Calculation of 

Releases of Radioactive Materials 

in Gaseous and Liquid Effluents 

from Pressurized Water Reactors 

PWR-GALE Code 

Manuscript Completed: March 1985 

Date Published: April 1985 

T. Chandrasekaran, J. Y. Lee, C. A. Willis 

Division of Systems Integration 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, D.C. 20555 

I' 

, r I, 

NUREG-0017 

Rev. 1 

NRC FORM 335 

{2·B41 

U.S. NUCLEAR REGULATORY COMMISSION ,. REPORT NUMBER (Auignodby TlDC, .dd Vol. No. ifony) 

NRCM 1102, 

3201,3202 BIBLIOGRAPHIC DATA SHEET NUREG-00l7, Rev. 1 

SEE INSTRUCTIONS ON THE REVERSE 

2. TITLE AND SUBTITLE 

Calculation of Releases of Radioactive Materials in 

Gaseous and Liquid Effluents From Pressurized Water 

Reactors (PWR-GALE Code) 

5. AUTHOR lSI 

T. Chandrasekaran, J.Y. Lee, C.A. Willis 7. PERFORMING ORGANIZATION NAME AND MAILING AODRESS (/nciudeZ,p Code) 

Division of Systems Integration 

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 

Washinqton DC 20555 10. SPONSORING ORGANIZATION NAME AND MAILING ADORESS (lnc/udeZip Code) 

Same as 7. above 

12 SUPPLEMENTAR Y NOTES 

13. ABSTRACT (200 word, or !ess) 

3. LEAVE BLANK 

4. DATE REPORT COMPLETEO 

MONTH YEAR 

March 1985 6. DATE REPORT ISSUED 

MONTH YEAR 

April 1985 B. PROJECT/TASKIWORK UNIT NUMBER 

9. FIN OR GRANT NUMBER 

lla. TYPE OF REPORT 

b. PERIOD COVERED (lnc/usive det .. ) 

This report revises the original issuance of NUREG-00l7, "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. 

14 OOCUMENT ANALYSIS •. KEYWORDS/OESCRIPTORS 

Radioactive Effluent 

Radioactive Waste 

Radioactive Materials 

b. IDENTIFIERS/OPEN·ENDED TERMS 

~u.s. GOVERNMENT PRINTING OFFICE: 1985-461-721:20097 

15. AVAILABILITY 

STATEMENT 

Unl imited 16. SECUR ITY CLASSI F ICATION 

(Thls~tlJ 

Unclassified 

(ThiS rtlporrJ 

Unclassified 

17. NUMBER OF PAGES 

lB PRICE 

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 

ABSTRACT • • • • • • • • • . . . . . . . . . • iii ACKNOWLEDGEMENTS ••••• · xi 

EXECUTIVE SUMMARY • • • • • • xi i i 

CHAPTER 1. PWR-GALE CODE 

1.1 

1.2 

Introduction 

Definitions •• 

• • • 1-1 

• • 1-3 

1.3 Gaseous Source Terms •• 1-6 

1 .4 Li qui d Sou rce Terms • • • 1 - 7 

1.5 Instructions for Completing PWR-GALE Code Input Data 

Cards. • • • • • • • • • • • • • • • • • • • • • • 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 

2.2.2 

2.2.3 

2.2.4 

2.2.5 

2.2.6 

2.2.7 

2.2.8 

2.2.9 

2.2.10 

2.2.11 

2.2.12 

2.2.13 

2.2. 14 

2.2.15 

2.2.16 

Thermal Power Level • • • • • ••• 

Plant Capacity Factor ••••••••••• 

Radionuclide Concentrations in the Primary and 

• 2-1 

2-1 

Secondary Coolant •••••••••••••• 2-2 

Iodine Releases from Building Ventilation 

Systems ...... ............. 2-33 

Radioactive Particulates Released in Gaseous 

Effluents •••••••••••••••••• 2-40 

Noble Gas Releases from Building Ventilation 

Systems ••• •••••••••••••••• 2-42 

Steam Generator Blowdown Flash Tank Vent •• 2-48 

Iodine Releases from Main Condenser Air Ejector 

Exhaust •••••••• • ••••••••• 

Containment Purge Frequency •••••••••• 

Containment Internal Cleanup System •••••• 

Radioiodine Removal Efficiencies for Charcoal 

Adsorbers and Particulate Removal 

• 2-48 

• 2-51 

• 2-55 

Effi ci enci es for HEPA Fi 1 ters •••••••• 2-56 

Waste Gas System Input Flow to Pressurized 

Storage Tanks •••• • • • • • 

Holdup Times for Charcoal Delay System 

Liquid Waste Inputs ••••••••• 

Detergent Waste • • • • • • • • • • • • • 

Chemical Wastes from Regeneration of 

Condensate Deminerallzers •• 

• • 2-57 

• • 2-60 

• • 2-61 

• • 2-61 

• 2-68 

Preceding page blank 

v 
Page 

2.2.17 Tritium Releases • • • • • • • • • 2-68 

2.2.18 Decontamination Factors for Deminera1izers ••• 2-77 

2.2.19 Decontamination Factors for Evaporators ••••• 2-79 

2.2.20 Decontamination Factors for Liquid Radwaste 

Filters • • • • • • • • • • • • •• • 2-80 

2.2.21 Decontamination Factors for Reverse Osmosis • 2-80 

2.2.22 Guideline for Calculating Liquid Waste Holdup 

Times •••••••••••••••••••• 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 

3.4 Listing of PWR-GALE Code 

3.4. 1 

3.4.2 

Nuclear Data Library 

FORTRAN Program Listing . . . . . . . 

CHAPTER 4. DATA FOR RADIOACTIVE SOURCE TERM CALCULATIONS 

FOR PRESSURIZED WATER REACTORS 

4.1 General 

4.2 Primary System . . . 4.3 Secondary System 

4.4 Liquid Waste Processing Systems. 

4.5 Gaseous Waste Processing System 

4.6 Ventilation and Exhaust Systems 

APPENDIX A 

• • 3-5 

• 3-5 

• 3-5 

• 3-15 

• • 4-1 

• • 4-1 

• • • 4- 1 

4-2 

••• 4-3 

• 4-3 

Liquid Source Term Calculational Procedures for Regenerant 

Wastes from Demineralizers Other Than Condensate 

Demineralizers • • . • • • • • • • . • • • • • . • • • •• A-l 

REFERENCES R-l 

vi 
LIST OF TABLES 

Table 

1-1 Radioiodine Releases from Building Ventilation Systems 

Prior to Treatment •••••••••••••••••• 1-9 

1-2 

1-3 

1-4 

Radioactive Particulate Releases from Building 

Ventilation Systems Prior to Treatment 

PWR Liquid Wastes ••• 

Decontamination Factors for PWR Liquid Waste 

Treatment Systems • • • • • • • • • • • • • . . . . 1-5 Assigned Removal Efficiencies for Charcoal Adsorbers 

• • 1-10 

• 1-18 

• • 1- 21 

for Radioiodine Removal •••••• 1-29 

2-1 Plant Capacity Factors at Operating PWRls • • 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 PWRls with U-Tube Steam 

Generators •••••••••••••••••••••• 2-12 

2-8 Adjustment Factors for PWRls with Once-Through 

Steam Generators •••••••••••••••• 2-13 

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

Coolant Concentrations in PWRls ••• 

2-10 Summary of Radionuclide Primary Coolant Concentrations 

• 2-21 

in PWRls • • • • • • • • • • 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 

2-13 

2-14 

2-15 

2-16 

2-17 

Annual Iodine Normalized Releases from Containment 

Ventilation Systems ••••••••••••••• 

Annual Iodine Normalized Releases from Auxiliary Bldg. 

Vent il at ion Systems · · · · · · · · · · · · · · · · 

Annual Iodine Normalized Releases from Refueling Area 

Ventilation Systems · · · · · · · · · · · · · · · · 

· 

· 

Annual Iodine Normalized Releases from Turbine Building 

Ventilation Systems · · · · · · · · · · · · · · · · · 

• 2-34 

· 2-35 

· 2-36 

· 2-37 

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 

Pool Area • • • • • • • • • • • • • • • • • • • • • • • 2-45 

2-21 Measured Releases Upstream Filters - Waste Gas 

System • • • • • • • • • • • • • • • 2-46 

2-22 Annual Iodine Normalized Releases from Main Condenser 

Air 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 

2-26 

2-27 

2-28 

2-29 

Waste Gas System Input Flow to Pressurized Storage 

Tanks for PWR's with Recombiners ••••• 

PWR Liquid Wastes 

Calculated Annual Release of Radioactive Material in 

Untreated Detergent Waste • • • • • • • • • 

Radionuclide Distribution of Detergent Waste 

Tritium Release Data from Operating PWR's with 

Zircaloy-Clad Fuels ••••••• 

2-30 Tritium Release Rate from Operating PWR's as a Function 

2-59 

2-65 

• 2-67 

• 2-69 

• 2-71 

of Number of Years of Operation. • ••••••• 2-75 

viii 
Table 

2-31 

2-32 

2-33 

2-34 

2-35 

2-36 

2-37 

2-38 

2-39 

2-40 

Tritium Release Rate from Operating PWR's - Percent 

of Total Tritium Released in Liquid Effluents •• 

Distribution of Tritium Release in Gaseous Effl uents 

Reverse Osmosis Decontamination Factors, 

Ginna Station . . . . . . . . . . . . . · . · . · 

Reverse Osmosis Decontamination Factors, 

Point Beach . . . . . . . . . . . . . . · . · . · 

. 

. Reverse Osmosis Decontamination Factors, 

H. B. Robinson No.2 Station ••••• · . . . . . Expected Reverse Osmosis Decontamination Factors for 

Specific Nuclides ••••••••••••••• 

Frequency and Extent of Unplanned Liquid Radwaste 

Releases from Operating Plants 

Carbon-14 Release Data from Operating PWR's •• 

Distribution of Carbon-14 Released in Gaseous 

• 2-76 

· · 2-78 

· · 2-82 

· · 2-83 

2-84 

2-85 

2-88 

• 2-91 

Effluents. • • ••••••••• · . · . . . 2-92 

Summary of Argon-4l Releases for Operating PWR's 

for 1973-1978 ••••••••••••••••• 

;x 

2-94 
Fi gu re 

2-1 

2-2 

2-3 

LIST OF FIGURES 

Removal Paths for Pressurized Water Reactor with 

U-Tube Steam Generators •••••••••••• 

Removal Paths for Pressurized Water Reactor with 

Once-Through Steam Generators •••••• 

Krypton and Xenon K Values as a Function of 

. . . . 

Reciprocal Temperature •••••• . . . . . 

2-14 

2-15 

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 

3-4 

Program Listing for Gaseous Determination 

Program Listing for Liquid Determination 

x 

•• 3-16 

3-26 
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 

Commission were the principal contributors. J. W. Mandler and F. Y. Tsang 

of EG&G 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 

Commission. 

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 IIsource term,lI* 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 ~ffluents) Code is a computerized mathematical model for calculatTng the 

releases of radioactive material in gaseous and liquid effluents (i.e., the 

gaseous and liquid source terms) from pressurized water reactors. The calculations are based on data generated from operating reactors, field and laboratory 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 Commission 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 experience where data were 

available. 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 values listed. A list of references used in developing the parameters is also included. The source term parameters used are believed to provide 

a realistic assessment of reactor and radwaste system operation. 

* IiSource termli as discussed in this report differs from lIaccident source 

term," which deals with potential releases resulting from nuclear reactor 

accidents. 

Preceding page blank xiii 
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 

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 I NTRODUCT ION 

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-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 average quantity of radioactive material released to the environment 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 deminera1izers 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 radio1ysis. 

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 

leaking from the primary coolant concentrate in the liquid phase in the 

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

generator b1owdown rate and condensate demineralizer flow rate. 

Since there is no liquid reservoir in a once-through steam generator, 

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 generator 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 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 ent~ 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: 

Activation Gases: The gases (including oxygen, nitrogen, and argon) 

that become radioactive as a result of irradiation in the core. 

Anticipated Operational Occurrences: Unplanned releases of radioactive 

materials from miscellaneous actions such as equipment failure, operator 

error, administrative error, that are not of consequence to be considered 

an accident. 

Chemical Waste Steam: Normally liquids that contain relatively high 

concentrations of decontaminants, regenerants, or chemical compounds 

other than detergents. These liquids originate primarily from resin 

regenerant and laboratory wastes. 

Clean Waste System: Normally tritiated, nonaerated, low-conductivity 

liquids consisting primarily of liquid waste collected from equipment 

leaks and drains and certain valve and pump seal leakoffs. These liquids 

originate from systems containing primary coolant and are normally 

reused as primary coolant makeup water. 

1-3 
Decontamination Factor OF: 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 

Effective Full Power Days = Integrated Thermal Power = 

Licensed Power Level 

. P. T. 

1 1 1 

Ptotal 

P. 

1 

Ptotal 

T. 

