM9 (I) OI3590

' IF (NUL'.GT . HUM) GO TO 490 013600 DO 280 N=NUL,NUM OI3610 J=LOCP(N) 013620 280 C IMNI =C I MN I + AP (N) *C IMV. (J ) 013630 290 CINNI=CIMNI+D(I)*CIM0(!) Oi3640 CIMNI=y0N*CIMNI Oi3650 IF ( ABS (C1dNI) .LT. ERR 31CIMNI:0.E0 Oi3660 CIMN(I)=CIMNI 0{3670 CSUM (I) =CSUM (I) +CINNI 013680 300 NUL=NONP(I)+1 Oi3690 310 CONTINOE 0i3700 00 320 I=1,ITOT Di3710 IF(CSUM(I).LT. ERR) CPUM(I)=0 0 OI3720 IF(LONG(I)) XNEW(M,I)1CSUH(I) 013730 320 CONTINUE 013740 NLTURN 0i3750 END OI3760

+0E'CK OECAY Oi3770 SUBROUTINE DECAY (M,T,AT0T) Oi37A0 C DECAY TREATS SMORT-LIVED ISOTOPES AT BEGIhNINy OF CMAINS USING OI3790 C DATEMAN EQUATIONS Di3800 LOGICALLONG 013810 COMMON /SEHIES/ XP(8001,XP$R(800),LohGt800) 013860 COMMON /CONST/MMN, ERR,MZERU COMMON /Eu/ AT EMP (8 0 0 ) ,KNE = (10, d O O ) , n (8 0 0) ,P ( 80 01 CUMMON/PATRIX/A(2500): LOC (2500),NON0(800),KD4400) OI3900 LEVEL 2,A, LOC,NONO,KD 050560 COMM0h/TEHM0/DD(100),9XP(1003',0UEUE(50),N4U(50),NQUEUE(50),NQ(800)0i3910 AXN=-ALOGt0.601) .

DU 10 I=1,110T 013970 XPAR(I)=0.0 Oj3930 LONG(Ig=. FALSE. Di3940 API =0 0 013950 DT=D(I)*T OI3960 IF(DT.LT.-50.) GO TO 10 0{3970 IF ( ABS (DT) .LE. AXN) LUNG (I)=.TRUE. 013980 XPI=EXp(OT) OI3990 3-41

10 XP(I)=XPI. OI4000 NUL=1 014010 90-160 L=1,ITUT '0I4020 X TE M = 0'. 0 OI4030 DL=-D(L) Oi4040 NUM=N9N0(L) .

DI4050 IF(M.9T.MMN.OR.M.EQ.M4ERO) NUM=KD(L) 0i4060 IF ( NUN'.LT .NUL ) GO TO &50 Oi4070 DO 140 NaNUL, HUM OI4080 J= LOC (N) 0{4090 DJa-D(J) 014100-IF(LONG(J)) GO TO i40 Di4110 C USE THIS FORM FOR Two NEA*LY EQUAL H4LF-LIVES 014120 IF(ABS (DL/DJ-1.0).LE.1.0E-5) XTEM=XTEM+XTEMP(J)eA(N)eXP(J)*T 014130 IF(ABS (DL/DJ-1.0).GT.).0E-5) 014140 1 XTEM=XTEM+XTEMP(J)*A(N)*(XP(J)*AP(L))/(DL-DJ) 014150 QUE=A(N)/UJ 014160 NQ(L)=6 OI4170-NQ(J)=( DI4180 NSAVE=0 OI4190 20 NUX=NONn(J) OI4200 IF(M.GT.MHN.OR.M.EQ.M4ERO) NUX=KD(J) OI4210 NUF=1 014220 IF(J.GT.1) NUF=NON0(9-1)*1 OI4230 IF(NUF.GT.NUX) GO TO 130 0I4240 00 120 K=NUF,NUX Oi4250 J1= LOC (K) 0i4260 IF(LONGtJ1)) GO TO 129 OI4270 KP=J OI4240 30 lIF(J1.EO.Nu(KP)) GO TO 120 014200 hP=NQ(KP) 014300 IF(KP.NE.0) 'GO TO 30 Oi4310 DJ==D(J1) OI4320 AKDJQsA(K)/DJo0UE 014330 IF ( AKDJQ.LE.1 0E-06) GO To i20 014340 NQ(J1)=J 0I4350 I=1 054360 KP=J1 014370 40- DD(I)= D(KP) 0i4380 DXP(I):XP(KP) 0i4390 KP=NQ (KP) OI4400 IF(KP.EO.0) GO TO 50 0[4410 181+1 014420 IF(I.LE.100) GO TO 40 014430 PRINT 9000, M.L,J1sJ,AKDJQ 0I4440 9000 FORMAT}"1",4IS,E12.5) 014450 GO TO 130 014460 50 W A T E = 0'. E 0 OI4470

'I1=I-1 0{4480 XPL=XPiL) 014400 C D R VONDY FORM OF BATEMAN EQUATIONS - ORNL-TM-361 OI4500 00 100 Kd=1,Il 014510 XPJ=D3P(K8) 014520

, IF(XPL;XPJ.LT. ERR) GO TO 100 0[4530 DA=DD(KB) 014540 PHOD=(OL/un-1,0) 034550 UKR= PROD Oi4560 IF( ABS (PROD).GT.I.E-?) GUTO' 014570 PROD =T*DK+XPJo(1 0-0.5e(DL-DJ), 0145A0 GO TO 70 Oi4590 60 PH00=(XPJ-APL)/P400 014600 PRO 1=XPJ/UKM 14610 342

it s

170- PI=1.0 -. -

-014620 Sl=2./(DK*T)- 014630.

DO 90.JK=1,Il Oi4640' IF(JK.EO.Ku) GO TO 90. .0I4650 A S=1 0-DK/DD(JK) .

Oi4660 IF( ASS (S).GT.1.E-47 90 TU 80 .

0i4670' C. .USE THIS F0HN FOR TWO NEAMLY EQUAL MALF-LIVES DI46R0 '

IF(ABS (DKR).GT.1 0E44& P80D PR01 0i4690 S=Sl~ Oi4700

- 80 . :PIsPlos .

014710

.IF ( A8p (PI) .GT.1.E25) -GO TO 100' 014720-

'0i4730.

~

901 -CONTINUE DATE 8 ATE + PROD /PI~ OI4740 100 CONTINUE .

'0i4750

=IF(BATE.LT.O.E0)PRINTY001,L,I6 BATE,XTEM,XTEMP(J11 AKDJG Oi4760 190011 FORMATin_L,1, BATE,XTENeXTEMP(J1),AKDJG = ",215,1P4E12.5)- 0{4770' IF(BATE.LT.u.EO) RATE =0.E0 '014780 XTEM=3 TEM +XTEMP(J13eA5090*8 ATE Oi4790 '.

IF(NSAVE.GE.50) GO TO 120 0I4800 110 NSAVE=NSAVE+1 . 0I4810 NQUEuE(NSAVE)=J1 054A20 QUEUE (NS AVE) = AKDJQ :' 0[4810 1NQU(N3AVEl=J 014840 "120- CONTINUE. .

0i4850 130 IF(NSAVE.LE.0) G0~TO 140 OI4860-J.NQUEUE(NSAVEf- Di4870-QUE= QUEUE (NhAVE) 0148a0 NQ (J) =NOU (NS AVE). :Di4890 HSAVE=NSAVE-IL .0i4900 GO-T0"2n' Di4910 140 CONTINUE 0i4920 IF(LONG(L)) XPAR(L)=XTEM/AP(L) 014930' .;

L150 NUL=NON0(L)+1 Oi4940 IF ( .N07.LONb (L) ) XNEWIM,L):XTEM+XTEMP(L)*AP(L) OI4950 160= CONTINUE- '0i4960 00 170 !=leITOT .. Oi4970 IF(LONG(I)) XTEMP('It:XTENHgI),XPAR(I) '

0{498D IFt.NQT.LONG(I)) XTEMP(I)=0 0 014990

'170 CONTINut- OIS000 HETURN: DIS 010

-END 0iS020

• DECK'EQUIL 0I5030.

tSU8ROU' TINE EQUIL(9,ITuT)' OISO40-C 0iS050 C- . EQUIL PUTS SHORT-LIVEV O AUGHTERS IN. EQUILIBRIUM WITH PAREMTS - DIS 060

-C EQUIL USES GAUSS-SEIVEL ITERATION TO GENLRATg STEADY STATE oj5070

. C .. CONCEhTRATIONS- 0}5080' 015090 C

LOGICALLONG Oi5100

+ LCOMMON/EQ/XTEMP(800),XNEh(10,800),8(800),9(800)

COMMON / MATRIX /A(2500)sLOCi2500),NON0(800),KD4?00) 0{5140-

-LEVEL 2,A, LUC,NON0,KD On0580 COMMON /CONST/MMN, ERR,NZERO COMMON / SERIES / XP(8001,XPPR (800),LONG (800) OiS160 QXN=0 001.

00 10-I 1,ITOT '

Oi5173

XPAR(I)=0.0 OI5180 sIF(.NOT.LONG(I)) GO Tu 10 OI5190

'XTEMP(I)=ATEMP(I)*xP(A) Oi5200 XP AH t ! ) = A'4 A A1 (XNEW (M,13 -XTEMP (I) ,0,0) 0i5210 oj5220

'10 : : CONTINUE

- ITER =1 015230-20- 'h=0 OiS240 3-43

-HIG=0 6 .