, 

is the ith power level, in MWt; 

is the licensed power level, in MWt; and 

is the time of operation at power level Pi' 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. 

where 

Gaseous Effluent Stream: Processed gaseous wastes containing radioactive 

materials resulting from the operation of a nuclear power reactor. 

Liguid Effluent Stream: Processed liquid wastes containing radioactive 

materials resulting from the operation of a nuclear power reactor. 

Partition Coefficient (PC): The ratio of the concentration of a 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 

powe r 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. 1 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 (Sr, 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 radioactive 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 lowlevel radioactivity primarily resulting from secandary system leakage, 

steam trap drains, sampling system drainage, and maintenance and waste 

drains. 

1.3 GASEOUS SOURCE TERMS 

The following sources are considered in calculating the releases of 

radioactive materials (noble gases, radioactive particulates, carbon-14, 

tritium, argon-41, and iodine) in gaseous effluents from normal operation, 

including anticipated operational occurrences: 

1. Waste gas processing system; 

2. Steam generator blowdown system; 

3. Condenser air ejector exhaust; 

4. Containment purge exhaust; 

5. Ventilation exhaust air from the auxiliary, and turbine 

buildings, and the spent fuel pool area; and 

6. Steam leakage from the secondary system. 

The releases of radioactive materials in gaseous effluents from the 

followiU g sources are calculated to be less than 1 Ci/yr of noble gases 

and 10 Ci/yr of iodine-131. Therefore, the following releases are 

considered negligible: 

1. Steam releases due to steam dumps to the atmosphere and lowpower 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~J1t during 

normal operation and from degassing the primary coolant for two cold 

shutdowns per year. For plants equipped with steam generator blowdown 

systems, the model considers iodine present in gases leaving the system 

1-6 
vent. The PWR-GALE Code calculates the release rates of noble gases and 

iodine to building atmospheres based on coolant leakage rates to buildings. 

Radioiodine releases are related to the iodine-131 coolant concentrations 

for the PWR being evaluated. Particulate release rates are based on 

measurements at operating 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 boron 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 PWR's using a deep-bed condensate demineralizer, the inventory 

of radionuclides collected on the demineralizer resins is calculated 

by considering the flow rate of condensate at main steam activity that 

1-7 
is processed through the deminera1izers and radionuc1ide removal using 

the decontamination factors given in Chapter 2. The activity on the 

condensate deminera1izer resins will also include the steam generator 

b1owdown activity if the b1owdown is recycled to the condensate deminera1izers. The radioactivity content of the deminera1izer 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 Chapter 2 of this document. 

1.5.1.3 Radioiodine 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 Tab 1 e 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 1b/day primary coolant 

leakage. Releases from the turbine building are based on 1700 1b/hr 

steam leakage. 

1-8 
TABLE 1_l t * 

RADIOIODINE RELEASES FROM BUILDING VENTILATION 

SYSTEMS PRIOR TO TREATMENT 

(Ci/yr/~Ci/g) 

Containment Auxiliary 

Building Bui1ding** 

Turbine 

Building *** 

Annual Norma1ized* Iodine 

Release Rate 

t* 

* 

** 

*** 

t 

tt 

Power Operation 8.0 x 10-4tt 0.72t 3.8 x 103 

Refueling/Maintenance 0.32** 2.59 4.2 x 102 

Outages 

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 radioiodine. It is the 

combination of the reactor water leakage rate into the building and 

the partitioning of the radioiodine between the water phase in the 

leakage and the gas phase where it is measured. For the turbine 

building the effective leak rate must consider the carryover for 

radioiodine from water to steam in the steam generator. 

To obtain the actual iodine release from these buildings in Ci/yr, 

multiply the normalized release by the iodine coolant concentration 

in ~Ci/g. 

To obtain the actual iodine release from the turbine building in 

Ci/yr, multiply the normalized release by the secondary coolant 

concentration in ~Ci/g and by the partition coefficient (NS) from 

Table 2-6. 

Includes contribution from the fuel pool area. 

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

coolant inventory of iodine and represents the effective leak 

rate for radioiodine. It is the combination of the reactor water 

leakage rate into the buildings, and the partitioning of the 

radioiodine between the water phase in the leakage and the gas 

phase where it is measured. In order to obtain the releases in 

curies/year during power operations from the containment building 

of a particular PWR, the normalized leak rates in Table 1-1, are 

multiplied 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. 

1-9 
TABLE 1-2 

RADIOACTIVE PARTICULATE RELEASES FROM BUILDING VENTILATION SYSTEMS 

PRIOR TO TREATMENT* 

(Ci/yr)/Unit 

Auxi 1 i ary Fuel Pool Waste Gas 

Nuclide Containment Building Area System 

Cr-51 9.2(-3)t 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 1.0% 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 
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 

Iodine 

Nonvolatiles 

Recirculation U-Tube 

Iodine 

Nonvolatiles 

PC 

1.0 

1.0 

0.01 

0.005 

1.5.1.9 Radioiodine 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/~Ci/g. The 

normalized release rate represents the effective release rate for radioiodine. It is the combination of the steam flow to the main condenser, 

the partitioning of radioiodine between the main condenser and the air 

ejector exhaust where it is measured, and the partition coefficient for 

radioiodine from water to steam in the steam generator. To obtain the 

actual iodine release from the main condenser air ejector exhaust in 

Ci/yr, multiply the normalized release by the secondary coolant concentration in ~Ci/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 containment atmosphere based 

on 16 hours of system operation prior to purging, an iodine removal 

efficiency for the charcoal adsorbers corresponding to Table 1-5, a 

particulate OF of 100 for HEPA filters and an internal mixing efficiency 

of 70%. 

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 ~Ci/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 Ci/yr, respectively. 

1.5.1.15 Decontamination Factors for Condensate Demineralizer 

Demi neral izer Anion 

Deep Bed 10 

Powdex 10 

Cs, Rb 

2 

2 

Other Nuclides 

10 

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 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 Demineralizers 

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 
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 

forward flow. The steam generator blowdown radioactivity is 

added to the condensate radioactivity if the blowdown is 

processed through the condensate demineralizer. 

2. All radionuclides removed from the secondary coolant by the 

demineralizer resins are removed from the resins during chemical 

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 

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 

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 

extent holdup capacity or an alternative evaporator is 

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)" 

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

discussed in Chapter 2 of this document. 

1.5.2.1 Card 1: Name of Reactor (SAR/ER) 

Enter in spaces 33-60 the name of the reactor. 

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

1.5.2.2 Ca rd 2: Therma 1 Power Level {SAR/ER} 

Enter in spaces 73-80 the maximum thermal power level (i n 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) 

Enter in spaces 73-80 the mass of coolant (in 103 lb) in the primary 

system at operating temperature and pressure. 

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 demineralizers are in service to obtain the average cation deminera1izer 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 106 1b/hr) for all 

steam generators. 

1.5.2.8 Card 8: Mass of Liguid 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 Slowdown Rate and B1owdown Treatment 

Method (SAR/ER) 

Enter in spaces 37-44 the steam generator b1owdown rate as given in 

the applicants SAR or ER. 

Enter total b1owdown rate in thousands of lb/hr in spaces 37-44. 

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

Describe the B1owdown Treatment Method as follows: 

1. Enter 0 in space 80 if the blowdown is recycled to the condensate 

system after treatment in the b1owdown system whether or not 

there are condensate demineralizers. 

2. Enter 1 in space 80 if the steam generator b1owdown is recycled 

directly to condensate system deminera1izers without prior 

treatment in the b1owdown system. 

1-14 
3. Enter 2 in space 80 if the steam generator blowdown is not 

recycled to the condensate system. 

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 

frequency 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 Deminera1izer 

(SAR/ER) 

Enter in spaces 73-80 the fraction of feedwater to the steam generator 

processed through the condensate demineralizers. If condensate deminera1izers 

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. 

2. Equipment Drain Waste, Cards 1 5-1 7. 

3. Clean Waste, Cards 18-20. 

4. Di rty 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 

Ca rds 15-17. 

1-15 
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) 

(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: 

23, 

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

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

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

The following entries are required on the third card (14, 1 7, 20, 

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 OFls as explained 

for Ca rd 13. 

1-16 
b. If the steam generator blowdown is recycled directly to 

the condensate system demineralizers without prior treatment 

in the blowdown system, enter OF of 1.0 for iodine in 

spaces 21-28, OF of 1.0 for cesium and rubidium in spaces 

34-41, and OF 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 OF's as explained for Card 13. 

3. Ca rd 26 

Complete Card 26 as explained for Card 14. 

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

1. Ca rd 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 regeneration 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. Ca rds 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 
I 

-' 

OJ 

SOURCE 

1. REACTOR CONTAINMENT 

a. Primary coolant pump seal 

leakage 

b. Primary coolant leakage, 

miscellaneous sources 

c. Primary coolant equipment 

drai ns 

2. PRIMARY COOLANT SYSTEMS 

(OUTSIDE OF CONTAINMENT) 

a. Primary coolant system 

equipment drains 

b. Spent fuel pit liner drains 

c. Primary coolant sampling 

system drai ns 

d. Auxiliary building floor 

d ra ins 

TABLE 1-3 

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) 

Deep-bed condo 

demineralizers 

with ultrasoni c 

resin cleaner 

20 

10 

500 

80 

700 

200 

200 

Deep-bed condo 

demi neral i zers 

without 

ultrasonic 

resin cleaner 

20 

10 

500 

80 

700 

200 

200 

Filterdemineralizer 

20 

10 

500 

80 

700 

200 

200 

Plant with 

blowdown treatment. Product 

not recycled to 

condenser or 

secondary coolant 

system 

20 

10 

500 

80 

700 

200 

200 

FRACTION OF 

PIHMARY COOLANT 

ACTIVITY (PCA) 

0.1 

1.67* 

0.001 

1.0 

0.001 

0.05 

0.1 
TABLE 1-3 (Continued) 

3. SECONDARY COOLANT 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 

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

I g. Turbine building floor 7200 7200 7200 7200 Calculated in <.0 d ra ins GALE Code 

4. DETERGENT AND DECONTAMINATION 

SYSTEMS 

a. On-site laundry faci 1 ity 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. 
tanks. For example, if the inlet streams A, B, and C enter the di rty 

waste collector tank at average rates and PCA as listed below, 

Stream A 1,000 gal/day at O.OlPCA 

Stream B 2,000 gal/day at 0.1 PCA 

Stream C 500 gal/day at 1 .OPCA 

the composite A, B, C activity would be calculated as follows: 

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

"Oi rty 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 OF'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. OF's are categorized by one of the following types of radionuc1ides: 

a. Halogens 

b. Cs, Rb 

c. Other Nuclides 

Note: A OF 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 OF for each inlet stream is the product of the 

individual equipment OF'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 OF. 

Table 1-4 shows the decontamination factors to be used for PWR 

systems. 

1-20 
TABLE 1-4 

DECONTAMINATION FACTORS FOR PWR LIQUID WASTE TREATMENT SYSTEMS 

TREATMENT SYSTEM 

Deminera1izer 

Mixed Bed 

Primary coolant letdown (CVCS) 

Radwa ste (H+OW) 

Anion 

DECONTAMINATION FACTOR 

Cs, Rb Other Nuclides 

2 

2 (1 0) 

Evaporator condensate polishing 5 

Bo ron recycle 

Steam generator blowdown 

Cation bed (any system) 

Anion bed (any system) 

Powdex (any system) 

Evaporators 

Miscellaneous radwaste 

Boric acid recovery 

Reverse Osmosis 

Laundry wastes 

Other liquid wastes 

Fi lters 

10 

1 02 (1 0) 

1 (1 ) 

102(10) 

10(10) 

All Nuclides 

Except Iodine 

103 

103 

2 

1 O( 10) 

10(10) 

1 (1 ) 

2( 10) 

All Nuclides 

30 

10 

10 

1 02 (1 0) 

10(10) 

1 (1 ) 

10(10) 

DF of 1 for all nuclides 

* For demineralizers in series, the DF for the second demineralizer is 

given in parentheses. 

1- 21 
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 deminera1izer; and collected for sampling. If required to meet 

discharge criteria, the contents of the waste sample (test) tank are 

processed through a mixed-bed deminera1izer for additional radionuc1ide 

removal. This example may be summarized graphically as: 

Deminera1izer 2 

Dirty waste 

collector tank - Filter - Evaporator - Deminera1izer 1 - Waste sample 

tank 

Extracting from Table 1-4 gives the following values for the example: 

Oemi ne ra 1 i ze r Oemi ne ra 1 i zer 

Fi Her Eva~orator 1 2 Product 

Iodine 1 102 5 

Cs, Rb 1 103 1 

Other Nuclides 103 10 

These values are obtained as follows: 

• 

• 

• 

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

A OF of 102 for iodine and 103 for Cs, Rb, and other nuclides 

is applied for the radwaste evaporator. 

A OF of 5 is applied for iodine, a OF of 1 for Cs, Rb and a 

5 x 

103 

104 

OF of 10 for the evaporator condensate polishing deminera1izer. 

• 

• 

A OF of 1 is applied to the second deminera1izer since this 

deminera1izer 's used is optional, and it is not used for 

normal operations. 

The product of the OF'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) 

102 

Collection time prior to processing is based on the input flow 

calculated above. Where redundant tanks are provided, assume the collection 

tank to be filled to 80% design capacity. If only one tank is provided, 

1-22 
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: 

Co 11 ect ion time (T ) = ~~,*="":'-~-;-"'--r- = 16 days c 

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 4B-53) 

Decay during processing and discharge of liquid wastes is shown 

graphically as follows: 

where 

R 

o 

Tank A - B Rb - Tank C R - Discharge Canal 

c 

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 ga 1 ; 

Rb is the equipment flow capacity of Process B, in gal/day; 

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

c 

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

gal/day. 

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

Tp = O.BA for redundant tanks, or Tp = 0.4A for a single tank 

Rb Rb 

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

in days: 

Td = °RBC for redundant tanks, or Td = °R4C for a single tank. 

c c 

1-23 
After performinq the above two calculations, calculate whether 

credit may be taken for decay during discharge by determining whether 

O.8C > Tp(Rb + Ro) for redundant tanks, or 

O.4C > Tp(Rb + Ro) for a single tank. 

If so, then 

Decay = Tp + O.5Td 

where "Decay" is the new processing and discharge time to be entered in 

spaces 48-53 of the third card for eacn input stream (Cards 14, 17, 20, 

23, 26, and 29). 

If, however, O.8C (or 0.4C, as appropriate) < T (Rb + R ), T is used - pop 

for the holdup time during processing, since Tank C may be discharged before 

Tank A has been completely processed. In this case, the Tp value should be 

entered in spaces 48-53 of the third card. 

For example, for the following input waste stream: 

FLOOR DRAINS 

1,000 GALfDA Y 

t 

FLOOR DRAIN 

TANK A 

20,000 GAL 

I 

I 

t 

FLOOR DRAIN 

TANK B 

20,000 GAL 

I 

WASTE 

DETERGENT WASTE 

540 GALIDAY 

WASTE 

SAMPLE 

TANK A 

40,000 GAL 

WASTE 

DEMIN. - 100 GAll f----- EVAPor~ATOR 

MIN 15 GALIMIN 

1-24 

WASTE 

SAMPLE 

TANK B 

40,000 GAL 

r--

I 

DISCHARGE 

PUMP 10 GAll 

MIN. 
Decay time during processing and discharge is calculated as follows: 

Process Time (T ) 

P 

= (0.8)~20,000 gal) = 0 7 d 

(15 ga1/mln)(1440 min/day) • ay 

Then, checking for decay credit, 0.8C/(Rb + Ro) = 1.45 days, which 

is greater than T ; therefore, credit is taken for (T + 0.5Td) or 1.8 days p p 

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 b10wdown 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 deminera1izer regenerations, or 

both according to the plant design. 

4. Turbine Building Floor Drain System - Releases of liquid from 

the turbine building floor drain system are calculated assuming 

no treatment prior to release. Straight decay time of 6 hours 

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.) 

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, cover gas) is calculated 

in the PWR-GALE Code. (For definition of "V", see 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: 

where 

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 days, assume the 

remaining gases are released after 90 days. 

The holdup time (Th) and fill time (Tf ) are calculated as follows: 

PV Tf = rTh = PV~n-2) 

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 

particular calculation); 

T

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

Th is the holdup time, in days; 

V is the volume of each tank, in ft3 (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 ft 3 /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 ft 3 /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 3/min; (see 1.5.2.l4.1.c, above) 

K is the dynami c adsorption coefficient, in cm3/g; 

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. 

Kr 

Xe 

DYNAMIC ABSORPTION COEFFICIENT, K (cm3jg) 

Ope rat i ng nOF 

Dew Point 45°F 

18.5 

330.0 

Operating nOF 

Dew Poi nt O°F 

25 

440 

Operating nOF 

Dew Point _40° 

70 

1160 

Operating O°F 

Dew Point _20° 

105 

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. 

3. 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 Card 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 

remove particulates or if the HEPA filters do not satisfy the 

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 radioactive particulates released from the fuel handling and auxiliary buildings, 

respectively. 

1-28 
TABLE 1-5 

ASSIGNED REMOVAL EFFICIENCIES FOR CHARCOAL ADSORBERS 

FOR RADIOIODINE REMOVAL 

Activated Carbona Bed Depth 

2 inches. Air filtration system designed to 

operate inside reactor containment 

2 inches. Air filtration system designed to 

operate outside the reactor containment and 

relative humidity is controlled at 70% 

4 inches. Air filtration system designed to 

operate outside the reactor containment and 

relative humidity is controlled at 70% 

6 inches. Air filtration system designed to 

operate outside the reactor containment and 

relative humidity is controlled to 70% 

Removal Efficienciesb 

for Radioiodine % 

90. 

70. 

90. 

99. 

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 radioiodine 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-Coo1ed Nuclear Power Plants" (Ref. 2). 

1-29 
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 

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) 

Enter the containment volume (in 106 ft3) in spaces 73-80. 

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

1. If the containment internal cleanup system uses 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 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. 

4. Enter the flow rate (in 103 ft 3

/min) through the internal cleanup 

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 hours prior to purging. 

d. A OF for the charcoal adsorber corresponding to the values in 

Table 1-5, a OF 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 
e. Continuous leakage of primary coolant during the operation of 

the internal cleanup system. 

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 

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. 

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), 

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. 

1-31 
4. Enter the continuous containment purge rate (ft3jmin) in spaces 

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 thro"gh 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. 

1.5.2.22 Card 42: Percentage of Iodine Removed by the Condenser Air 

Ejector Offgas Treatment System 

1. 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 radioiodine corresponding to the depth of 

charcoal as indicated in Table 1-5. 

2. If the offgases 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. 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 (ljDF) in 

spaces 73-80. The parameters in Chapter 2 of this document 

should be used in determining the DF for the treatment applied 

to detergent wastes. 

1-32 
CHAPTER 2. PRINCIPAL PARAMETERS USED IN PWR SOURCE TERM 

CALCULATIONS AND THEIR BASES 

2.1 INTRODUCTION 

The principal parameters used in source term calculations have been 

compiled to standardize the calculation of radioactive source terms. 

The following sections describe parameters used in the evaluation 

of radwaste treatment systems. The parameters have been derived from 

reactor operating experience where data were available. 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 

values listed. A list of references used in developing the parameters is 

also included. 

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

published in revisions to 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 considerations 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, the evaluation of the radwaste management systems 

need not be repeated when the applicant applies for a stretch power license 

at a later date. Past experience indicates that most utilities request 

approval to operate at maximum power soon after reaching commercial 

operation. 

2.2.2 PLANT CAPACITY FACTOR 

2.2.2.1 Parameter 

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

powe r days. 

2-1 
2.2.2.2 Bases 

The source term calculations are based on a plant capacity factor of 

80% averaged over the 30-year operating life of the plant, i.e., the plant 

operates at 100% power 80% of the time. The plant capacity factors 

experienced at PWR's are listed in Table 2-1 for the period 1972 through 

1977. 

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 problems and extended 

refueling outages that have contributed to the lower plant capacity factors 

will be overcome and that the plants will achieve the 80% capacity factor 

when averaged over 30 years of operation. 

2.2.3 RADIONUCLIDE CONCENTRATIONS IN THE PRIMARY AND SECONDARY COOLANT 

2.2.3.1 Parameter 

As used in the PWR-GALE Code, Tables 2-2 and 2-3 list the expected 

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

factors in Tables 2-6, 2-7, and 2-8. Figures 2-1 and 2-2 show the graphical 

relationship of the design parameters. 

2.2.3.2 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. 1) but have 

been updated based on a recent compilation of available operating data 

concerning primary coolant concentrations, steam generator tube leakage, 

and seconda~ side radionuclide behavior. Therefore, the concentration 

values in NUREG-0017, Rev. 1 differ from the ANSI N237 values. 

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 and 2-5. The values in 

Tables 2-2 and 2-3 are those determined to be representative of radionuclide 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 and the equations given in Tables 2-7 and 2-8. It is 

recognized that some systems will have design parameters that are outside 

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

adjusting the concentrations to the actual design parameters has been 

provided in Tables 2-6 through 2-8. The adjustment factors in Table 2-6 

through 2-8 are based on the following expression; 

2-2 
TABLE 2-1 

PLANT CAPACITY FACTORS AT OPERATING PWR's a 

FACILITYb 

Date of 

Commercial c Operation 1972 

86 

72 

58 

69 

78 

61 

1973 

48d 

60 

81 

67 

65 

40e 

72 

55 

51 f 17 

1974 

89 

83d 50 

76 

81 

1975 

84 

85 

73 

70 

71 

46e 

88 

76 

60 

69 

76 

71 

68 

68 

54 

83 

68 

75 

79 

68 

68 

69 

69 

73 

1976 

81 

66 

52 

78 

82d 

50 

1977 

Haddam Neck 

San Onofre 1 

R. E. Ginna 

Point Beach 1 

H. B. Robinson 2 

Palisades 

Point Beach 2 

Turkey Point 3 

Su rry 1 

Maine Yankee 

Su rry 2 

Oconee 1 

Indian Point 2 

Tu rkey Poi nt 4 

Fort Ca 1 houn 

Prairie Island 

Zi on 1 

Kewaunee 

Three Mile Island 

Oconee 2 

Zi on 2 

Oconee 3 

Arkansas 1 

Prairie Island 2 

Rancho Seco 

Calvert Cliffs 

Cook 1 

Millstone 2 

Trojan 

Indian Point 3 

Beaver Valley 1 

St. Lucie 1 

AVERAGE 

1/68 

1/68 

7/70 

12/70 

3/71 

12/71 

10/72 

12/72 

12/72 

12/72 

5/73 

7/73 

8/73 

9/73 

9/73 

12/73 

12/73 

6/74 

9/74 

9/74 

9/74 

12/74 

12/74 

12/74 

4/75 

5/75 

8/75 

12/75 

5/76 

8/76 

10/76 

12/76 

71 64 

a From monthly Operating Units Status Reports. 

d 

77 

61 d 50 

54d 

40 

54 

51 

71 

61 

36e 

49g 

69 72 

86 

75 

67 

91 e 51 

54d 

31 

64 

57 

73 

55 

75 

63 

58 

54 

64 

54 

69 

28g 

88 

75 

68 

69 

82 

62 

83 

85 

74 

78 

82 

78 

78 

77 

65 

54 

73 

62 

76 

83 

58 

77 

79d 53 

71 

71 

73 

87 

75 

65f 54 

63 

71 

72 

44e 

78 

74 

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 include the following plants 

which began commercial operation in 1977 and 1978: Calvert Cliffs 2, 

Cook 2, Crystal River 3, Davis Besse 1, Farley 1, Salem 1, North Anna 1, 

and Three Mile Island 2. 

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

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

f Not included due to extended operation at reudced power. 

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

2-3 
TABLE 2-2 

NUMERICAL VALUES - CONCENTRATIONS IN PRINCIPAL FLUID STREAMS 

OF THE REFERENCE PWR WITH U-TUBE STEAM GENERATORS 

(].lCi/g) 

Secondary Coo1ant* 

Isotope Reactor Coo1ant** ~Jater*** Steamtt ----

Noble Gases 

Kr-85m 1. 6 ( - 1 ) ttt 3.4(-8) 

Kr-85 4.3(-1) 8.9(-8) 

Kr-87 1.5(-1) 3.0(-8) 

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

Xe-131m 7.3(-1) 1.5(-7) 

Xe-133m 7 .O( -2) 1.5(-8) 

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) 7.5(-8) 7.5(-10) 

1-1 31 4.5(-2) 1.8(-6) 1.8(-8) 

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

1-133 1.4(-1) 4.8(-6) 4.8(-8) 

1-134 3.4(-1) 2.4(-6) 2.4(-8) 

1-135 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. O( -6) 1.0(-7) 

Triti um 

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

Other Nuclides 

Na-24 4.7(-2) 1.5(-6) 7.5(-9) 

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 
TABLE 2-2 (continued) 

Secondary Coo1ant* 

Isotope Reactor Coo1ant** Water*** Steam tt 

Fe-55 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) 

Sr-91 9.6(-4} 2.8(-8} 1.4(-10} 

y- 91 m 4.6(-4} 3.2(-9) 1.6(-11} 

y- 91 5.2(-6} 2.1(-10) 1.1(-12} 

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-l03 7.5(-3) 3.1(-7) 1.6(-9) 

Ru-l06 9.0( -2) 3.7(-6} 1.8(-8) 

Ag-ll0m 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-131m 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(-11) 

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 lbjday. 