015250 D0.60 !=1,ITOT' 015260 NUM=NON0(I)-N OiS270 dis-D(I) U[5280 IF(LONG(I)) GO TO 50 015290 FNW=B(!) OI5300 IF(M.GT.MMN.0R.M.EQ.M4ERO) NUM=KD(I).N OI5310 I F ( NUN'. EG . 0 ) GO TO 31 OI5320' D0 30 Kal,NUM Oi5330 N=N+1 OiS340

. J= LOC (N) - 0i5350 DJa-D(J) 015360 XJmXPAR(J) 015370

.IF(LONGtJ)) XJ=XJ+XTEMP(J)/(1,0-DJ /DI) DI5380

'XNWsXNW+A(N)eXJ Oi5390

'30 CONTINUE Oi540 0' 31 XNW=XNW/DI Di5410 I F ( XNW'.L T .1. 0E-5 0 ) Gu TO 40 OiS420 ARG*AS3((XNh-XPAR(I))/Xhh) OiS430 IF(ARG'GT. BIG) RIG =AUG 0i5440

.40 .APAR(ItaXNW OI5450

50. N=NON0(I)- 055460 60 CONTINHE 015470 I F ( B I G'.L T . uXN ) G010 70 0i5480

. ITER = ITER +1 015490 IF(ITER.LT.100) GO TH 20 015500

. PHINT 9000 0i5510 STOP ' .

015520 70 00 do I 1,ITOT 015530 IF(.NOT. LUNG (I)) XNE w (M ,1) =XNEW (M ,I) + AP AH (I) 055540 80 CONTINUE OiS550 NETURh DI5560 9000 FORMAT (* GAUSS SEIDEL ITEMATION DID NOT CUNVENGE IN EQUILH) OI5570 END Oi5580

• DECK NUDATA dis 590 SUBROUTINE NUDATA(NLI@El OiS600 C NUDATA VER510N TO HANULE THREE TYPES OF NUCLEAR DATA LIBRARIES Oi5610 C. HAS POINTER, NLI8E = 1 fur HTGR- 0{5620 C = 2 FOR LIGHT WATER REACTHR 015630 C = 3 FUR LMF8R 0[5640 C = 4 fur MS8H 015650 i INTEGERELE(99),STA(2) OiS670

! DIMENS ION CUEFF ( 7,8001,hPdOD (T ,8 0 0 ) ,C APT (6) , Y,I. ELD (5,50 0) l DIMENSION -Y (5) ,NSORS (4) , T YLD (5) ,NUC AL (6)

D1 HENS!0N SKIP (20),MSOS(20),NAME(3)

DIMENSION TuCAP(800), TIPS (100), TITLE (20),4(809),FGt800),

I ALPHAN (100) ,SPONF (100 J , ABUND (5001, KAP (800) ,MH8X (800)

COMMON / LABEL / ELE,STA Di5750

' COMMON /CONST/MMN, ERR,MZERU COMMON /EQ/X1EMP(800),XNEhf10,800),8(800),D(800)

COMMON /FLUXN/ REGENT.DAS(dPO),ILITE,IA9T,ITOT COMMON /0UT/NUCL(800) .

' COMMON / MATRIX /A(2500)tLOC(2500),NON0(800),KD(400) 015850 l-LEVEL 2,A, LOC,NONO,KD 050600 COMMON /CCUEFF/COEFF LEVEL p,CUEFF EuulVALENCE (XNEW(1,401),hPROD(1,1))

Euu! VALENCE (Ai,0 LAM) 015880 DATA NUCAL/-20030,-10000,10,11,-10,-9/ 015890 DATA NSRS/922330,922350,902320,922380,942390,922330,922350,942410,0iS900

.) 922380.9423Yo,942410,922350,942400,Y22380,9423'i,'!2330,0iS910

?

922350,902320,922380,94g90/,

Oi5920

C 015930 C PROGRAM TO COMPUTE A NATRIX (TRANSITIUN MATRIA) FROM NUCLEAR DATA 015940 C- 0I5050 READ 9511, (T&TLE(!),1=1,18),NLIBE 0j5960 C IF(NLIBE.LT.0) PROGRAM WILL READ TAPE,IN CASotR FUHMAT 015970 IGWC=0 015980

.IF(NLIRE.GT.0) GO TO 10 015990 IGWC=1 ui6000

'NLIBE=1NLIBE O 6010

~ PRINT 9000 .

0 6020 9000 FORMAT (1HO,"WILL READ TAP 6 GENERATED BY CASD4HH) 0 6030 10 'N1=4-hLIBE Oi6040 20 HEAD 9561, THERM,RBS, FAST, ERR, NMO,NDAY,NYR,NPCTAB,INPT,IR OI6050 PRINT.9005, NMO,NWAY,NYR OI6060 PRINT 9006 Oi6070 PRINT 9007 016080

' PRINT 9008 016000 PRINT 9009 OI6100 PHINT 9010 0{6110 PRINT 9n13 016120 PHINT 9n14 Di6130

.C Di6140 C' -THCHM s NATIO OF-THERHAL 6 LUX TO TOTAL FLUX Oi6150 C :RLS = RATIu 0F RESON$NCE FLUX TO TOT $L FhUX Oi6160 C' FAST = RATIO OF FAST [ LUX TO TOTAL FLUX Oi6170 C EHR = TRUNCATION ERRON LIMIT Di6180 C Oi6100 C READ DATA FOR LIGHT EhEMENTS Oi6200 C- Oi6210 K=5*(NLIdE-1) 0i6270 00'30 K1=1,b Oi6230 K2=K.Kj 0i6240

'30 ' NSORS (K11 =MSRS '( K2 ) Oj6P50 PHINT 9018, THERM,RES, FAST,tNSORS(K),K=1,5),NLIBE 016P60

!=0 Oi6270 NOTAPE=n 016790 40 ~I=I*1 016290 50 NE AD (6,9 034) NUCL (I) ,Dh AM ,10,FB1,FP ,FP1.F T ,F A s tSF ,0 (I) ,FG i l ) , ABUN0 ( Di630 0.

11),DUNY1,00MY2 IF ( EOF '( 8 ) .NE. 0 ) GOT0269 016370 IFt1 GWC.GT.0) GO TO {0 016330 DO 60 N=1,NLIBE OI6340 60 READ (6,9035) SIGTH.FNhl,FMA,FNP,RITH,FINA, FIN (,SIGMEV,FN2N1,FFNA, 0i6350 1 FFNP,IT 0i6360 GO TO 90 0I6370 70 DO 80 N=1,NLIBE 016380 60 NEA0(8,9040) SIGTH FN91,FNA,FNP,RITH,FINA, FIN [,SIGHEV,FN2N1,FFNA, 016390 1 FFNP,IT Oi6400 90 IF(N1.EO.0) GO TO 110 016410 00 100 N=1,N1 016420 100 : HEAD (6,9036) SKIP 016430 110 IF(IT.EQ.0) GO TO 50 016440 120 Ms0 016450 CALL NALF(A1,IU) 0{6460 NUCLI=NUCL(1) 016470 IF(NUCL I.Eu.G) GO TO 260 0164R0 CALL NOAH (NUCLI,NAME) Oi6400 IF(MOD (I-1,50) .Eu 01 PHINT 9012, (TIILE (N),N=1,18) 0{6500 IF(MOD (I-1,50) .EG. 01 PHINT 9016 016510 SIGTH=THEHM*SIGTH Oj6570 HITH=RFS*HITH 016530 SIGMEVsFAdT*SIGHEV Oi6540 SIGNA sIGTH+FNA+RITH*lIhA*SIGHEveFFN4 Di6550 3-45

_ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _]

SIGNP:$IGIM*FNP.RITM*tIhP+SIGMEV*FFNP 016560

-FNG=1 0 FNA-FNP 016570 IFiFN9.LT.1.0E-4)FNG=V. Oi6580 F. I NG= 1'. 0-F I N A-F I NP 0J6590 IF(FING.LT.1.0E-4)FINus0. 016600

- FN2N=I*.0-FFN A-FFNP Oi66io IF(FN2N.LT.1.0E-4)FN2N=0. Oi6620 SIGNG=SIGTM*FNG+RITM*GING OI6630 SIGN 2N=SIGHEV*FN2N 016640 130 RRINT 9033, .NAME, DLAM.F91.FR,FP1,FT,EA SIGNG, 016650 1 FNG1,SAGN2N,FN2N1, SIGN A ,SIONP,4 (I) ,FG (I) , ABUND (I) OI6660 C TEST RADIOACTIVITY 0i6670 C .

OI6680 140 IF(A1.LE. ERR) GO TO 180 Oi6690 ABETAsi.0 Oi670q C-016710 C TEST POSITRON EMISSION 016720 C Oi6730 IF(FP ".LT. ERR) GO TO 150 OI6740 M*M+1 Oi6750 COEFF(M,I)=FP*Al 0I6760 NPROD(M,1)=NUCLI-10000 Oi6770 ABETA=ARETA-FP Oib780 C Oi6700 C. TEST POSIIRON EMISSION To EXCITED STATE Of PRuDUCT NUCLIDE Oi6800

.C _

OI6810

.IF(FPI .LT. ERR) GO TV 150 Oi6820 H=M+1 016R30 COEFF (M,I) =f Pl*COEFF (M-1,1) 016840 NPROD(4,1)=NPROD(M-1,1)*1 Di6850 COEFF(M-1,1)=COEFF(M-1,1)-COEFF(H,Il Di6R60 C Oi6870 C TEST ISOMERIC TRANSITA0N Oi6880 C ' 016890 150 IF(FT .LT.EHR) GO TO 160 016900 M=M*1 036910 COEFF(M,I)=FT*Al Di6920 NPROD(M.I):NUCLI Gi6930 ABETA=ABETA-FT 016940 C 016950 C TEST ALPHA EMISSION Oi6960 C. 0I6970 160 IF (FA *.LT. ERR) GO TO 170 016980 H=M+1 016990 COEFF (u,1) =F A* Al 057000 NPROD(M,I)=NUCLI-20040 017010 M=M*1 OI7020 COEFF(M,1)=COEFF(M-1,A) 017030 NPROD(4,I)=20040 017040 AbETA=APETA-FA Di7050 C 0'7060 1

C TEST hEGATRON EMISSION OI7070

, C OI7080 170 IF(ABETA.LT.1.E-4) GO TO 180 017090 M*M+1 017100 COEFF (u ,1 ) = ABE T A* A t Oi7110 NPH00(M.I)=NUCLI+10000 Oi7120 C- Di7130 C TEST hEGATRON EMISSION TO EXCITED STATE OF PRuDUCT NUCLIDE 017140 C. OI7150 IF (FH1 .LT. ERRIGO TO 180 0i7160 M*M+1 3-46 OI7170

. CUEFF (M. I) =F BI COEFF (M-1,1) Oj71A0 HPHOD(M,1)=NPROD(M-1,.1)+1 017140 COEFF (M-1,1) =COEFF (M-1,1)-COEF F (M,I) OI7200 C Di7210 C COMPUTE NEUTRON CAPTUWE CROSS SECTIONS IN THdEE REGIONS 0i7220 C Oi7230 180 NAP (I)=M Oi7240 00 190 MI=1,6 017250 190 'CAPT(MI) =0 0 Oi7260 CAPT(1)= SIGNA 0}7270 CAPT(2)=SIGNP 017280 CAPT(4)=SIbhGeFNG1 OI7290 CAPT(3)=SIGNG-CAPT(4) Oi7300 CAPT(ol= SIGN 2NeFN2N1 Di7310 CAPT(5)=51GN2N-CAPit61 Di7320 200 TOCAP(I)=0.0 Oi7330' C TOTAL NEUTRON CROSS SECTION FOR NUCLIDE(I) 017340 DO 220 K=1,6 Di7350 CAPKI4CAPT(K) DI7360 IF (C APKI .LT. ERR) GO {0 220 Oj7370 McM*1 017380 NPROD(M,1)=NUCLI+NUCAL(K) OI7390 COEFF(M,I)=CAPKI OI7400 TOCAP(I)=TOLAP(11+CAPQI Di7410 I F ( K .!'E .1 ) GO TO 210 OI7420 M=M+1 OIT430 CUEFF(4,1)=COEFF(M-1,1) Oi7440 NPROD(M,I)=20040 Oi7450 210. IF (K.NE.21 60 TO 220 Oj7460 M=H+1 017470 COEFF(M,1)=COEFF(M-1,A) OI74HO NPHOD(H,I)=10010 Oj7440 220 CONTIh0E 017500

'230 IF (MOD (NUCLI, 10).EQe0) ?O TO 250 017510 00 240 K=1,M 0175?0 240 NPROD(K,1)=NPROD(K,I)*1 Oi7530 250 MMAX(I) =M 017540 IF(M.0T.7) PRINT 903Y, M 017550 DIS (!) = Al 0i7560 90 TO 40 057570 260 ILITE = I-1 017580 IACT=0 Di7540 C OIT600 C READ DATA ON ACTINIDE) Oj7610 C 017620 270 RE AD ($ ,9 0 34 ) NUCL (I) ,Dh Ad e lU,FB1,FP,FP1,F T s FA ,(SF ,0 (I) ,FG (I) , DUMMY ,0iT6 3 0 1DUMY1,00MY2 I F (EOF, ( 8 ) .NE. 0 ) GOT0459 Di7650 DO 280 N=1,NLIBE Ui7660 READ (6,9037) SIGNG,RINGeFNG1,SIGF,RIF,SIGFreS.I.GN2N,FN2N1, SIGN 3N,IT0iT670 280 CONTIhUE 017680 IF(N1.EQ.0) GD TO 300 017690 DO 290 N=1,N1 OI7700

.290 NEAU(0,9036) SKIP OI7710 300 IF (IT .Eu. 01 GO TO 210 0{7720

'310 N=0 017730

.NUCLI=NUCL(I) Oi7740 IF(NUCLI.Lu.0) GO TO 450 Oi7750 00 320 K=1,5 017760

- IF (NUCL I .Eu.NSOHS (K)) K50RSth)=1 OI7770 320 CONTINUE Oi7780 CALL NALF(A1,IU) 017790 CALL NOAM(NUCLI,NAME) Oi7800 3-47

SIGNG9THEbM*SIGNG+HES* RING '0I7810-SIGF gyHEHM*SIGF'+RES?RIF + FAST *SIGFF 017820 SIGN 2N= SIGN 2N* FAST Oi7830 SIGN 3N=$IGN3N* FAST 0i7840 IF(M00(IACT,50).EG.0) PRINT 9012, (TIlLE (N),N=1,18) 0i7850 330 IF(M0p(IACT,50).EQ.0) PRINT 9024 Oi7860

PRINT 9026, NAMEe DLAM,F81.FP,FP1,FT.FA,FSF,$IGNG,0I7870 1 FNG 1, S} GF , SIGN 2N , S IGN3N ,0 (1 ) , F) ( I ) DIT880 340' IACT=IACT+1 Oj7890 C

C 017900 TEST RADI0 ACTIVITY OI7910 C .

Oi7920.

IF(A1.LT. ERR) GO To d80 OI7930 ABETAsi.0 OI7940 C TEST POSITRON EMISSION DI7959 IF(FP 'LT. LRR).G0 TO 350 Oi7960 ASETA=ARETA-FP Oi7970 M*H+1 OI7980 COEFF(M,1)=FP*Al OI7990 NPROD(M,I)=NUCLI-10000 ci8000 C POSITRON EMISSION To EXCITED STATE OTRO10 IF(FP1'.LT. ERR)GO TO 350 OI8020 M=M*1 0i8030 COEFF (M ,1 ) =F P l *COCFF (M-1,1) .0i8040 NPROD(M.I)=NPROD(M-1,1)*1 .

0i8050 COEFF(M-1,1)=COEFF(M-1,11-COEFF(M,Il OI8n60 C ISOMEMIC THANSITION 0iA070 350 IF(FT .LT.ERRIGO TO 300 M=M*1 0{8080 018090 COEFF (M,1) =FT* Al NPROD(M,1):NUCLI DiB100 018110 ABETA=ABETA-FT Oia120 s

C ALPHA EMISSION 10j8130 360' IF(FA .LT.EHRIGO TO 3{0 018140 H=H+1 Oi8150 COEFF(M,1)=FA*Al DI8160 NPROD(M,1)=NUCLI-20040 Gia170 H=M+1 - Dis 180 00EFF(M,1)=COEFF(M-1,A) OI8190 l N"'00(M,1)*20040 Oi8200 AbiTA=ARETA-FA C UETA DECAY 0}8240 018220 370 IF(ABETA.LT.I.E-4) GO TO 380 0{8230 MsH+1 018240 L COEFF(M,1)=ABETA*Al 0I8250 NPROD(M,I)=NUCLI+10000 OI8260 IF(FBI .LT. ERR)GO TO 390 OI8270 M=M+1 COEFF(4,I)=COEFF(M-1,()*FW1 018280 018290 COEFF (M-1,1 ) =COEFF (M-1,117COEFF (M ,1 )

NPROD(M,1)=Npn00(M-1,1)+1 0{8300 018310 C

I C 018320 NEUTRON CAPTURE CROSS SECTIONS 018330 C

380 KAP (I)=M 0{8340 00 390 K=1,6 018350 018360 390 CAPT(K  !=0 0 Dia370 CAPT(d)=SIGNGeFNG1 , 018380 CAPT(1)=SIGNG-CAPT(2) OI8390 j CAPT(4)=C1GN2NeFN2N1 OI8400 CAPT(3)= SIGN 2N CAPT(41 OI8410 1 400 FISS(IACT)=SIGF 0 8420 348

__ , , , , , , _ . * - + + - - r v'er WV'* *'T ""D*

TOCAP(I)=0.0 018430 DO 410 K=1,4 018440 CAPK!*CAPT(K) .

Oi8450 IF(CAPKI.LT. ERR) GO TO 410 Oi8460 M=M+1 0I8470 TOGAP(!)=TOCAP(I)+ CAP 61 OI8480

'COEFF(M,1)=CAPKI Di8490 NPROD(M I)=NUCLI+NUCAh(K+E) OI8500 410 CONTINUE OiS510 TOCAP(I)=TOCAR(1)+FISf(IAET) Di8520 C- N-3N CROSS SECTION OiB530 A17= SIGN 3N Oi8540 IF ( A1 7 '.LT. ERR) GO TO $30 0i8550 M=d+1 .0i8560 COEFF(M ,1)= Ali Di8570 NPROD(M ,Ils NUCLI.29 Oi8580 TOC AP (I) = TOC AP (I) + A]' Oi8500 420- IF (M09 (NUCLI,10) .EQ.01 GO TO 440 OiB600 DO 430 K=1,M OI8610 430 'NPROD(K,1)=NPROD(K,1)?1 018620 440 MMAX(I)=M .