** 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 10-1• 

2-5 
TABLE 2-3 

NUMERICAL VALUES - CONCENTRATIONS IN PRINCIPAL FLUID STREAMS 

OF THE REFERENCE PWR WITH ONCE-THROUGH STEAM GENERATORS 

(].ICi/g) 

Isotope Reactor Coo1ant* 

Noble Gases 

Kr-85m 1.6(-1) 

Kr-85 4.3(-1) 

Kr-87 1.5(-1) 

Kr-88 2.8(-1) 

Xe-131m 7.3(-1) 

Xe-133m 7.0(-2) 

Xe-133 2.6(0) 

Xe-135m 1.3(-1) 

Xe-135 8.5(-1) 

Xe-137 3.4(-2) 

Xe-138 1.2(-1) 

Halogens 

Br-84 1. 6( -2) 

1-131 4.5(-2) 

1-132 2.1(-1) 

1-133 1.4(-1) 

1-134 3.4(-1) 

1-135 2.6(-1) 

Cs, Rb 

Rb-88 1.9(-1) 

Cs-134 7.1(-3) 

Cs-136 8.7(-4) 

Cs-137 9.4(-3) 

Water Activation Products 

N-16 

Tritium 

H-3 

Other Nuclides 

Na-24 

Cr-51 

Mn-54 

Fe-55 

Fe-59 

Co-58 

Co-60 

4.0(+1) 

1.0(0) 

4.7(-2) 

3.1(-3) 

1.6(-3) 

1.2(-3) 

3.0(-4) 

4. 6( -3) 

5.3(-4) 

2-6 

Secondarl Coo1ant** 

3.4(-8) 

8.9(-8) 

3.0(-8) 

5.9(-8) 

1.5(-7) 

1.5(-8) 

5.4(-7) 

2.7(-8) 

1.8(-7) 

7.1(-9) 

2.5(-8) 

1.8(-8) 

5.2(-8) 

2.4(-7) 

1.6(-7) 

3.8( -7) 

3.0(-7) 

6.0(-7) 

3.0(-8) 

3.6(-9) 

3.9(-8) 

1.0(-6) 

1. O( -3) 

1.0(-7) 

6.9(-9} 

3.6(-9) 

2.7(-9} 

6.7(-10} 

1.0(-8) 

1.2(-9} 
Isotope 

Zn-65 

Sr-89 

Sr-90 

Sr-91 

Y-91m 

Y- 91 

Y-93 

Zr-95 

Nb-95 

Mo-99 

Tc-99m 

Ru-l03 

Ru-l06 

Ag-11 Om 

Te-129m 

Te-129 

Te-131 m 

Te-131 

Te-132 

Ba-140 

La-140 

Ce-141 

Ce-143 

Ce-144 

W-187 

Np-239 

TABLE 2-3 (continued) 

Reactor Coolant* 

5.1(-4) 

1.4(-4) 

1.2(-5) 

9.6(-4) 

4.6(-4) 

5.2(-6) 

4.2(-3) 

3.9(-4) 

2.8(-4) 

6.4(-3) 

4.7(-3) 

7.5(-3) 

9.0(-2) 

1.3(-3) 

1.9(-4) 

2.4( -2) 

1.5(-3) 

7.7(-3) 

1.7(-3) 

1.3(-2) 

2.5( -2) 

1.5(-4) 

2.8(-3) 

3.9(-3) 

2.5(-3) 

2.2(-3) 

Secondary Coolant** 

1.1(-9) 

3.1(-10) 

2.7(-11) 

2.1(-9) 

9.7(-10) 

1.2(-11) 

9.3(-9) 

8.7(-10) 

6.2(-10) 

1. 4( -8) 

1.0(-8) 

1.7(-8) 

2.0(-7) 

2.9(-9) 

4.2(-10) 

5.1(-8) 

3.3(-9) 

1.5(-8) 

3.8(-9) 

2.9(-8) 

5.6(-8) 

3.3(-10) 

6.2(-9) 

8.7(-9) 

5.6(-9) 

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 

H-3 concentrations are obtained from Reference 1. 

** Based on primary-to-secondary leakage of 75 lbjday. The concentrations 

given are for steam leaving a steam generator. 

2-7 
TABLE 2-4 

PARAMETERS USED TO DESCRIBE THE REFERENCE PRESSURIZED WATER 

REACTOR WITH U-TUBE STEAM GENERATORS 

Range 

Parameter Symbol 

P 

Units 

Nominal 

Value Maximum Minimum 

Thermal Power 

Steam fl ow rate 

Weight of water in reactor coolant 

system 

Weight of water in all steam 

generators 

Reactor coolant letdown flow 

(purification) 

Reactor coolant letdown flow (yearly 

average for boron control) 

Steam generator blowdown flo~ (total) 

Fraction of radioactivity in blowdown 

stream that is not returned to the 

secondary coolant system 

Flow through the purification system 

cation demineralizer 

Ratio of condensate demineralizer flow 

rate to the total steam flow rate 

Ratio of the total amount of noble 

gases routed to gaseous radwaste 

from the purification system to the 

total amount of noble gases routed 

from the primary coolant system to 

the purification system (not including the boron recovery system) 

FS 

WP 

WS 

FD 

FB 

FBD 

NBD 

FA 

NC 

y 

MWt 

1 b/hr 

lb 

lb 

1 b/h r 

1 b/hr 

1 b/hr 

1 b/h r 

3,400 

1.5(7) 

5.5(5) 

4.5(5) 

3.7(4) 

500 

75,000 

1.0* 

3,700 

0.0** 

0.0 

3,800 

1.7(7) 

6.0(5) 

5.0(5) 

4.2(4) 

1,000 

100,000 

1.0 

7,500 

0.01 

0.01 

3,000 

1.3(7) 

5.0(5) 

4.0(5) 

3.2(4) 

250 

50,000 

0.9 

0.0 

0.0 

0.0 

* This value is based on a nominal case of blowdown through blowdown demineralizers 

back to the main condenser (no condensate demineralizers). Value taken from blowdown demineralizer DF's in Section 2.2.18. Value for cesium and rubidium is 0.9. 

** 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 ;s 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 
TABLE 2-5 

PARAMETERS USED TO DESCRIBE THE REFERENCE PRESSURIZED WATER 

REACTOR WITH ONCE-THROUGH STEAM GENERATORS 

Nominal Range 

Parameter S,lmbo 1 Units Value Maximum 

The rma 1 Powe r 

Steam flow rate 

Weight of water in reactor coolant 

system 

Weight of water in all steam 

generators 

Reactor coolant letdown flow 

(purification) 

Reactor coolant letdown flow (yearly 

average for boron control) 

Flow through the purification system 

cation deminera1izer 

Ratio of condensate deminera1izer 

flow rate to the total steam flow 

rate 

Ratio of the total amount of noble 

gases routed to gaseous radwaste 

from the purification system to the 

total amount routed from the 

primary coolant system to the 

purification system (not including 

the boron recovery system) 

P MWt 

FS 1 b/h r 

WP 1b 

WS 1b 

FD 1 b/h r 

FB 1 b/h r 

FA 1 b/h r 

NC 

y 

3,400 3,800 

1.5(7) 1.7(7) 

5.5(5) 6.0(5) 

1.0(5) * 

3.7(4) 4.2(4) 

500 1,000 

3,700 7,500 

0.65** 0.75 

0.0 0.01 

* The secondary coolant inventory is not of importance in a once-through steam 

generator plant because decay is not an important removal mechanism for most 

of the isotopes. 

Minimum 

3,000 

1.3(7) 

5.0(5) 

* 

3.2(4) 

250 

0.0 

0.55 

0.0 

** For a 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 deminera1izer, a value of NC = 1.0 is 

used by the PWR-GALE Code. 

2-9 
TABLE 2-6 

VALUES USED IN DETERMINING ADJUSTMENT FACTORS FOR 

PRESSURIZED WATER REACTORS 

Element Class 

Water 

Noble Activation Other 

Symbol Description Gases Halogens Cs, Rb Products H-3 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 in passing 

through the purification 

demineralizer 

R Removal rate :1 reactor 0.0009 0.067 0.037 0.0 *** 0.066 

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 

demi ne ra 1 i zers 

r Removal rate :lsecondary 

coolant (Hr )ttt 

U-tube steam generator t O. 1 7 0.15 tt *** 0.17 

Once-through steam t 27 7.5 tt *** 14 

generator 

FL Primary-to-secondary 75 75 75 75 75 75 

leakage (lbjday) 

* 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 
TABLE 2-6 (continued) 

** These values of R apply to the reference PWRls whose parameters are given in 

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

PWRls not included in Tables 2-4 and 2-5, the appropriate value for R may be 

determined 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 

WP 

*** 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 PWRls with a moderate 

amount of tritium recycle and can be used to calculate source terms in accordance 

with Regulatory Guide 1.112, IICalculations of Releases of Radioactive Materials in 

Gaseous and Liquid Effluents from Light-Water-Cooled Power Reactors.1I 

t Noble gases are rapidly transported out of the water in the steam generator and 

swept out of the vessel in the steam; therefore, the concentration in the water 

is negligible and the concentration in the steam is approximately equal to the 

ratio of the release rate to the steam generator and the steam flow rate. These 

noble gases are removed from the system at the main condenser. 

tt Water activation products exhibit varying chemical and physical properties in 

reactor 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 PWRls whose parameters are given in 

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

not included in Tables 2-4 and 2-5, the appropriate value for r may be determined 

by the following equation: 

r = (FBD)(NBD) + (NS)(FS)(NC)(NX) 

WS for halogens, Cs, Rb, and other nuclides 

2-11 
N 

I 

N 

El ement Cl as·s 

Noble gases 

Halogens 

Cs, Rb 

Water activation 

products 

Trit ium 

Other nuclides 

TABLE 2-7 

ADJUSTMENT FACTORS FOR PWR's WITH U-TUBE STEAM GENERATORS 

Adjustment Factors 

Secondary Coolant 

Reactor Water (f)* 

162P 

WP 

0.0009 + A ** 

R + A 

162P 0.067+A 

WP R + A 

162P 

--wP 

'162P 

WP 

0.037 + A 

R + A 

1.0 

*** 

0.066 + A 

R + A 

Water 

5 4.5xlO 0.17+A 

WS r + A 

5 4.5xlO 0.15+A 

WS r + A 

4.5 x 10 5 

--ws 

*** 

5 4.5 x 10 0.17 +A 

WS r + A 

f 

f 

f 

Steam 

1.5 x 107 

f FS 

5 4.5 x 10 0.17 + A 

WS r + A 

5 4.5 x 10 0.15 + A 

WS r + A 

4.5 x 10 5 

WS 

*** 

5 4.5x10 0.17+A 

WS r + A 

f 

f 

f 

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

** A is the isotopic decay constant (hr-l ). 

*** 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. 
TABLE 2-8 

ADJUSTMENT FACTORS FOR PWR's WITH ONCE-THROUGH STEAM GENERATORS 

Nuclide 

Noble gases 

Halogens 

Cs, Rb 

Water activation 

products 

Triti urn 

Other nuclides 

Adjustment Factors 

Reactor Water (f)* 

162P 0.0009 + A 

---wP R + A 

162P 0.067 + A ~ R +A 

162P 0.037 + A 

---wv- R + X 

1.0 

** 

162P 0.066+A 

""""Wi' R + X 

Secondary Coolant 

7 1.5 x 10 

FS 

5 1.0 x 10 

WS 

** 

f 

105 (14 + A) f WS r + X 

* 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. 

2-13 
N 

I 

~ 

REACTOR 

VESSEL 

f----~ 

FD 

LEAKAGE 

~ STEAM 

GENERATORS 

PRIMARY 

SIDE 

STEAM 

GENERATORS 

SECONDARY 

SIDE 

IFS 

~ 

NS_ ~ • I 

FEEDWATER 

FS(l-NC) 

r-' 

I 

I 

r-~::>~D~~N _____ --4- __ J 

FBD 

MAIN 

CONDENSER 

NC·FS 

CONDENSATE 

DEMINERALIZER 

NX 

OFF GAS 

PURIFICATION I FA I CATION 

'----.. ~IDEMINERALIZER I "DEMINERALIZER 

NB 

SYMBOLS ARE DEFINED IN TABLES 2-4 AND 2-6 

L..-___ ... ·~,I~F~B_.~ TO BORON RECOVERY SYSTEM 

PURIFICATION I V 

SYSTEM r---- DEGASSING 

TANK 

FIGURE 2-1 REMOVAL PATHS FOR PRESSURIZED WATER REACTOR 

WITH U-TUBE STEAM GENERATORS 
N 

I 

--' 

U1 

REACTOR 

VESSEL 

FD 

LEAKAGE 

STEAM 

....... :-/ 

STEAM 

FSNS .... ~ 

------ STEAM MAIN GENERATORS GENERATORS CONDENSER PRIMARY SECONDARY FS(1·NC) SIDE SIDE 

FEEDWATER 

NC·FS 

CONDENSATE 

DEMINERALIZER 

PURIFICATION FA CATION 

DEMINERALIZER DEMINERALIZER 

SYMBOl 

NB FB 

TO BOROI 

PURIFICATION Y 

SYSTEM ...... I 

TANK 

L---.-

FIGURE 2·2 REMOVAL PATHS FOR PRESSURIZED WATER REACTOR 

WITH ONCE·THROUGH STEAM GENERATORS 

OF F GAS 
where 

C s 

w(,\ + R)K 

C is the specific activity (in ~Ci/g) 

K is a conversion factor, 454 g/lb 

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 ~Ci/hr) 

w is the fluid weight (in lb), and 

is the decay constant (hr~). 