Oi8610 IF (M.GT.7) PRINT'903y, M 0[8640

.SPONF(IACT)=FSF*Al*6.023E23 018650 4LPHAh(IACT)=FA*Al*6.023E13*0(!)**3 65 0i8660

. DIS (I)=Al OiB670

!=I+1 Dia680 90 TO 770 0i8690 450 IL=0 018700 DO 4h0 K=1,5 Dia710 460 ' TYLD (K) =0.0 0i8770 C Oi8730 C READ DATA FOR FISSION PRODUCTS Oia740 C 018790 470 HE AD (S .9034) NUCL (I) ,DL AM,10,Ful ,FP,FP1,FT ,FA e tS F ,0 (I) ,FG (I) ,00MMY ,0I8760 IUuMY1,0UHY2 IF(EOF (8).NE.0)GOT0690 Oi8780 DO 480 N=1,NLI8E Dia700 480 NEAD(8,9038) SIGNG. RING,FNG1,Y,IT '0i8800 IF(N1.EO.0) GO TO 500 Oie810 DO 490 N=1,N1 0i8820 490 REA0(0,9036) SKIP OIA830 500 IF (IT ".EQ. 0) GO TO 4To 0{8840 510' M=0 018850 CALL HALF (A1,1U) OI8860 1 520 NUCLI=NUCL(I) -0[8870 IF(NUCLI.EQ.0) GO TO 990 018880 CALL NOAH (NUCLI,NAME) Oi8800

IF(MOD (IL,50).EQ. Oy PRINT 90126 (TITLE (N),N=1,18) 0i8900 SIGNG9 THERM *SIGNG+RES!R.ING Oi8910 IF(NLIRE.EQ.3) GO TO 3 0 0{8020 530 IF(MOP ( IL 501.EO.0) PRINT 9019 018930 PRINT 9021, NAME, DLAM FB1,FP,FP1,FT,SIGNG, 018940 1 FNG1,YeQ(I),FGfIl 018950 GO TO 550 0 8960 540_ IF(MODg!L,50).EQ.01 PKINT 9020 0{i8070 PRINT 9022, NAME

• DLAM,Fal,FP,FP1.FT.SIGNG.FNG1, 0I8980 1 Y(2),Yl4),Y(5),0(I),FG(I) OIB990 C 019000 C TEST SADIUACTIVITY OI9010 C Oi9020 550 IF(A1.LT. ERR) 00 TO 000 0I9030 ABETA=1.0 0{9040 C. P051T80N LMISSION 019050 3-49

.l I

l

~

A3=FP .

OI9060 IFIA3.LT. ERR) GO TO hT0 019070-

'AbETAgABETA-A3 OI9090 APl=A8eFP1 019090 APmA3-API 019100

'IF(AP.LT. ERR) GO TO $60 Di9110 H=M+1 OI9120 CC'iFF (M,1) = ape A1 OI9130

'NfROD(M,1)=NUCLI-10009 0I9140 560~ IF ( APl*.LT. ERR) GO TO 510 0I9150

'M=M+1 OI9160 COEFF(M,1)=APleA1 Di91T6 NPROD(M,1)=NUCLI-9999 0J9180 C ISOMESic TRANSITION 039190 ST0 IF(FT '.LT. ERR) GO TO 500 OI9200 MsM*1 019210 COEFF(M,1)=FTeAl 019220

, NPR00(M,I)=NUCLI OI9230

-ABETA=ABETA-FT 0i9240 C NEGATRON EMISSION OI9250 580 IF(ABETA.LT.1 0E-43 GU TC 600 Oi9260 A2=F81 Di9270 ABl=AWETA*A2 Oi92n0 AB=ABETA-Ad1 DI9290 IF(AB.LT.1.E-43 GO TO 590 Oi9300 '

.M=M+1 OI9310 COEFF(4,1)=ABoAl Oi9320 NPROD(4,1)=h0CLI+10000 Oi4330 590 IF ( AH1'.L T .1.E-6) Go TU 600 Oi9340 M=M*1 OI9350 COEFF(p.1)=AR1*Al Oi9360 NPRUD(p,I)=NUCLI+I000& OI9370-C Di93R0 C NEUTRON CAPTURE CROSS SECTIONS FOR FISSION PauDUCTS USING.THREE 019390 C REGION APPROXIMATION 010400 C OI9410 600 KAP (!)=M 0i9420 DO 610 K=1,6 Di9430 610 .CAPT(K)=0 0 Oi9440 CAPT(d)=SIGNGeFNG1 Oj9450 CAPT(1)=SIGi4G-CAPT(2) 019460 TOCAP(I)=0 0 Oi94TO DO 620 K=1,2 0194A0 CAPKI=CAPT(K) 0194?0 IF(CAPKI.LT. ERR) GO TO 680 OI9500 M=M*1 019510 TUCAP(I)=TOCAP'(I)+ CAP 61 019520 COEFF (M,1 ) =C APKI 019530 NPROD(M,1)=NUCLI+NUCAL(K+2) 019540 620 CONTINUE Di9550 630 IF(M0p(NUCLI,10).EQ.01 GO T0 650 OI9560

~DO 640 KaleM 059570 640 NPROD(K.I)=NPROD(K,1)*1 0399R0 650 IL=IL+i OI9590 00 660 Jal,5 019600 YJ=Y(J)*0.010 019630 TYLD(J)=TYLD(J)+YJ Oj9620 660 YIELO(J IL)=YJ 019630 IF(NLIRE.EG.I.OR.NLISE.EG,4) GO TO 680 OI9640

'6TO IF(NLIpE.Eu.3) YIELDil,IL)=YJ OI9650 YIELD (3,IL)=YJ Oiq660 4

6n0 MM Ax (1) =M 3-50 Oi96TO

F

.IF(M.9T.73- PRINT 903Y, M OI9680 DIS (IlsA1 019690 I=I+1 Oi9700 GO TO 470 DI9710

.690 IFP=IL DI9720 C 0i9730 CL ALL. DATA ON NUCLIDES MAS WEEN READ, BEGIN TO LOMPUTE HATRIX COEFF OI9740 C Oi9750 ITOT=I.1 Di9760 C .

Oj9770'

'C FIND PRODUCT NUCLIDES FOR REACTIONS OF LI9HT ELEMENTS 019780 C' OI9790 NONz0 0[9800 DO 700 K=1,ITOT 019810 700~ NON0 (K) s0 Oi9820 IF(ILITE.LT.1) GO TO 760 01sC3F 00 750 !=leILITE 019840 NUCLI=NUCL(I) OI9850 UO 720 J81,ILITE OI9860 KMAX= KAP (J) 0[9870 IF(KMAX.LT.1) 00 TD (20 0198a0 00 710 Mal,KMAx- 0{9890 IF (NUCLI .NE.NPROD (M,J1) GW TO 710 019900 NON0(1)=NON0(I)+1 OI9910 NONaNON+1 0[9420 I F (N0h'.GT .25 0 0 ) PRINI 9041, NON,NUCL(I) 019930 A (NON) mCOEFF (M,J) 0[9940 JT=J 019950 LOC (NON)=JT 039960 710 CONTINUE 019970

.720 CONTINUE 0i9980 KD (1) sNON0 (1) Di9090 00 740 Jul,ILITE 070000 K1 =K AP '( J) + 1 070010 KMAXsMMAA(J) 070020 IF (KMAx.LT.K1) GO TO 740 070030 00 730 M*K1,KMAX 070040 IF(NUCLI.NE.NPROD(M,J1) GU TO 730 070050 NON0(I)sNON0(1)*1 070060 NON=NON+1 020070 IF (N0h'.GT .250 0 ) PRINI 9041, NON,NUCL(Il 070080 A(NON)=COEFF(M,J) Op0090 JTgJ 070100 LOC (NON)=JT 070110 730 CONTINUE 070120 740 CONTINUE' 020130 750 CONTINUE 070140 C .

070150 C NON ZERO MATRIX ELEMENTS (OR THE ACTINIDES OP0160 C 070170 760 IF(IACT.LT.1) GO TO 920 070180 Zu=ILITE+1 020190 11=ILITE+1ACT 070200 DO 810'!=10,11 020210 NUCL!sNUCL(I) 070220 00 780 Jz!O,Il 070230 MAX *KbP(J) 070240 IF (M AN.L T.1) GO TO 790 070250 DO 770 MaleMAX .

OP0260

. IF (NUCLI .NE.NPROD (H,JII 40 TO 770 0P0270

- NONO (!) sNON0 (!) +1 070280 NON=NON 1 070200 IF (NON*.GT .2500) PR1H1 9041, NON,NUCL(!) 070300 3-51

l A (NON) =COEF F (H ,J) 020310  ;

JT=J 070320 LOC (NON)sJT 070330 770 CONTINUE OP0340

_780 - CONTINUE ,

020350 KD(I)=NON0(1) 070360-DO 800 JsIO,Il 070370 M1gKARIJ)+1 070380 M2sMMAX(J) . _

070390 IF(M2eLT.M1)_ GO TO 890 020400 p0 790 M=M1,M2 070410 IF(NUCLI.NE.NPROD(M,J1) GU TO 790 020420 NON0(!) ANON 0(I)*1 020430 NON=NON.1 070440 IF (NON'.G T . 2500 ) PRINI 9061, NON,NUCL(I) 070459 A(NON)=COEFF(M,J) 070460 JT=J 070470 LOC (NON)=JT 070480 790 CONTINUE 020400 000 CONTINUE 070500 810 CONTINUE 070510 C 070520 C MATRIX ELEMENTS FOR FISSION PRODUCTS 070510 C 070540 820 I F ( I FP'. L T .1 ) RETURN IM=ILITE+1ACT 020560 10mIM i 070570 IF(ITOT.LT.IO) RETURN 00 880 !=IO,ITOT 070590 NUCLI=NUCL(1) 070600 12 = M A X g' ( 10,1 -10 ) 070610 13= MIN 0(ITuf.I.10) 070620 00 840 J8I2,I3 070630 KMAAzKAP(J) 070640 IF(KMAX.LT.1) GO 70 940 0F0650 DO 830 Mal,nHAK 070660 IF(NUCLI.NE.NPROD(H,J1) Gd TO 830 07c670 NON0(IjsNON0(I)+1 070680 NON=NON.1 ~ 070600 I F ( N0h'. G T . 250 0 ) PRINI 90#1, NON,NUCL(I) 020700.