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. 

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 

Steam flow rate, lb/hr 17 x 106 13 x 106 - 17 x 106 

Mass of reactor coolant, lb 

Water weight in all steam generators, lb 

Reactor coolant letdown, lb/hr 

Cation demineralizer flow, lb/hr 

Shim bleed rate - yearly average, lb/hr 

Steam generator blowdown flow, lb/hr 

Fraction of blowdown activity not 

returned to secondary system 

Cation demineralizer flow, lb/hr 

Condensate demineralizer flow fraction 

Y (see definition in Table 2-4 and 

page 1-26) 

2-16 

5.5 x 

4.4 x 

4.9 x 

4.9 x 

650 

60,000 

0.99 

4900 

0.0 

105 

105 

104 

103 

5.0 x 105 - 6.0 x 

4.0 x 105 - 5.0 x 

3.2 x 104 - 4.2 x 

o - 7.5 x 103 

250 - 1000 

50,000 - 100,000 

0.9 - 1.0 

0.0 - 7500 

0.0 - 0.01 

105 

105 

104 
Since in this example the parameter for reactor coolant letdown 

rate (4.9 x 104 lb/hr) is outside the range specified in Table 2-4 

(3.2 - 4.2 x 104 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 

---wr 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) 

WP 

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

(1 ) 

(2) 

Use the sample case parameters given above and the noble gas parameters 

given in Table 2-6 and substitute in Equation (2) above. 

R = 650 + (4.9 x 104 - 650) x 0.25 = 

5.5 x 105 

Use the value of R in Equation (1) above. 

0.023 

-3 f = 162 x 3800 0.0009 + 5.5 x 1~3 = 0.25 

5.5 x 105 0.023 + 5.5 x 10 

The adjusted Xe-133 primary coolant concentration 

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

= 0.25 x 2.6 ~Ci/q = 0.65 ~Ci/g 

2. Halogens (1-131 is used as an example) 

Using the equation for halogens in Table 2-7, the adjustment 

factor, f, is calculated as follows: 

2-17 
f - l62P 0.067 + >.. 

---wv- R+>" 

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. 

R = (FO)(NB) + (1 - NB)(FB + (FA)(NA)) 

WP 

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

(3 ) 

(4) 

Use the sample case parameters given above and the halogen parameters 

given in Table 2-6 and substitute in Equation (4) above. 

R = (4.9 x 104 

x 0.99) + (1 - 0.99)(650 + (4900)(0.0)) = 

5.5 x 105 0.088 

Use the value of R in Equation (3) above. 

f 162( 3800) 

5.5 x 105 

0.067 + 3.6 x 10-3 

0.088 + 3.6 x 10- 3 

The adjusted 1-131 concentration 

0.86 

= (adjustment factor) x (standard 1-131 concentration) 

= 0.86 x 0.045 ~Ci/g = 0.039 ~Ci/g 

3. Cs, R~ (Cs-137 is used as an example) 

Using the equation for Cs and Rb in Table 2-7, the adjustment 

factor, f, is calculated as follows: 

f - 162P 0.037 + >.. 

---wr R+>" 

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

(5) 

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. 

R = (4.9 x 104 x 0.5) + (0.5)(650 + (4900)(0.9)) = 0.05 

5.5 x 105 

2-18 
Use the value of R in Equation (5) above. 

-6 

f -_ 162(3800) ---~~-~ 

0.037 + 2.6 x 10 = 0.83 5 -6 5.5 x 10 0.05 + 2.6 x 10 

The adjusted Cs-137 concentration 

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

-3 -3 

= 0.83 x 9.4 x 10 11Ci/g = 7.8 x 10 11Ci/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: 

f = 162P 0.066 + A. 

WP R + A. (6) 

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. 

R = (4.9 x 104

) (0.98) + (1 - 0.98)(650 + (4900)(0.9)) = 0.087 

5 5 105 

• x 

Use the value of R in equation (6) above. 

f = 162 (3800~ 

5.5 x 10 

-3 (0.066 + 8.9 x 10 ) 

-3 0.087 + 8.9 x 10 

The adjusted concentration of Te-132 

= 0.87 

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

-3 -3 = 0.87 x 1.7 x 10 IlCi/g = 1.5 x 10 11Ci/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 factors, 

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 stea~ generators. 

The secondary coolant concentrations are based on 75 lb/day primaryto-secondary leakage. The primary-to-secondary leakage rate experience 

for 79 years of experience at operating 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 of the ratio of the condensate demineralizer flow 

rate to the total steam flow rate is 0.65. This indicates that the 

nominal case is a design 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 
Isotope 

1-131 

1-133 

Isoto~ 

1-131 

TABLE 2-9 

SUMMARY OF 1-131 AND 1-133 PRIMARY COOLANT CONCENTRATIONS IN PWR'S* 

(~Ci/g) 

H.B. Robinson 2 

(1973-1978)** 

3.lE-03 

1.2E-02 

Fort Calhoun 

(1976-1977) 

1.8E-01 

Arkansas 

(1976) 

7.3E-03 

*** 

Zion 1/2 

(1975-1978 ) 

2.3E-02 

D.C. Cook 1 

(1976-1978) 

7.8E-03 

1.8E-02 

Turkey Point 3/4 

--.0274-1978) 

2.1E-02 

Trojan 

(1977-1978) 

1.3E-02 

1.5E-02 

Palisades 

(1972-1976) 

1.2E-02 

1.6E-02 

Point Beach 1/2 

(1972-1979) 

7.7E-02 

3.6E-Ol 

Three Mile Island Ca 1 vert Cl i ffs 

(1975-1977) (19 7 6) 

2.lE-02 3.6E-02 

R.E. Ginna 

(1971-1978) 

2.2E-Ol 

6.9E-01 

Beaver Va 11 ey 

(1977-1978) 

1.8E-03 ~ 1-133 1.6E-01 6.9E-02 6.9E-02 *** *** 5.:>E-03 N 

Indian Point 2/3 Kewaunee Prairie Island 1/2 Surry 1/2 J.M. Farley Yankee Rowe Rancho Seco 

Isotope (1 975-1 978) (1975-1978t ( 1 97~)-1}:3.IJ . _ llYJ}-.1.'J7B) (1978 ) (1975) (1978) 

1-131 2.5E-02 5.3E-03 8.1E-03 2.1E-02 5.4E-04 4.9E-03 1.3E-02 

1-133 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 1-131 and 1-133 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 radionuc1ide. 

*** No value reported. 
TABLE 2-10 

SUMMARY OF RADIONUCLIDE PRIMARY COOLANT CONCENTRATIONS IN PWR'S* 

(\lCi /g) 

H.B. Robinson 2 Arkansas D.C. Cook 1 Trojan Palisades Poi nt Beach 1/2 R.E. Gi nna Isoto~e ~1973-1978)** (1976) (1976-1978) (1977-1978) (1972-1976) ~1972-1979) (1971-1978) 

Kr-85m 1.8E-02 8.0E-03 3.4E-02 2.2E-02 4.0E-Ol 2.2E-0l 2. 4E- 01 

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

Kr-87 1.7E.-02 6.7E-03 4.7E-02 4.2E-02 4.6E-Ol 1.1E-Ol 3.8E-Ol 

Kr-88 2.3E-02 1.3E-02 5.7E-02 4.0E-02 7.5E-Ol 3.0E-Ol 6.3E-Ol 

Xe-131rn *** *** *** *** 4.4E+OO 9.4E-Ol *** 

Xe-133m 1.61::-03 3.3E-03 1.5E-02 2.7E-03 1.9E-Ol 6.5E-02 *** 

Xe-133 2.3E-Ol 2.1E-Ol 5.8E-Ol 5.7E-Ol 4.9E+OO 2.8E+00 6.0E+00 

Xe-135m 2.1 E-02 *** *** *** 1.0E-02 1.4E-Ol 1.5E-Ol 

Xe-135 7.8E-02 2.7E-02 1.9E-Ol 1.2E-Ol 1. lE+OO 1.lE+OO 2.2E+00 

Xe-137 *** *** *** *** *** *** *** 

Xe-138 *** *** *** 6.7E-02 2.2E-03 1. 7E- 01 *** 

r;v Br-84 *** *** *** *** *** *** *** ~ 1-132 1.5E-02 *** *** 1.9E-02 7.1 E- 03 3.6E-Ol 7.3E-Ol 

1-134 3.2E-02 *** 2.4E-02 2.2E-02 1. OE-02 6.2 E-Ol 1.2E+00 

1-135 1.9E-02 *** 2.0E-02 1.7E-02 9.2E-03 5.7E-Ol 6.6E-Ol 

Rb-88 *** *** *** 3.5E-02 2.6E-02 1.7E-Ol 3.7E-Ol 

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.lt-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.lE-02 3.1E-02 

N-16 *** *** *** *** *** *** *** 

H-3 *** 6.3E-02 2.1E-01 *** 7.5E-02 5.0E-Ol 6.7E-Ol 

Na-24 1. 3E- 01 8.7E-02 1.2E-02 1.3E-02 5.6E-03 7.6E-02 *** 

Cr-51 3.5E-04 3.2E-03 *** *** 9.lE-03 *** 1.1 E-04 

Mn-54 3.4E-04 7.6E-04 8.3E-03 9.7E-04 1. lE-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 *** *** 
TABLE 2-10 (continued) 

SUMMARY OF RADIONUCLIDE PRIMARY COOLANT CONCENTRATIONS IN PWR'S* - - ( fl C i / g) ------

H.B. Robinson 2 Arkansas 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-91 *** *** *** *** *** *** *** 

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

Zr-95 1.3E-05 3.4E-04 4.5E-03 *** 1.0E-04 *** 1.5E-03 

Nb-95 1.3E-05 3.1 E-04 2.4E-03 *** 7.6E-05 3.6E-04 8.1 E-05 

Mo-99 *** 7.2E-05 *** *** 5.7E-04 3.8E-02 4.1 E-04 

Tc-99m *** *** *** *** 7.3E-04 2.5E-02 *** 

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

N Ru-106 *** *** *** *** *** *** *** 

~ Ag- 110m *** *** *** *** *** 8. 8E- 03 *** 

w Te-129m *** *** *** *** *** *** *** 

Le-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 *** *** *** *** *** *** *** 

Ce-143 *** *** *** *** *** *** *** 

Ce-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* 

(]lCi/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 

IsotoEe (l976-1977) (1975-1978) (l974-1978) (1975-1978) (l975 ) (1976) (1975-1977 ) (1981 ) (1979) 

Kr-85m 1.9E-Ol *** 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-Ol 

Kr-87 1.9E-Ol *** 9.0E-02 *** 7.6E-03 *** *** 1.lE-03 5.9E-02 

Kr-88 3.2E-Ol *** 1.3E-Ol 7.3E-02 1.9E-02 *** *** 1.lE-03 9.9E-02 

Xe-131m 6.8E-02 *** 1.2E-03 *** *** *** *** 4.2E-05 3.5E-03 

Xe-133m 1.6E-Ol *** 9.1E-03 *** *** *** *** 7.2E-05 4.5E-02 

Xe-133 6.7E+OO *** 8.8E-Ol 8.3E-Ol 2.1E-Ol *** *** 2.2E-03 1.5E+00 

Xe-135m 9.5E-02 *** 1.7E-Ol 1.0E-Ol *** *** *** 1.4E-03 6.0E-Ol 

Xe-135 9.3E-Ol *** 5.lE-Ol 1.9E-Ol 3.0E-02 *** *** 3.6E-03 4.6E-Ol ~ Xe-137 *** *** 3.4E-02 *** *** *** *** *** ***t ~ Xe-138 1.8E-Ol *** 7.6E-02 *** *** *** *** 2.9E-03 1.7E-01 

Br-84 *** *** 1. lE-02 *** *** *** *** 1.0E-03 5.5E-02 

1-132 7.1E-02 9.6E-02 9.3E-02 *** 1.8E-02 *** 'tf** 5.1E-03 5.3E-02 

1-134 3.8E-02 1.3E-01 1.5E-Ol *** *** *** *** 9.0E-03 8.3E-02 

1-135 7.4E-02 1.1E-01 8.6E-02 *** *** *** *** 5.8E-03 6.0E-02 

Rb-88 5.0E-Ol 2.3E-Ol 1.0E-Ol *** *** *** *** 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. lE-04 *** *** *** *** 3.2E-06 1. 9E-04 

Cs-137 2.0E-02 1.2E-02 3.lE-03 2.4E-02 *** *** *** 6.7E-05 9.4E-03 

N-16 *** *** *** *** *** *** *** *** *** 

H-3 1.3E-Ol 1.5E-Ol *** *** *** 4.4E-02 1.2E-Ol 2.9E-Ol 2.5E-01 

Na-24 8.8E-03 1.0E-Ol 1.0E-02 3.6E-03 *** *** *** 9.0E-03 1.4E-02 

Cr-51 1.5E-02 2.1E-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. 1 E-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-04 

Zn-65 2.6E-03 2.4E-03 1.6E-05 *** *** *** *** 1.7E-06 2.2E-05 

Sr-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 *** 
TABLE 2-10 (continued) 

SUMMARY OF RADIONUCLIDE PRIMARY COOLANT CONCENTRATIONS IN PWR'S* 

(lJCi /g) 

Fort Turkey Indian Yankee Ca 1 vert Three Mi le Prairie Rancho 

Calhoun 1 Zi on 1/2 Point 3/4 Pt 2/3 Rowe Cl iffs 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 *** *** *** *** 2.1 E- 04 2.6E-03 

Zr-95 1.5E-03 4.2E-04 4.5E-05 *** 2.8E-04 *** *** 4.6E-06 2.9E-04 

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 8.1E-04 *** 5.0E-03 *** *** 1.3E-04 1.7E-03 

Tc-99m 4.1 E-04 *** 2.7E-06 *** 4.9E-03 *** *** *** *** N Ru-l03 5.4E-02 1.8E-04 2.1 E-05 *** *** *** *** 6.7E-07 7.0E-05 

~ Ru-l06 *** 9.0E-02 *** *** *** *** *** *** *** U1 Ag-llOm 2.2E-04 3.1E-03 1.1 E-05 *** *** *** *** 3.7E-06 9.7E-05 

Te-129m 2.1E-04 3.8E-04 1.9E-04 *** *** *** *** 2.0E-06 *** 

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

Te-131m *** 2.1E-03 3.7E-04 *** *** *** *** *** *** 

Te-131 *** *** 7.9E-03 *** *** *** *** *** 7.4E-03 

Te-132 *** 1.8E-04 4.0E-05 *** *** *** *** 1.2E-06 3.1E-05 

Ba-140 1.lE-03 1.0E-03 1.1 E-04 *** *** *** *** 1. 9E-05 2.5E-04 

La-140 4.2E-04 1.8E-03 1.3E-04 *** *** *** *** 1.4E-05 1.1E-04 

Ce-141 4.3E-04 1.lE-04 1.7E-05 *** *** *** *** *** 4.6E-05 

Ce-143 8.2E-03 4.6E-04 4.9E-05 *** *** *** *** 1. 9E-05 *** 

Ce-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. 
TABLE 2-11 

MONTHLY AVERAGE* PRIMARY/SECONDARY LEAKAGE 

gal/day at 70°F; density = 

1970 

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

San Onofre 4 4 4 4 3 9 11 8 14 S** S 0 

Connecticut 0 10 0 S 0 0 20 10 20 0 0 0 

Yankee 

R. E. Ginna 0 0 0 0 0 0 

Point Beach 0 

1971 

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

San Onofre 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 1 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 0 

San Onofre S 0 0 0 0 0 22 0 10 30 4 31 

Connecticut 40 40 40 40 40 S 0 0 0 0 0 0 

Yankee 

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

H. B. Robi nson 60 60 60 60 3 0 0 0 0 0 0 0 

Point Beach 40 50 55 55 55 55 55 55 55 S S S 

Point Beach 2 0 0 0 

Su rry 1 0 

Tu rkey Poi nt 3 0 

* Leakage values listed begi n with the first year of commercial operation. 

** Shutdown not included in average. 

NA - Not Available. 

2-26 
TABLE 2-11 (continued) 

MONTHLY AVERAGE* PRIMARY/SECONDARY LEAKAGE 

(gal/day at 70°F; density = 8.3 1b/ga1) 

Plant 

San Onofre 

Connecticut 

Yankee 

R. E. Ginna 

H. B. Robinson 

Point Beach 

Point Beach 2 

Surry 

Turkey Point 3 

Surry 2 

Turkey Point 4 

Plant 

San Onofre 

Connecticut 

Yankee 

R. E. Ginna 

H. B. Robinson 2 

Point Beach 

Point Beach 2 

Su rry 1 

Tu rkey Poi nt 3 

Su rry 2 

Turkey Point 4 

Zi on 1 

Zion 2 

Indian Point 2 

Prairie Island 

J 

3 

o 

o 

6 

S 

o 

o 

o 

J 

o 

o 

S 

2 

o 

o 

S 

o 

o 

s 

s 

F 

3 

o 

o 

6 

S 

o 

o 

o 

M 

o 

o 

o 

6 

o 

o 

o 

o 

A 

o 

o 

o 

S 

o 

o 

o 

o 

1973 

M J 

o 0 

lOS 

o 0 

o 0 

o 0 

o 0 

o 0 

o 0 

o 0 

1974 

J 

o 

o 

o 

1 

o 

o 

o 

o 

o 

F M A M J J 

44 60 60 0 o 

o 

o 

o 0 S 0 o 

S S 0 

10 112 98 

o 

o 

S 

o 

o 

o 

s 

o 

o 

o 

o 

o 

o 

s 

s 

o 

o 

o 

o 

o 

o 

2-27 

o a a 

NA 19 2 

o 

o 

o 

o 

S 

o 

o 

o 0 

o 0 

115 55 

o 0 

38 0 

o 0 

o s 

o 

A 

o 

S 

o 

1 

o 

o 

o 

o 

o 

A 

o 

o 

o 

1 

o 

o 

S 

o 

S 

o 

1 

o 

o 

o 

o 

o 

o 

S 

o 

o 

o 

1 

o 

o 

o 

o 

S 

o 

1 

o 

o 

o 

o 

o 

o 

o 

2 

o 

o 

1 

o 

o 

115 115 4 

NA NA S 

o 0 S 

22 0 0 

S 0 0 

o 0 

000 

o o o 

N 

S 

S 

o 

7 

o 

o 

o 

o 

o 

o 

N 

2 

o 

o 

1 

o 

S 

S 

S 

S 

o 

o 

o 

o 

o 

o 

S 

o 

o 

5 

o 

o 

o 

o 

o 

o 

D 

2 

o 

o 

1 

o 

S 

S 

S 

S 

o 

o 

o 

o 

o 
TABLE 2-11 (continued) 

MONTHLY AVERAGE* PRIMARY/SECONDARY LEAKAGE 

(gal/day at 70°F; density = 8.3 1b/ga1) 

1975 

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

San Onofre 2 2 2 2 3 5 0 a a a a 0 

Connecticut 0 a 0 0 0 S a 0 0 0 0 0 

Yankee 

R. E. Ginna a a 3 S 0 a a a 0 0 0 a 

H. B. Robinson 2 3 5 3 2 0 a S 7 

Point Beach 0 61 S 0 2 2 2 2 S S 

Point Beach 2 0 0 0 0 0 0 a 0 0 0 0 

Su rry 1 S 0 0 a 0 a a a 125 S S 26 

Tu rkey Poi nt 3 0 0 0 0 a 0 0 0 0 0 S S 

Surry 2 0 0 0 0 S 0 0 0 0 0 a 0 

Tu rkey Poi nt 4 0 0 0 S S S 7 20 79 0 0 50 

Zion 1 0 0 S 0 0 S 0 0 S 0 0 0 

Zion 2 0 S a 0 0 s 0 0 S 0 0 0 

Indian Point 2 a 102 S 0 0 a 0 0 a s 0 0 

Prairie Island 1 a 0 0 0 0 0 a 0 0 a 0 a 

Prairie Island 2 0 0 0 0 a 0 0 0 0 0 0 0 

Cook 1 0 a 0 0 

2-28 
TABLE 2-11 (continued) 

MONTHLY AVERAGE* PRIMARY/SECONDARY LEAKAGE 

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

1976 

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

San Onofre 0 0 0 0 0 0 46 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 2 2 2 2 6 S S 

Point Beach 2 32 200 5 29 10 12 13 21 23 25 25 25 

Surry 0 0 28 86 NA 19 39 14 33 S S 

Tu rkey Poi nt 3 12 6 14 0 11 19 0 12 S S S 

Surry 2 95 31 10 0 S 0 0 0 6 S S 200 

Tu rkey Poi nt 4 62 0 0 S S S 0 0 80 42 S 0 

Zion 0 0 S S S S 0 0 0 S 0 0 

Zi on 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 

Prai ri e Island 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 
Pl ant 

San Onofre 

Connecticut 

R. E. Gi nna 

TABLE 2-11 (continued) 

MONTHLY AVERAGE* PRIMARY/SECONDARY LEAKAGE 

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

J F 

S S 

o o 

o o 

M 

S 

o 

o 

A 

o 

o 

S 

1977 

M J 

o 

o o 

S o 

J 

2 

o 

o 

A 

2 

o 

o 

H. B. Robinson 2 o o o 

Point Beach 4 5 3 6 3 5 5 5 

Point Beach 2 25 35 33 S o o a a 

S o 

S o 

o S 

o o 

o 6 

4 s 

a o 

N D 

2 

S o 

a o 

41 52 

8 

o 

7 

o 

Surry S 77 144 53 0 o o 26 58 58 21 0 

Turk ey Poi n t 3 

Surry 2 

Turkey Point 4 

Zion 1 

Zion 2 

Indian Point 2 

Prairie Island 

Prairie Island 2 

Cook 1 

Trojan 

Indian Point 3 

Beaver Valley 

Salem 1 

Fa rl ey 1 

o o o o o o 28 72 72 56 S 

548 360 S 

o 

o NA 18 10 8 4 o 14 0 

23 29 71 96 7 

000 a 0 

S S S 0 0 

o 0 0 S 0 

o 0 0 S 0 

o o o a o 

S S o o o 

o o o o S 

o o o o o 

o a 

2-30 

S 

o 

o 

o 

o 

o 

a 

S 

o 

S 

S 

o 

o 

S 

a 

o 

o 

a 

a 

o 

o 

o 

o 

o 

a 

o 

o 

o 

o 

a 

o 

o 

o 

S 

o 

o 

a 

o 

a 

o 

o 

S 

o 

4 

S 

o 

o 

a 

a 

o 

S 

S 

S 

o 

S 

a 

o 

a 

S 

a 

a 

S 

o 

S 

o 

o 

o 

a 

a 

S 

a 

o 

S 

o 

a 

a 
TABLE 2-11 (continued) 

MONTHLY AVERAGE* PRIMARY/SECONDARY LEAKAGE 

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

1978 

Plant J F M A M J Average, * gal/da}: 

San Onofre 1 S 4.6 

Connecticut 0 0 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 

Poi nt Beach 20 7 7 7 120 7 15 

Point Beach 2 0 0 0 S 0 0 7.9 

Surry 0 0 0 0 S S 22 

Tu rkey Poi nt 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 1 7. 

Zi on 1 0 0 0 0 0 0 0 

Zion 2 0 S S S S 0 0 

Indian Point 2 0 S S S S 0 7.8 

Prairie Island 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 Va 11 ey 0 0 0 0 S S 0 

Salem 1 0 0 0 S S S 0 

Fa r1 ey 1 0 0 0 0 0 0 0 

Operation Weighted Average 9 

* Ave rage daily value for each reactor is obtained by the sum of the total 

month 1y leakage rates divided by the total number of days in operation. 

2-31 
Facility 

Palisades 

Kansai 

Point Beach 

Tu rkey Poi nt 3 

Tu rkey Poi nt 4 

TABLE 2-12 

MOISTURE CARRYOVERS IN RECIRCULATING U-TUBE 

STEAM GENERATORS* 

Average 

Percent Carryovers 

0.08 

0.05 

0.2 

0.6 

1.6 

0.5 

* Measurement based on Na concentration. 

2-32 

Reference 

10, 11 

10, 11 

8, 1 2 

6 

6 
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 from building ventilation systems prior to treatment are calculated by the P~JR-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 radioiodine at PWR's has been conducted by 

C. Pelletier, et ale (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 iorline-131 building vent releases 

are directly related to the reactor coolant iodine-131 concentration. As 

a result, the releases of iodine are expressed as "normalized" releases, 

that is, the absolute measured release rate in Ci/yr is divided by the 

reactor coolant concentration in l1Ci/g to give a "normalized" release 

rate of iodine-131 in Ci/yr/~Ci/g as shown in the following equation: 

where 

RN = normalized release rate of iodine-131, Ci/yr/~Ci/g 

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

CRW = measured reactor water iodine-131 concentration, ~Ci/g 

2-33 
TABLE 2-13 

ANNUAL IODINE NORMALIZED RELEASES 

FROM CONTAINMENT VENTILATION SYSTEMSt 

NORMAL OPERATION LEAK RATE* 

Data Source 

Ft. Calhoun (Ref. 4) 

Three Mile Island 1 (Ref. 7) 

Turkey Point 3/4 (Ref. 6) 

Main Yankee (Ref. 16) 

Gi nna (Ref. 19) 

Yankee Rowe (Ref. 14, 16) 

Prairie Island 1/2 (Ref. 42) 

Rancho Seco (Ref. 43) 

Average 

Normalized Release/Unit 

-3 10 %/day 

0.0014 

2.5 

0.9 

O. 1 

0.064 

1.0 

0.005 

2.56 

0.80 

RELEASE FOR EXTENDED OUTAGES** 

Data Source 

Three Mi le Island 1 (Ref. 7) 

Calvert Cliffs 1 (Ref. 7) 

Average 

Normalized Release/Unit 

(Ci/yr/jJCi/g) 

0.44 

0.19 

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 radioiodine between 

the water phase in the leakage and the gas phase where it is measured. 