A(NON)=COEFF(M,J) 020710 JT=J. 070720 LOC (NON)=JT 020730 830 CONTINUE 070740 840 CONTIr:0E 070750 KD(I) ANON 0(I) 020760 -

DO 860 J=12,13 i

070770 K1sKAPid)+1 070780 KMAXaNMAX(J) 070790 IF(KMAx,LT.K1) GO TO '60 070800 DO 850 N=nl,KMAX 070810 IF(NUCLI.NE,NPROD(M,J1) GW TD 850 070820 NON0(!)=NON0(I)+1 070830 NON=NON.1 070840 IF(N0h.GT.2500) PRINT 9061, NON,NUCL(I) 070850 A(NON)=COEFF(M,J) 070860 JT=J 070870 LOC (NON)=JT 070P80

.850 CONTINHE 070890 860 CONTINUE 070900 IF(IACT.LT.1) GO TO 680_ 070910 DU 870 Kal,5 070920 3-52 L

ILal IM . 070930 IF(YIELD (K,IL).LT. ERR 1 GU TO 870 020940 NON=NON+1 070950 I F ( N0b'.GT . 250 0 ) PRINT 9041, NON,NUCL(1) 020960 NON0 (I):NON0(1)+1 070970 KKsNSORS(K) 070980 LOC (NON)anK 070990 KF=KK-ILITE 051000 A(NON) YIELD (K,IL)* PISS (K6) 071010 870 CONTINUE 071020 880 CONTINUE 021030 IF (I FP'.LE . 0 ) GO TO 909 071040 IF(NLIBE.NE.3) GO TO 990 051050 PRINT 9027 TYLDt21,TYLD(4),TYLD(5) 071060 GO TO 900 071070 890 PRINT 9030, (TYLD(I),1 1,5) 071080-C 071090 C ALL HATRIX ELEMENTS AHE h0W COMPUTED 071100 C. HEGIN TRANSIENT SOLUTAON 071110 C 071120 C 071130 C TEMPORARILY WRITE OUT HATNIX ELEMENTS 071140 C 071150 900 IF(IR '.Eu. 0) RETU9N 071160 PRINT gn29 071170 N=0 0211A0 00 910 !=1,ITOT 051190 NUM=NONn(1) OE1700 IF (NUN.LE.0) GO TO 910 071710 N1=N+hpM 071270 N=N+1 071730 PRINT on28, I . DIS II) ,IOC AP (1) , ( A (K) , LUC (K) ,K=N,N1) 071240 N=N1 071750 910 CONTINUE. 071260 RETURh 071270 920 STOP 071?R0 C 071200 C FORMATS FORMATS FORMATS FORMATS 071300 C 021310

'9001 FORMAT (4F10.5,612) .

071320 TEVISED ",12',"/",I2,"071370 9005 FORMAT (IH1,43X," NUCLEAR THANSMUTATION DATA 1/",I2,/,"0HUCL s NUCLADE F 10000

• ATOMIC NO

• 10

• MASS NO + ISOM071380 2ERIC STATE (0 OR 1)",10X,"DLAM = DECAY CONSTANT (1/SEC).Hi/," FB, 071390 3FP, FA, FT = FRACTIONAL DbCAy BY BETA, POSITRUN (OR ELECTRON CAPTUO71400 4RE), ALPHA, INTERNAL TRANSITION. FB = 1. FP - F A - FT"i/," FB1,071410 5 FP1, FNG1, FN2N1 FHACTION UF BETA, POSITRQH,~N-GAMMA, N-2N TRAN0?1420 6SITIONS TO EXCITED ST6TE OF PRODUCT NUCLIDE"e/," SIGTH, SIGNG, SIG071430 7F, SIONA, SIGNP = THEHMAL CROSS SECTI,uNS (BARNS) FOR ABSORPTION, N051440 8-GAHMA, F1SSION, N.AL6HA, N-PHOTON.") 0?1450

'9006 FORMAT (" SIGNG = Slu1H * (1 - FNA FNP), S.I.GNA = SIGTH + FNA. 071460 ISIGNP SIGlH

• FNP. fha *, FNP a FRACTION THETMAL N-ALPHA, N-PROT 0071470 2N,",/," RITH, RING, RIF, BINA, RINP a RESONANhE INTEGRAL FOR ABSOR071480 3Pl!ON, N-GAMMA, FISSIPN, N-ALPHA, N-PROTON.",/," RING s RITH + (071490 41 - FINA - FINP). RINA = HITH + FINA. RINF = MITH

• FINP. FINA, F071500 SINP = FRACTION RESONANCE N. ALPHA, N-PROTON."e/," SIGMEV, SIGFF, $1071510 6GN2N, SIGNAF, SIGNPF y FAST CROSS SECTIONS (StRNS) FOR ABSORPTION,071520 7 FISSION, N-2N, N-ALPN3, h-PRO T ON.",/," blGNdN a SIGHEV * (1 FF071530 8HA - FFNP). SIGNAF = SIGHEV e FFNA. SIGNPF

• SIGHEV

• FFNP. FFN071540 9Ai FFhp = FRACTION FAST N7ALPNA, N P.") 071550 9007 FORMAT (" Y23, Y25, Y0d, Y28, Y49 = FISS10H YIELD (PEHCENT) FROM 23021560 13-U, 235-ue 232-TH, 2J8-Ut 239- PU.",/," w = OEAT PER DIS 1NTEGRATIO71570 20N. FG = FRACTION OF HL A7 IN GAMHAS OF E!*ERGT GREATER THAN 0.7 8E071580 3V.",/,"O LftECTIVE CdOSS bECTIONS FOR A VOLUME AVERAGED THERNAL (LO71540 3-53

1 47.0 876 EY) FLUX ARE AS FULLOWS.",/," '

H GAUMA - ~ SIGNG

• THEnM021600 5 + RING *-RES.",/," FlySION - SIGF

• THEHM + RIF

• RLS + SIGF071610 .

6F

• FAST.",10X," THERM = 1/V CORRECTION FON THERMAL SPECTRUM AND TE021620 7MPERATURE.",/," N-4N - $1GN2N

• FASTn",36X,"RES = RATIO 021630 80F RESnNANCE FLUX PER LETUARGY UNIT TO TNLRMAh FLUX.") 071640

.900s FORMAT (" N-ALPHA ? SIGNA

• THERM + RINA

• RES + SIGNAF

• FAST 071650 1.",7X," FAST =fl.45

• ratio 0F FAST (GT 1*0 M$V) TO THERMAL FLUX "0s1660

-2/" N PROTON - SIGNP

• THERM + RINP

• YES

• SIGNPF

• FeST.") 021670 9009'FORHAfi1H0,59X,"PEFERENCEb",/," HALF L4VESv DECAY SCHEMES, AND OP16R0 1 THERMAL POWER"g/," C 4 LE9ERER, J H HOLLANDER, AND I PERLMAN "" TAB 0g1690 2L E OF- ISOTOPES - SIXTQ EOLTION"" JOHN WIL?Y AND SONS,' INC (1967)",071700 3/," 8'S DZNELEPOV AND L K PEKER "" DECAY SCHEMES OF RADI0 ACTIVE NuC071710 4 LEI"" PERGAMMON PRESS (1981)",/e" D T GOLUMAN AND JAMES R.ROSSER "071720 5" CHART OF THE NUCL1DEy"" NINTH EDITION GENERAL ELECTRIC CO (JULY 071730-61966)",/," E D ARNOLD "" PROGRAM SPECTMA"" APPENDIX A 0F ORNL-3576 071740 7(APRIL 1964)") 021750 9010 ORMATi" CROSS SECTIOh6 AND FLUX SPECTRA"#/," 8 E PRINCE ""NEUT071760

'N REACTION RATES IN THE MSRE SPECTRUM"" ORNH-4119, PP 79-83 (JULO71770

'967)",/," B E PRINSE "" NEUTRON ENERGY SPECTRA IN MSRE AND MSBH"071780

. JRNL-4191, PP 50-58-(DEC 1967)",/," M D GOLUBERG ET AL "" NEUTRON 071790 4 CROSS SECTIONS"" RNLt325. SECOND ED, SUPP NO 2 (MAY 1964 - AUG 19021800 566) ALSO EAHLIER EDITA0NS",/," H T KERR, UNPUOLISHED ERC COMPILATIO21810 ADN (FE8 1968)",/," M h DROKE ""A COMPILATION uF RESONANCE INTEGRAL 071820 75"" NUCLEONICS, VOL 24, hu 8, PP 108-111 (AUG 1966)",/," BNWL STAF0pl830 BF "" INVESTIGATION OF N 2N CROSS SECTIONS"" BNFC-98, PP 44-98 (JUNE 021840 9 1965)") 021850 9011 F0HMAT(18A4,13) 071860 9012 FUdMAT(1H1,20X,18A4t 021870-9013 F0HMAT(" H ALTER AND E E WEBER "" PRODUCTION Qt H AND HE IN METALS Oh1880 IUURING REACIOR IRRADI$T 1 0H"" J NUCL MATLS, V0h 16, PP 68 73 (19651071890 2",/," L L BENNETT "" hec 0MNENDtD FISSguN PMODubT CHAINS FOR USE I4 071900 3HEACTOR EVALUATION STUDIES"" ORNL-TM-165S (SEUT 19 6 6)") 071910 9014 FORMAT (" FISSION PH000CT YIELDS",/," M E HEEK AND B F kIDEH, ""051920 I-1

SUMMARY

OF FISSION PRUDUCT YIELDS FOR U-235, U-238, PU-239, AND PUO21930 -

2-241 AT THERMAL, FISSION SPECTRUM AND"/" 14 MEV NEUTRON ENERGI0plo40 3ES"" APED-5398.A(REV.1,(OkT. 1968)"/" S KMTCDEF "" FISSION PRODUCT 021950 4Y1 ELDS FHUM NEUTRON INDUC6D FISSION"" NUChEONACS, VOL 18, NO 11, 071960 5(NOV 1960)"/" N O DVDEY "" REVIEW OF LOW-MASS ATOM PRODUCTION IN F071970 6AST REACTOHS9" ANL-74J4,(APRIL 1968) ") 071980 9016 FORMAT (1H0e20X," LIGHT ELENENTS, MATERIALS OF EONSTRUCTIONi AND.ACT022030