** The normalized release rate, expressed in Ci/yr/~Ci/g, represents the 

effective leak rate for radioiodine. It is the combination of the 

reactor water iodine leakage rate into the buildings, and the partiti~ning 

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 1331• 

2-34 
TABLE 2-14 

ANNUAL IODINE NORMALIZED RELEASES* 

FROM AUXILIARY BUILDING VENTILATION SYSTEMSt 

NORMAL OPERATION 

Data Source 

Zion 1/2 (Ref. 5) 

Fort Calhoun (Ref. 4) 

Ginna (Ref. 7) 

Calvert Cliffs 1 (Ref. 7) 

Three Mi le Island 1 (Ref. 7) 

Turkey Point 3/4 (Ref. 6) 

Prairie Island 1/2 (Ref. 42) 

Rancho Seco (Ref. 43) 

Average 

SHUTDOWN 

Data Source 

Ginna (Ref. 7) 

Calvert Cliffs 1 (Ref. 7) 

Three Mile Island 1 (Ref. 7) 

Turkey Point 3/4 (Ref. 6) 

Rancho Seco (Ref. 43) 

Average 

Normalized Release/Unit 

(Ci/yrhlCi/g) 

1.0 

0.12 

0.032 

0.57 

0.034 

1.85 

0.013 

0.97 

0.68 

Normalized Release/Unit 

(Ci /yr/llCi /g) 

0.08 

0.016 

0.14 

6.8 

1. 14 

2.50 

* The normalized release rate, expressed in Ci/yr/llCi/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 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 133 1• 

2-35 
TABLE 2-15 

ANNUAL IODINE NORMALIZED RELEASES* 

FROM REFUELING AREA VENTILATION SYSTEMSt 

NORMAL OPERATION 

Data Source 

Ginna (Ref. 7) 

Ca 1 vert Cl iffs 1 (Ref. 7) 

Three Mi le Island 1 (Ref. 7) 

Turkey Point 3 (Ref. 6) 

Prairie Island 1/2 (Ref. 42) 

Rancho Seco (Ref. 43) 

Average 

SHUTDOWN 

Data Sou rce 

Ginna (Ref. 7) 

Calvert Cliffs 1 (Ref. 7) 

Three Mile Island 1 (Ref. 7) 

Turkey Point 3 (Ref. 6) 

Rancho Seco (Ref. 43) 

Average 

Normalized Release/Unit 

(Ci /yr/}1Ci /g) 

0.008 

0.049 

0.0012 

O. 16 

0.019 

0.01 

0.038 

Normalized Release/Unit (Ci/yr/~Ci/g) 

0.014 

0.039 

0.06 

0.05 

0.30 

0.093 

* The normalized release rate, expressed in Ci/yr/~Ci/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 

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 1 data. The normalized release . 131 133 rates are appllcable to both I and I. 

2-36 
TABLE 2-16* 

ANNUAL IODINE NORMALIZED RELEASES** 

FROM TURBINE BUILDING VENTILATION SYSTEMSt 

NORMAL OPERATION 

Data Source 

Monticello 

Oyster Creek 

Vermont Yankee 

Pil grim 

Browns Fe rry 

References 3, 5 of Ref. 15 

Average 

EXTENDED SHUTDOWN 

Data Source 

Monticello 

Oyster Creek 

Vermont Yankee 

Browns Fe rry 

References 3, 5 of Ref. 15 

Average 

Normalized Release/Unit 

( C i / y r /lJ C i / 9 ) 

3.1 x 103 

6.0 x 103 

0.35 x 103 

8.5 x 103 

1. 3 x 103 

3.3 x 103 

3.8 x 103 

Normalized Release/Unit 

(Ci /yr/lJCi /g) 

1.7 x 102 

3.5 x 102 

0.63 x 102 

1 .3 x 102 

1.4 x 103 

4.2 x 102 

* 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 Ci/yr/lJCi/g during different 

modes of operation represents the effective leak rate for radioiodine. 

It is a function of iodine leak rate via steam and the partition 

coefficient for radioiodine 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 1331• 

2-37 
The normalized reactor water release rate, expressed in Ci/yr/~Ci/g 

represents an effective leak rate for reactor water containing iodine. 

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 directly related to the secondary coolant iodine-131 concentration. 

Therefore, for the turbine building, the normalized iodine release, RN, is determined using the following expression: 

where 

RA = 

CRW = 

PC 

normalized release rate of secondary coolant water containing 

iodine-131, Ci/yr/~Ci/g 

absolute (measured) iodine-131 release rate, Ci/yr 

measured secondary coolant iodine-131 concentration, ~Ci/g 

measured iodine partition coefficient from secondary coolant 

water to steam. 

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. 

Data on normalized release rates from the three reactors used in 

the EPRI study and the five reactors used in the NRC sponsored study are 

given for normal operation and shutdown periods in Tables 2-13 through 

2-15, for the containment building, auxiliary building and refueling 

area, respectively. Also given in Table 2-13 is the normalized value of 

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 

of PWR and BWR power generation equipment which is housed in the turbine 

building is essentially identical, the turbine building leakage rates 

from PWR's and BWR's should be similar. Therefore, for the PWR turbine 

building normalized 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. 

The data in Tables 2-14 through 2-16 are expressed as total 

normalized releases during power operation of 300 days an~ the total 

normalized releases during shutdowns of 65 days. Since the reactors 

used in the EPRI study and the NRC study experienced several intermittent 

2-38 
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 

from 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 PWRGALE Code by the iodine concentrations in the reactor coolant for that 

particular PWR using the following expression: 

where 

= calculated annual release rate for particular PWR for iodine 

isotope i, Ci/yr 

normalized annual release rate of iodine from Tables 2-14 

and 2-15, Ci/yr/~Ci/g 

calculated reactor water concentration for particular 

PWR for iodine isotope i, ~Ci/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: 

where 

= calculated annual release rate for particular PWR for iodine 

isotope i, Ci /yr 

= normalized annual release rate of iodine from Table 2-16, 

Ci/yrNCi/g 

2-39 
calculated secondary coolant concentration for particular 

PWR for iodine isotope i, ~Ci/g 

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. 

Iodine released from PWR building ventilation systems appear in one 

of the following chemical forms: particulate, elemental, hypoiodious 

acid (HOI) 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 IODINE APPEARING IN EACH CHEMICAL FORM 

FROM PWR BUILDING VENTILATION SYSTEMS 

Containment Auxil i ary Tu rbi ne Fuel Handling 

Particulate 0.09 0.04 * 0.01 

Elemental 0.21 0.21 0.78 0.17 

HOI 0.21 0.22 * 0.57 

Organic 0.49 0.53 * 0.25 

* No data on breakdown of other species. 

2.2.5 RADIOACTIVE PARTICULATES RELEASED IN GASEOUS EFFLUENTS 

2.2.5.1 Parameter 

Use the radioactive particulate release rates in gaseous effluents 

given in Table 2-17. 

2-40 
TABLE 2-17 

PARTICULATE RELEASE RATE FOR GASEOUS EFFLUENTS* 

(Ci/yr)/unit 

Auxi li ary Fuel Pool Waste Gas 

Nuclide Containment Building Area S~stem 

Cr-51 9.2(-3) 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) 

* 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. 

2-41 
2.2.5.2 Bases 

Tables 2-18 through 2-21 list measured particulate releases at 12 

operating 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 NOBLE GAS RELEASES FROM BUILDING VENTILATION SYSTEMS 

2.2.6.1 Parameter 

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 wCi/cc for Main Yankee and 

7 x 10- 3 wCi/cc for Ginna. The containment volumes at these facilities 

are approximately 1.8 x 106 ft 3 for Maine Yankee and 1 x 106 ft 3 for 

Ginna. Based on these values, the total microcuries of xenon-133 in 

the containment building atmosphere are 

Maine Yankee 

(5 x 10-3 wCi/cc)(1.8 x 106 ft3)(2.83 x 104 cc/ft 3) = 2.5 x 108 wCi Xe-133 

Ginna 

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 

2.5 x 108 wCi 

(5.3 dayjO.693) 

7 3.3 x 10 wCi/day Xe-133 leakage 

2-42 
TABLE 2-18 

MEASURED RELEASES UPSTREAM OF HEPA FILTERS - CONTAINMENT 

(Ci /yr) 

Turkey Prairie 

Three Mile Point Ca 1 vert Island Rancho 

Island 1 Fort Calhoun Zion 1 & 2 3 & 4 Cl iffs 1 Ginna 1 & 2 Seco Average 

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 ND ND NA NA NA NA 9.2( -3) 

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) 

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) 

r;" Zr- 95 NA NA NA NA NA NA NA NA NA 

t; Nb-95 1.1(-2) ND ND ND NA NA NA NA 1.8(-3) 

Ru-l03 9.5(-3) ND ND ND NA NA NA NA 1.6( -3) 

Ru-l06 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) 

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-140 NA NA NA NA NA NA NA NA NA 

Ce-141 8.0(-3) ND ND ND NA NA NA NA 1.3(-3) 

NO = Not detected. For averaging purposes, a value of zero was assumed. 

NA = Not analyzed (or no n~asurement taken); plants not included in averaging. 
Three Mile 

Island 1 

Nuc 1 ide (Ref. 7} 

Cr-51 1.4( -3) 

Mn-54 1.1(-4) 

Co-57 NA 

Co-58 1.1(-3) 

Co-60 2.0( -4) 

Fe-59 2.3(-4) 

Zr-95 2. 7( -4) N Nb-95 1.4(-4) .1. Ru-103 9.1(-5) +:> Ru-106 NA 

Sb-125 NA 

Cs-134 8.0(-5) 

Cs-136 NA 

Cs-137 2.0(-4) 

Ba-140 NA 

Ce-141 1.5(-4) 

ND = Not detected. 

TABLE 2-19 

MEASURED RELEASES UPSTREAM OF HEPA FILTERS - AUXILIARY BUILDING 

(Ci /yr) 

Tu rkey 

Point Ca 1 vert 

Fort Calhoun Zi on 1 & 2 3 & 4 C1 iffs 1 Ginna 

(Ref. 4} (Ref. 5} {Ref. 6} (Ref. 7} (Ref. 7} 

ND NA ND NA 1.9(-4) NA NA 6.3(-5) 3.0( -4) 6.7(-5) 

NA NA NA NA NA 

2.0( -3) NA 1.1(-3) 4.8( -4) 6.3( -4) 

2.7(-4) NA 6.0(-4) 2.0(-3) 7.7(-4) 

ND ND ND NA 1.9(-5) ND ND ND 7.9(-3) 4.1(-5) 

ND ND ND NA 6.0(-5) 

ND NA ND NA 6.9(-5) 

ND NA ND NA 2.4(-5) 

NA NA NA NA NA 

1.6(-3) NA 7.9(-4) 2.0( -3) 3.4(-4) 

ND NA ND NA 1.9(-4) 

1.8( -3) NA 1.4( -3) 1.9(-3) 1.1(-3) 

ND ND ND NA 1.6( -3) ND NA ND NA 2.8(-5) 

For averaging purposes, a value of zero was assumed. 