( IIVATTION rdODUCTS ",/,"O NUCL DLOM F#1 FP " 050640 2"FPI FT FA S.1GMG FNG1 SIbN2N FN2N1" 050650 3" SIGNA SIGNP 9 FG ABUNDANCE") 000660 -

9018 FORMAi(1HO,10X," THERM 9 "F10,5,5X,"RES" "F10.565X," FAST = "f10,5, 022000 1//,1X,"NEUTHON SOURCEy "5(Il0%5X)iSX,"NLIPE= "I3) 022100 e 9019 EORMATi1H0,36X,"FISSIMN PWODUCTS"i/,"O NUCL DLAM " 060690 l' " FBI FP FP1 FT -SIGNG FNG1 Y23 " 060700 2 "Y25 YO2 Y29 Y49 0 FG") 000710

.9020 FORMAT (1HO,36X," FISSION PUODUCTS",/,"O NUCL DLAM FBI " 000720 1 "FP FP1 FT SIGNG FNG1 Y2P Y28" 000730 2 " Y49 'O FG") 090740

-9021'FORMATj1H ,A2,I3.A1,1PE10 2,0P4F7.3,1PE10 2,0tFT.3, On0750 1 1PSE10.2,0P2F7.33 060760 9022 FORMATi1H ,A2,13 A1,1fE10 2,0P4F7 3,1PE10 2, 050770 1 ' 0PF7' 3,1P3E10.2,0P2t.7.3) 060780 9024 . F 0W M AT ,( 1 H 0,32X , " ACTINIDE? AND THEIR DAUGUTERb",// OP22PO

'1" !NUCL 'DLAH (01 FP FP1 FT" 060800 2" FA FSF E+6 SIGNG FNG21 SIGF" 060810 3" SIGN 2N SIGN 3N Q FG") 050820 9026 FORMAT (1H e d2. !3, A1,1(E10 2,0P5F 7 3,6PF10 1,1tE10.2, 000830 1 ~ OPF7'.3,1P3E10.2,0PFO 3,P6.2) 060840

.902T FORMATg"05UM OF YIELUS OF ALL FISSION PR0pVCT) =",15A,1P3E9.2) 022200 3-54 s - . .. _ _ . . . _ _ _ _.__ _ _ _ ~ _ _ _ _ . _ - _ _ _ _ _ _ - - _ - - - - - - - - - - - - -

9028 F uRM A! (15 2K,1PE10.3,3x ,E 10 3,5 (2X,E10.3 3X,14) / (30x.5 (2X',E 10 3, 022300 3x,15))) 072310 1

072320 9029 FORMAT (alNON.ZEHOlMATHIX 4LEMLNTS AND THEIR LuCATIONSn/

I- DIS (I) cat (I) A(1.J) J A(1,J) 072330 in J n) 022340 2J .

A(1,J) J A(1,J) J A(I,J) 9030 FORMATi63M05UM OF YIELDS UF ALL FISSION PNODubTS 1 1,5(E9.2,1X))

'9033 FORMAT (IM .A2,13,A1,1fE10 2,0P5F7.3,1PE10.2,0tF7.3,1PE10.2, 050860 1 OPF Z'.3,1P2E10.2,0P2t7.3 % F8.3 ) 060870 9034 FORMAT'(I7,F9.3,II,5F5,3,1HE9.2,0P2FS.3,F7 3,2E6 0) 022420 9035 FORMAT (7x,F9.2,3FS.3,F9.2',2F5 3,F9.2,3F5.3, 5A,11) 022430 9036 FORMAT (20A4) 072440 9037 FORMAT (7x,2F9.2,F5.3,4F9.2,F4.1.F9.2,11) 000890 9038FuMMAT(7x,2F9.2,F5.3,hF9.2,4X,II) 022460 WARNING, MuuT OF RANGE IN NUDATA, =n 15) 072470 9039 FORMAT ("O 9040 FORMAT ( 7A,F9.2,3F8,6,F6.2,2F3.1 F9 2,3FS.3,5X,11) 072480-9041 f0HMATi"0 NON MAS EXCEEDED 2500, EaVAL TO n219) 072490 Ed3 OP2500

~

• DECK COLLECT 0725io hu8 ROUTINE COLLECT (TMp,CWASTE,ITOT)

COMMON / Eu/ATEMP (800) ,XNEW (10,800) ,8 (800) ,P(800) 072550 OIMENSION CHASTE (800) 072560 IF ( TM8.LT .1) RETURN 00 10 !=1,110T 072570

'H(I)=CWASTE(I) 072580

_ATEMP(!)=0 0 072590 10 CALL DECAY (1.TMB,ITOT1 072600 CALL TERM (TMB,1,ITOT)

CALL EQUIL(1,ITOT) 072670 uu 20 !=1,110T 072630 20 'CWASTE(1)=XNEW(1,11/TMa 0FF640 RETURN 072650 ENO 072660

• DECK STORAG 022670 SUBROUTINE STORAG(TMBACWAbTE,ITOT)

COMMON /EQ/XTEMP(A00),XNEW(10,000),8(800),P(800)

DIMENSION CWASTE(ITOT1 0727io 072720 IF (TMB'.LT .1) RETURN 072730 DELT=TMB DO 10 I=1,ITOT 072740 U(I)=0.0 072750 10 ATEMP(I)=CWASTE(I) 072760 CALL DECAY (1,DELT.IT01) 072770 CALL TERM (TMR,1,ITOT)

CALL EQUIL(1,ITOT) 072790 DO 20 y=1,ITOT 072800 20 CWASTE(I)=XNEW(1,I) 072810 NETURN 072820 END 072830

• DECK ULKDAfi OP2840

-C PH0 GRAM OLOCK DATA 072850 WLOCK DATA ULKDATI 072860 INTEGERELE(99),STA(2) 072870 COMMON / LABEL / ELE,STA 072880 UATA ELE /n Mu,nHEn.nll,n enbEn," B",n CH," Nn," On n Fu,nNE",nNAn,nM072890 3Gn,HALu e nS!ne n Pn,n So,nCLn,nAHn," Kne nCA",nSynenTIn n Vn',nCRn,unN072900 2n,nFEn.ncon enN!","CUnsuiN","GAn,nGE","ASn,nSE",n8Rn,nKRn,nRBn uS4n072910 3," "HU",nRMn,nP0n,nAGnenCDn,nINn e nbN","SBn,072920 Yu,nZH"e"NDu,HMon,"Tk",n,LAn,"CE",nPRn enTEn," In,"AE" nCSnengA" e nNDu,"PMa,"SMn,"Eun,nGDH,n072930 n wne nRE",nOSn nIRn,uP072940 gTHuenpyn.nMone nER n , n Trin , n f 8n ,"LU", nMFu , nTen e 6T","Aun nNO",nTLu,nPB","HIn.nP0n,"ATo e nRN",nF$ ,HRAn,nACH,uTM",nPA022950 7",n Un.nNPu,uPUn e nAMnenCMa,Hahn,nCFn,"ESn/ 072960 DATA STA/" n onM n/ 072970 3-55

END

• DECK NALF, 022980 072990 SUBROUTINE HALF (A,1) 073000

.C SUBROUTINE HALF CONVENTS DALF-LIFE TO DECAY CuNSTANT (1/SEC) 073010 t; DIHENSION C(9) 073020 t e;~

DATA C/6.9315E-01,1.1552E-02,1.9254E-04,8 0229E-06,2.1965E-00,0,0,073030 1 2.1905E111,2.1Y65E,14,2.1965E 1T/ 073040 lW" IF(A.97.0 0) GO TO 19 023050 IF(1.Eo.6) GO To 20 023060 2

A=9.99 023070 8ETURb 073040 10 8sC(!)/A 053090 RETURN 073100 20 A=0 0 073110 RETURb OE3120 END 023130

• 0ECK IdOAH 073140

'SUdROUTINE NOAH (NUCLIaNAMp) OP3150 C SUdROUTINE NOAH CONVEdTS blX DIGIT IDENTItIER TO ALPHAMERIC SyMBOLOP3160

' INTEGERNAME(3) 073170

-INTEGERELE(99),STA(2) 023180 COMMON / LABEL / ELE.STA 023100 IS= HOD (NUCLI,103 1 023200 NZ =NUCLI/10000 073210 MW=NUCLI/10-NZ *1000 073220 NAME(1)= ELE (NZ) OP3230 NAME(2)=Hw 073240 NAME(3)sSTA(IS) 073250 RETURN 073260 END 073270 3-56 1

CHAPTER 4. DATA FOR RADI0 ACTIVE SOURCE TERM CALCULATIONS FOR PRESSURIZED WATER REACTORS (PWR's)

~

This chapter lists the information needed to generate source terms for PWR's.- The information is provided by the applicant and is consistent with the contents for the Safety Analysis Report (SAR) and the Environmental Report (ER) of the proposed pressurized water reactor. This information constitutes the basic data required in calculating the releases of radioactive material in liquid and gaseous effluents (the source terms).

All data are on a per-reactor basis.

4.1 GENERAL

1. -The maximum core thermal power (MWt) evaluated for safety considerations in the SAR.

Note: All the information required in calculating the releases should be adjusted to this power level.

2. The quantity of tritium released in liquid and gaseous effluents

-(Ci/yr per reactor).

4.2 PRIMARY SYSTEM

1. The total mass (lb) of coolant in the primary system, excluding the pressurizer and primary coolant purification system, at full power.

2. The average primary system letdown rate (gal / min) to the primary coolant purification system.

3. The average flow rate (gal / min) through the primary coolant purification system cation demineralizers.

Note: The letdown rate should include the fraction of time the cation demineralizers are in service.

e

4. The average shim bleed flow rate (gal / min).

4.3 SECONDARY SYSTEM'

l. The number and type of steam generators and the carryover factor used in the evaluation for iodine and nonvolatiles.

2. The total steam flow rate (lb/hr) in the secondary system.

3. The mass of liquid in each steam generator (1b) at full power.

4. The primary-to-secondary system leakage rate (lb/ day) used in'the evaluation.

4-1 o

.- . m .

! (

,a

, L51 .Descrihtion,ofthesteamganerator.blowdownpurificationsystem.-

'.The average. steam. generator blowdown rate (lb/hr) used in the- ..

o- >

evaluation.: '.

6. - The fraction.'of the steam generator feedwater processed through.

,the condensate.demineralizers'.and'the DF's used in the evaluation

~

for^ the condensate demineralizer system.

~

7.. -Condensate'demineralizers

a. Average flow ' rate (1b/hr);,

b. .

Demineralizer type (deep bed or. powdered resin);

^ -

c. Number and size (ft3 ) of demineralizers;

d. - Regeneration frequency;

- e.- ' Indication whether ultrasonic resin cleaning is used and the' waste. liquid volume associated with its use; and

f. ' Regenerant volume (gal / event) and activity.

1 e4.4 LIQUID WASTE PROCESSING SYSTEMS

~

' 1. - .For each . liquid waste processing system, including the shim z bleed,-steam generator blowdown, and detergent waste processing

systems, provide in tabular form the following information:.

a. Sources, flow rates (gal / day), and. expected activities

.(f raction of' primary coolant activity) for:all inputs to each system.

.# b. Holdup times associated with' collection, processing, and. discharge of.all liquid streams.

~

c. Capacities of all tanks (gal) and processing equipment (gal / day) considered in ~ calculating holdup times.

d. . Decontamination factors for each processing step.

+

e.- Fraction of each processing stream expected to be

j. discharged over the life of the plant.

.f. . For. demineralizer regeneration, provide time between reger.erations, regenerant volumes and activities, treatment of regenerants, and fraction of 'regenerant-

. discharged. Include parameters used in making these.

E~ determinations.

g. Liquid source term by radionuclide in C1/yr for normal operation, including anticipated operational occurrences.

4-2 E'

2. Provide piping and instrumentation diagrams (P&ID's) and process flow diagrams for the liquid radwaste systems along with all other systems influencing the source term calculations.

4.5 GASEOUS WASTE PROCESSING SYSTEM For the waste gas processing system, provide the following:

1. The method of stripping gases from the primary coolant, the volumes (f 3t /yr) of gases stripped from the primary coolant, the bases for these volumes.

2. Description of the process used to hold up gases stripped from the primary system during normal operations and reactor shutdown.

If pressurized storage tanks are used, include a process flow diagram of the system indicating the capacities (ft 3), number, and design and operating storage pressures for the storage tanks.

3. Describe the normal operation of the system, e.g., number of tanks held in reserve for back-to-back shutdown, fill time for tanks. Indicate the minimum holdup time used in the evaluation and the basis for this number.

4.- 'If HEPA filters are used downstream of the pressurized storage tanks, provide the decontamination factor used in the evaluation.

5. If a charcoal delay system is used, describe this system and indicate the minimum holdup times for each radionuclide considered in the evaluation. List3 all parameters, including mass of charcoal (Ib), flow rate ft / min), operating and dew point temperatures, and the dynamic adsorption coefficients for Xe and Kr used in calculating holdup times.

6. Provide piping and instrumentation diagrams (P&ID's) and process flow diagrams for the gaseous radwaste systems along with other systems influencing the source term calculations.

4.6 VENTILATION AND EXHAUST SYSTEMS For each building housing systems that contain radioactive materials, the steam generator blowdown system vent exhaust, gaseous waste processing system vent, and the main condenser air removal system, provide the following:

1. Provisions incorporated to reduce radioactivity releases through the ventilation or exhaust systems.

.2. Decontamination factors assumed and the bases (include charcoal adsorbers, depth of charcoal beds, HEPA filters, and mechanical devices).

3. Release rates for radioiodine, noble gases, and radioactive particulates (Ci/yr), radioactive particulate size distribution,

, and the bases.

4-3

t

4. Release point description, including height above grade, height above relative location to adjacent structures, relative temperature difference between gaseous effluents and ambient air, flow rate, velocity, and size and shape of flow orifice.

5. For the containment building, the building free volume (ft3) ,

and a thorough description of the internal recirculation system

. (if provided), including the recirculation rate, charcoal bed depth, operating time assumed, and mixing efficiency.

Indicated the expected purge and venting frequencies and duration and' continuous purge rate (if used).

4-4

APPENDIX A LIQUID SOURCE TERM CALCULATIONAL PROCEDURE FOR REGENERANT WASTES FROM DEMINERALIZERS OTHER THAN CONDENSATE DEMINERALIZERS Often in PWR radwaste systems, demineralizers other than the condensate demineralizers may undergo regeneration, for example, the radwaste demineralizer in the dirty waste system. The PWR-GALE Code can calculate the liquid effluent resulting from periodic regeneration of non-condensate demineralizers by following the procedure outlined

-below.

1. Input to Cards 1-11 and Cards 27-42 A separate computer run for calculating the regeneration waste effluent from non-condensate demineralizers is required. Cards 1-11 should be filled out as indicated for the specific plant in Sections 1.5.2.1 through 1.5.2.11 of this report. Also Cards 27 through 41 may be left blank-(except that values of 1.0 must be entered for Card 28 entries). Card 42 should be left blank.

2. Input -to Cards 12-26 The only liquid source term data cards completed (Cards 12-26) should be the three card sets used in the input data for the stream in which the demineralizer to be regenerated is located. The remaining card sets should have a zero entered for the input flow rate.

a. Input Flow and Activity (Card 12,15,18, 21 or 24)

The input flow rate and input activity should be the average daily input flow rate and input activity processed through the demineralizer to be regenerated. For example, if the demineralizer to be regenerated is used to process a shim bleed waste stream, the total input flow rate might be 1440 gallons per day.

Note that it is not the flow rate and activity which is due to the regenerant waste which is entered, it is the normal flow rate and activity through the compenent to be regenerated which is entered.

b. Regeneration Frequency (Card 14,17, 20, 23 or 26)

Enter the time between regenerations in days as the " collection time." If a regeneration frequency is stated by the applicant, it may be used; otherwise the following frequency may be used:

A-1

c  %

~

TABLE A-1

' Demineralizer Service Regeneration Frequency e Primary Coolant Letdown 180 days Boron Recovery System 180 days Equipment Drain Wastes-

• Floor Drain Wastes

• Steam Generator Blowdown 90 days a

• Regeneration frequency is calculated by dividing the waste quantity (gallons) by the waste flow rate in gallons per day. The waste 3

quantity is 25000 gal /ft times the volume in ft3 of resin for 3

equipment drain waste and 2000 gal3/ft times the volume in ft of

-resin for floor drain waste. The calculated values of 25,000 and

%000. gal /ft3 of resin for the waste are based on 12,000 g 3

CaC0 ion exchange capacity per ft of resin and 5 pmho/cm and 50 pMho/cm average conductivity for equipment and floor. drain liquid wastes.

By inputting the normal flow rate and activity in Item a and the. regeneration frequency as the collection time in Item b the PWR-GALE Code will accumulate all of the activity processed through the demineralizer during its normal operation and decay the activity as a function of the time over which it was collected.

c. Process Time and Fraction Discharged Use the same " process time" and " fraction discharged" as indicated for the stream in which the regeneration wastes are

! processed as indicated in Section 1.5.2.12.4 of this document.

d. Decontamination Factors (Card 13,16,19, 22 or 25)

The decontamination factors entered should consider radionuclide removal by the equipment used to process the regenerant wastes using the normal source term procedures of 1.5.2.12.2. In addition, the decontamination factors entered should be used to adjust the source term for the fraction of the activity in the process stream flowing through the demineralizer during normal operation which-is not removed by the demineralizer.

A-2 i

e. Sample Case A demineralizer is used to process shim bleed waste and is to be regenerated. The normal flow rate for the demineralizer is 1440 gpd and the activity is calculated in the PWR-GALE Code.

The regenerant wastes will be processed through an evaporator and discharged.

Fill in the Cards 12-14 in the following manner:

Card 12 Spaces 18-41 enter - shim bleed demin regen Spaces 42-49 enter - 1440.0 Card 13 The wastes will be processed through an evaporator which will provide the following DF's according to Table 1-4 of Section 1.5.2.12.2.

I - 10 23 Cs, Rb - 10 Others - 10 While in operation, referring to Table 1-4 of Section 1.5.2.12.2 demineralizer DF's are:

I - 10 Cs, Rb - 2 Others - 10 Therefore, for "I" and "Others," 90% of the activity processed through the demineralizer is removed by the resins and no adjustment is needed. Only 50% of the Cs and Rb in the waste stream is removed by the resins, however, so the DF entered for Cs should be adjusted. Thus, the DF's entered on Card 13 would be:

I - 100.0 Cs, Rb - 2000.0 Others - 1000.0 Card 14 Spaces 29-33 " Collection Time." Using the value from Table A-1 of 180 days for the regeneration frequency Enter 180.0 days in spaces 29-33.

A-3

, , .n .- . . :. .w , , ; . , - , m ,. . n. w ., 7. - , . .n -

.v?.-- & . . - . . . : - a . . - . . . .. : ..

=..y....

f.j L. ' ., ;y a,[  ;! , l *.s Q .. llsn%

9 . . ',

.,.s t i.; Use the same " Process time" and " fraction discharged" as is ",, -(f -=

. . indicated for the stream in which the regeneration wastes are 3.. s.

processed as indicated in Section 1.5.2.12.4 of this report.

~

.. y, .; , ;,5 Note: If there is more than one stream for which non-condensate 9 E. 7'

.' regenerant demineralizer is used, follow the same procedures P f.i ';-

fc explained under item A2 for the other stream or streams. y g .; .

,c . -.. ._ :

3. Components in Service 7 ; '. -

'T a. If the waste is processed through a component other than a  %-($;.

? regenerable demineralizer prior to processing by the regenerable 3.j.. 3 /

demineralizer, the activity in the steam entering the demineralizer 0. .M V' will be less than the activity entered as described above. To J'V^-

C compensate for this difference, the DF's for the regenerant d l,' 4:

i

.D waste calculation should be adjusted in a manner similar to that described above. The product of the DF's should be used.

N-i gy 1.

.) -

b. If two regenerable demineralizers are used in series, follow i$.; n i

'g.'- -

'7 the procedure in a above. Adjust the DF for nuclides removed W( .'

llp from the waste stream, by using the product of the DF's for f.g% y

>v two demineralizers in series, i.e., consider the two demineralizers  : t .'

4 as one larger demineralizer. Z- '

n. (~ Q:

4. Use of Computer Calculated Result -f .,

's Combine the values printed out in the individual liquid source term kg, -@j.

..' columns for the system in which the demineralizer is being regenerated .

f +

J. (not the adjusted total value) with the normal liquid source term "j

run values. Do not use the adjusted total value from the right hand

. .li d column since the source term run to which the regenerant waste run @? .'

will be added has already been adjusted. t5 Q_ iX l-  : __. *

, ,' .0 ? . ' ;' :.

,/ l.Y-[..-

h .

'i 3-

n

$w 5.F: ?:.

~., f :

I. ,..L y

y? $

T- . i

.! #p/ .'j, \ , #

l h'2 b Y.

, ..f '

4.5

l .d: - .M.[r -

.. .l ;. _ 7 h.y' ' k . .{

}g .;.t g .{

< ~ I

?6 d{g g' e

,,#y

( . .y

1. h, >-

.; i ' s . .., +

tp-  ;' .

,.y M- A-4 , , ; p.,

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3 REFERENCES

1. American National Stan'd ards Source Term Specification, ANSI N237-1976, i American National Standards Institute.

2. Regulatory Guide 1.140, " Design, Testing, and Maintenance Criteria lfor Normal Ventilation Exhaust. System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants," March 1978, Office of. Standards Development, U. S. Nuclear Regulatory Commission.

, 3. . Letter from H. Till, Electric Power Research Institute, to J. Collins, NRC, March 8, 1978.

4. NUREG/CR-0140, "In-Plant Source Term Measurements at Ft. Calhoun Station,-Unit 1," July 1978.

^

5.- NUREG/CR-0715, "In-Plant Source -Term Measurements at Zion Station,"

May , 1979.

'6. NUREG/CR-1629, "In-Plant-Source Term Measurements at Turkey Point -

Station - Unit 3 and 4," September,1980.

7. Electric Power Research Institute Report EPRI NP-939, " Sources of Radioiodine at Pressurized Water Reactors," November,1978.

8. Westinghouse' Electric Corporation, WCAP-8253, " Source Term Data for Westinghouse Pressurized Water Reactors," July,1975.

-9. Letter from T. M. Anderson, Westinghouse Electric Corp. to J. Collins, NRC, April 17,1979. '

10. Combustion Engineering, CENPD-67, Rev.1, " Iodine . Decontamination

. ~ Factors During PWR Steam Generation and Steam Venting," J. A. Martucci, November, 1974.

11. Combustion Engineering, CENPD-67, Addendum IP, " Iodine Decontamination Factors During PWR Steam Generation and Steam Venting," November 1974.

12. Westinghouse Electric Corporation, WCAP-8215, " Steam Side Iodine Transport Study at Point Beach Unit No. ~1 of Wisconsin Electric Power Company," October 1973.

13. General Electric Company, Figure 5 of Draft Report, " Fission Product Transport Measurements at Brunswick - 2," C. Lin and H. Kenitzer (tobepublished).

14. . NUREG-0017, " Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE Code)," April 1976.-

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15. NUREG-0016, Rev.1, " Calculation of Releases of Radioactive Materials 2 in Gaseous and Liquid Effluents from Boiling ifater Reactors (BWR- _,

S GALE Code)," January 1979.

16. Pilgrim 2 Preliminary Safety Analysis Report (PSAR) (Docket No. j 50-471), Appendix 11E, Amendment 12, November 1974. J

17. Letter from K. Seyfrit, Technical Assistance Branch, AEC, Regulatory  ;

2 Operations, to G. Lainas, Containment Systems Branch, AEC, Licensing "PWR Purging and Venting Experience," September 3,1974. i i

18. NUREG-75/087, "U. S. Nuclear Regulatory Commission Standard Review _

Plan," Section 6.2.4, " Containment Isolation System," Rev.1, a November 1978. .

19. Rochester Gas and Electric Corporation, " Radioactivity in the b Containment Building Atmosphere of Ginna Station," A. R. Piccot, 1971.

4

20. Nuclear Containment Systems, Incorporated, NCS-1101, " Dynamic j Adsorption Coefficient and Its Application for Krypton-Xenon Delay Bed Design," J. L. Kovach, Draf t, November 1971. ,

=

21. L. R. Michaels and N. R. Horton, " Improved BWR Offgas Systems,"

12th Air Cleaning Conference, San Jose, California, August 1972.  ;

22. W. E. Browning et al., " Removal of Fission Product Gases from Reactor Offgas Streams by Adsorption," 0RNL Central Files Number 59-6-47, -

June 11,1959. _

23. H. J. Schroeder et al., "Offgas Facility at the Gundremmingen Nuclear ]

Power Plant," Journal for Nuclear Engineers and Scientists," No. 5, ]

May 1971, pp. 205-213. 2

24. Letter from Kerndraftwerk Lingen GMBH to Peter Lang, North American -

Carbon, " Gas Delay System at KWL," December 30, 1970.  :

~

25. General Electric Company, NED0-10751, " Experimental and Operational Confirmation of Of fgas System Design Parameters," C. W. Miller,  ; "

proprietary report, October 1972.

26. Letter from J. L. Kovach, Nuclear Containment Systems, Inc., to  :

V. Benaroya, AEC, " Gas Delay Systems," December 1,1971. _

27. D. P. Siegwarth et al., " Measurement of Dynamic Adsorption Coeffi-  ;

cients for Noble Gases on Activated Carbon," 12th Air Cleaning -

Conference, August 1972. _

28. General Electric Co., NED0-20ll6, " Experimental and Operational Confirmation of Offgas System Design Parameters," C. W. Miller, i October 1973. ,

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'29.: ANSI /ANS 55.6-1979, "American N' a tional Standard Liquid Radioactive

-Waste Processing System for Light Water Reactor Plants," American -

National Standards Institute, April 1979.

30. NWT 133-1, "0TSG Secondary Water Chemistry Study," Nuclear. Water and Waste-Technology, March 1978.

31. - ' NWT 133-2,. "0TSG Secondary Water Chemistry Study," Nuclear Water

and Waste Technology, June 1978.

.32.: NUREG/CR-0143, "The Use of Ion Exchange to Treat Radioactive Liquids in Light-Water-Cooled Nuclear Power Plants," August 1978.

33. NUREG/CR-0142, "The Use of Evaporation to Treat Radioactive Liquids in

. Light-Water-Cooled Nuclear Power Plants," September,1978.-

34. - NUREG/CR-0141, "The Use of Filtration to Treat Radioactive Liquids in Light-Water-Cooled Nuclear Power Plants," September 1978.

35. - A "

Stud' y of Reverse Osmosis Applicability to Light Water Reactor Radwaste Processing," J. Markind, T. Van Tran, November 1978.

36. W. R. Greenway et al., " Treatment of Radioactive Steam Generator-Blowdown," 33rd Annual Meeting, International Water Conference of the Engineers' Society-of Western Pennsylvania, October 24-26, 1972.

37. C. Kunz et al., "C-14 Gaseous Effluent From Pressurized Water Reactors," CONF-741018, Symposium on Population Exposures, Proceedings of the Eighth Midyear Topical Symposium of Health Physics Society, Knoxville, Tennessee October 21-24,1974, pp. 229-234.

38. Westinghouse Electric Corporation, WCAP-7702, " Interim Report on Study of Iodine Transport in PWR Steam Systems," May,1971.

39. Letter from T. M. Anderson, Westinghouse Electric Corp., to R. Bangart.

NRC, November 8,1979.

'40. Letter from J. J. Barton, Metropolitan Edison Co., to J. Collins, NRC, December 4,1979.

41. U.S.E.P. A., EPA-520/5-76-003, " Radiological Surveillance Studies at the Oyster Creek BWR Nuclear Generating Station," June 1976.

42. "In-Plant Source Term Measurements at Prairie Island Nuclear Generating Station." To be published as a NUREG document.

43. "In-Plant Source Term Measurements at Rancho Seco Station," NUREG/

CR-2348, October 1981.

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pressurized water reactors. e PWR-GALE , de is a computerized mathematical model for -8 calculating the releases of 'dioactive mat tal in gaseous and liquid effluents --

(i.e., the gaseous and liqu source terms).\ The U.S. Nuclear Regulatory Commission 7 uses the PWR-GALE Code to termine conformanc with the requirements of Appendix I to y 10 CFR Part 50. -

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