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

Prai rie 

Island Rancho 

1 & 2 Seco Average 

(Ref. 42} (Ref. 43) (Ci/yr)/unit 

NA NA 3.2( -4) 

2.7(-6) NA 7.8(-5) 

NA NA NA 

4.0(-5) 1.2(-2) 1.9(-3) 

4.5(-5) 7.3(-4) 5.1(-4) 

NA NA 5.0(-5) 

5.7(-6) NA 1.0(-3) 

1.0( -5) NA 3.0(-5) 

2.7(-6) NA 2.3(-5) 

NA NA 6.0(-6) 

7.7(-6) NA 3.9(-6) 

1.5(-6) 5.2(-5) 5.4(-4) 

NA NA 4.8(-5) 

9.4(-6) 8.0(-5) 7.2(-4) 

NA NA 4.0( -4) 

1.5(-6) NA 2.6( -5) 

Measurements 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. 
TABLE 2-20 

MEASURED RELEASES UPSTREAM OF HEPA FILTERS - FUEL POOL AREA 

(Ci/yr)/unit 

Turkey Prairie 

Three Mile Point Ca 1 vert Island Rancho 

Island 1 Fort Calhoun Zion 1 & 2 3 & 4 Cl i ffs 1 & 2 Ginna 1 & 2 Seco 

Nuclide (Ref. ?L (Ref. 4} (Ref. 5) (Ref. 6) (Ref. 7} (Ref. 7} (Ref. 42) (Ref. 43) Average 

Cr-51 l.8( -4) NA NA NA NA NA NA NA 1.8( -4) 

Mn-54 l.0( -5) NA NA NA 1.2(-3) NA 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) NA NA NA 5.0(-3) NA 6.9(-6) 7.6(-6) 8.4(-3) 

Fe-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) Nb-95 3.0(-5) NA NA NA 9.5(-3) NA 1.7(-5) NA 1.9(-3) 

r;v Ru-103 9.8(-5) NA NA NA NA NA 1.7(-5) NA 3.8(-5) 

""" Ru-106 6.9(-5) NA NA NA NA NA NA NA 6.9(-5) U1 Sb-125 1.7(-4) NA NA NA NA NA NO NA 5.7(-5) 

Cs-134 9.0( -6) NA 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) 

NO = 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 Ca 1 vert Island Rancho 

Island 1 Fort Ca 1 houn Zi on 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) (Ci/yr)/unit 

Cr-51 8.4(-5) NO NO NO NA NA NA NA 1.4(-5) 

Mn-54 1.1(-5) NO 4.0(-6) NO 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) 

Co-60 8.0(-5) NA 3.2(-6) 2.9(-7) NA NA NA 5.9(-8) 1.4(-5) 

Fe-59 7.2(-6) 1.8( -6) 1.9(-6) NO NA NA NA NA 1.8(-6) N Zr-95 1.9(-5) NO NO NA NA NA NA NA 4.8(-6) 

J,.. Nb-95 2.2(-5) NO NO NO NA NA NA NA 3.7(-6) 0'1 Ru-l03 1.9(-5) NO NO NO NA NA NA NA 3.2(-6) 

Ru-l06 1.6(-5) NO NO NO 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 NO NO 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) NO NO NO NA NA NA NA 2.3(-5) 

Ce-141 1. 3( -5) NO NO NO NA NA NA NA 2.2(-6) 

NO = Not detected. For averaging purposes, a value of zero was assumed. 

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

2 x 10

8 ~Ci _ 2 6 7. X 33 (5.3 day/0.693) - • x 10 ~Cl/day e-1 leakage 

Based on the xenon-133 concentration during power operation (Ref. 29) 

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 

3.3 x 107 ~Ci/day = 

(lOt ~Ci/cc x 28,300 cc/ft3 

x 11,000 ft 3) 

O.Ol/day = 1 %/day 

Ginna 

2.6 x 1 0

7 _-,-____ ...J. 11 __ C i / "---O<.- day __ --;;~ =. 0 005/ d ay - 0 • 5%/ day 

(30t ~Ci/cc x 28,300 x 6,234 ft3) 

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 magnitude. 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. 

In the auxiliary building, the source term calculation is based 

on an assumed primary coolant leakage rate of 160 1b/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. 

t The reactor coolant concentrations for Xe-133 are measured values 

during 12/73 - 6/74 for Main Yankee and September and October of 1971 

for Ginna (Ref. 16). 

2-47 
In the turbine building, it is assumed that steam will leak to the 

turbine building atmosphere at a rate of 1700 1b/hr. The leakage is 

considered to be from many sources, each too small to be detected 

individually, but which, taken collectively, total 1700 1b/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 

1. Pressurized water reactors, with U-tube steam generators, that 

are currently under design, either direct their b10wdown through 

a heat exchanger to cool the b10wdown 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 b10wdown 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 b10wdown 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 b10wdown flash tank 

conditions (240°F, sat.). Although the iodine species in the b10wdown 

stream will be predominantly nonvolatile (volatile species are degassed 

in the steam 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 PWRls which have provisions 

to prevent flashing (cooling b10wdown 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 

The iodine releases from the main condenser air ejector exhaust are 

based on secondary side measurements made by EPR! 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 { C i /.l r /J.l C i / 9 ) 

Turkey Poi nt 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 

radioiodine. It is the combination of the steam flow to the main 

condenser, the partitioning of radioiodine between the main condenser 

and the air ejector exhaust where it is measured and the partition 

coefficient for radioiodine from water to steam in the steam generator. 

2-49 
In a manner similar to the discussion of normalized releases for 

building ventilation releases in Section 2.2.4, the main condenser air 

ejector exhaust iodine releases are directly related to the secondary 

coolant iodine-13l concentration. Therefore, for the air ejector 

exhaust, the normalized iodine release, RN, is determined using the 

following expression: 

where 

RN normalized effective release rate of iodine-13l, Ci/yr/wCi/g 

RA measured (absolute) iodine-13l release rate, Ci/yr 

CR~J = measured secondary coolant iodine-13l concentration, wCi/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: 

where 

calculated annual release for particular PWR for iodine 

isotope i, Ci/yr 

normalized annual release rate of iodine from Table 2-22, 

Ci/yr/wCi/g 

calculated secondary coolant concentration for particular 

PWR for iodine isotope i, wCi/g 

= Iodine partition coefficient from water to steam in the 

steam generator for the particular PWR (see Table 2-6) 

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 hotwe11 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 app1icant '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 
TABLE 2-23 

PWR CONTAINMENT PURGING AND VENTING EXPERIENCE (REF. 17) 

Yankee Rowe 

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 hours 

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 hours 

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 hours 

Connecticut Yankee 

Purge - Cannot purge during operation, only during shutdown 

Reason: Primarily to remove activity 

Duration: 1 to 2 days 

San Onofre 

Purge each cooldown approximately 4 times per year, no purging 

during power operation 

Purge for at least 24 hours, 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. 

2-52 
TABLE 2-23 (continued) 

PWR CONTAINMENT PURGING AND VENTING EXPERIENCE (REF. 17) 

Oconee 2 

Continuous purge since startup, lowest purge rate approximately 

20,000 ft3/min 

Reason: Reduction of gaseous activity for maintenance, etc. 

Robinson 2 

Purge approximately 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 hours 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 

Total time 

14 

502 hours* 

253 hours 

3 hou rs 

Maximum duration (1 purge) 

Minimum duration (1 purge) 

Infrequent purges or vents of 10 minutes for pressure control. 

Turkey Point 4 

For period 1/1/74 to 7/1/74 

Total purges Total time 

Maximum time (1 purge) 

Minimum time (1 purge) 

Su rry 1 and 2 

5 

984 hours* 

742 hours 

5 hou rs 

Containment operates at negative pressure. Discharge from vacuum 

pumps through filters to stack. During cold shutdown, there 

is continuous purging of containment. 

Prairie Island 

Frequency: 

Reason: 

Once per week for about 8 hours 

To relieve pressure buildup due to instrument air leakage 

to containment 

2-53 
TABLE 2-23 (continued) 

PWR CONTAINMENT PURGING AND VENTING EXPERIENCE (REF. 17) 

Kewaunee 

Frequency: 5 times in 60 days usually for less than 1 hour, longer 

if for personnel entry. 

Reason: Pressure control. During the 60-day period, purging 

occurred for personnel entry. 

Point Beach 

Continuous venting through a monitoring line at about 10 ft 3

/min 

flow. Gas filtered on way to stack. 

Pa1idades 

One per week for about 10 minutes duration (planned upon resumption 

of power operation) 

Reason: To control pressure buildups 

Zion 

Venting for pressure buildup about twice per week depending on 

outside temperature. 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 hours. 

Millstone 2** 

* 

For period from 7/1/75 through 12/31/77. 

About 45 purges per year with an average duration of about 9 hours. 

Generally, long purges occur during plant outages while at cold 

shutdown conditions. 

** Data for these plants was obtained from the Semi-annual Release Reports 

for the plants for the period indicated. 

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 hours prior 

to purging, that it provides a OF for radioiodine removal on charcoal 

adsorbers corresponding to the values in Table 1-5, and a OF 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 

hours prior to purging. It is considered that charcoal adsorbers provide 

a OF for iodine corresponding to the values in Table 1-5, that HEPA filters 

provide a OF of 100 for particulates, and that the containment air mixing 

efficiency is 70%. The system operation time of 16 hours considers 

that two shifts will elapse following a decision to enter the containment. 

The time period of two shifts is a reasonable amount of time for 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 

Length of test run 

Initial iodine activity 

Final iodine activity 

Containment volume 

Internal recirculation 

system fl ow rate 

Symbol 

T 

A 

o 

A 

V 

F 

2-55 

Value 

6 hours 

-8 1.2 x 10 j.lCi/cc 

1.2 x 10-9 j.lCi/cc 

106 ft 3 

6. 1 x 105 ft 3 /h r 
The efficiency of iodine removal, E, can be estimated from 

Ao FET A = exp (-v-) 

Substituting Ginna data into the equation 

10 = exp (3.7E), therefore E = 0.63. 

The iodine removal efficiency E is a function of filter efficiency, E

a , and mixing efficiency, Em' 

E = Elm = 0.63 

In calculating Em we used the assumed OF of 10 for charcoal derived from 

Table 1-5, (90% removal). Using Ea equal to 0.9, Em ;s calculated to be 70%. 

Em = EIEa = 0.63/0.9 = 0.7 

2.2.11 RADIOIODINE REMOVAL EFFICIENCIES FOR CHARCOAL 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 

radioiodine are as follows: 

Activated Carbon Bed Deptha 

2 inches. Air filtration system designed 

to operate inside primary containment. 

2 inches. Air filtration system designed 

to operate outside the primary containment 

and relative humidity is controlled to 70%. 

Removal Efficiencies 

For Radioiodine(%) 

90 

70 

a Multiple beds, e.g., two 2-inch beds in series, should be treated as 

single bed of aggregate depth of 4 inches. 

2-56 
Activated Carbon Bed Deptha 

4 inches. Air filtration system designed 

to operate outside the primary containment 

and relative humidity is controlled to 70%. 

6 inches. Air filtration system designed 

to operate outside the primary containment 

and relative humidity is controlled to 70%. 

2.2.11.2 Bases 

Removal Efficiencies 

For Radioiodine(%) 

90 

99 

The removal efficiencies assigned to HEPA filters for particulate 

removal and charcoal adsorbers for radioiodine removal are based on the 

design, testing and maintenance criteria recommended in 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 ft3 /day (STP) per reactor for PWR's without recombiners and 30 ft3 /day 

(STP) per reactor for PWR's with recombiners. 

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. 

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 
TABLE 2-24 

WASTE GAS SYSTEM INPUT FLOW TO PRESSURIZED STORAGE TANKS 

FOR PWR's WITHOUT RECOMBINERS 

Net Flow per Reactor 

Reactor ft 3

/day (STP) 

San Onofre 2/3 57 

Waterford 3 171 

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 

Average Net Flow for PWR's 

without recombiners = 170 ft3/day (STP) per reactor 

2-58 
TABLE 2-25 

WASTE GAS SYSTEM INPUT FLOW TO PRESSURIZED STORAGE TANKS 

FOR PWR's WITH RECOMBINERS 

Net Flow per Reactor 

Reactor ft 3

/day (STP) 

WPPSS 1 96 

Farley 1/2 3 

McGuire 1/2 18 

Average Net Flow for PWR's 3 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). 

2-59 
2.2.13 HOLDUP TIMES FOR CHARCOAL DELAY SYSTEMS 

2.2.13.1 Parameter 

where 

Kr 

Xe 

T = 0.011 MK/F 

T is the holdup time, in days; and 

K 3 is the dynamic adsorption coefficient, in cm /g, (see chart 

M 

F 

be lOw); 

is the mass of charcoal adsorber, in 103 lbs; 

is the system flow rate, in ft 3/min; 

0.011 is the factor to convert from (103 lb. cm3/g)/(ft3/min) to days. 

Dynamic adsorption coefficients, K, (in cm

3jg) are as follows: 

Operating nOF 

Dew Point 45°F 

18.5 

330 

Operating nOF 

Dew Point OaF 

25 

440 

Operating nOF 

Dew Point -40°F 

70 

1160 

Operati ng OaF 

Dew Point -20°F 

105 

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 

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 

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 

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: 

T (days) M(103 lbs) K{cm3

/g)(454 g/lb) 3.53 x 10-5 ft3

/cm3) 

F(ft3/min)(1440 min/day) 

T = 0.011 ~K 

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 
co, 

--u 

2 

tZ 

w 

~ 

LL. 

LL. 

W 

0 

U 

Z 

0 

i= 

~ 

ex: 

0 

Vl 

0 

<t: 

..J 

<t: 

0 

u 

ex: 

<t: 

J: 

U 

105 r---------------------------------------------rr---------------. 

O 

103 

102 

V DATA FROM UNDERHILL 

o DATA FROM BROWNING 

I I I I 

+100 +32 0 -40 

I I I I 100~ ______ _L ________ L_ ______ _L ________ ~ ______ _..l ________ _L ______ ~ 

o 0.001 0.002 0.003 0.004 0.005 0.006 

RECIPROCAL TEMPERATURE (I/°K) 

FIGURE 2-3 KRYPTON AND XENON K VALUES AS A FUNCTION 

OF RECIPROCAL TEMPERATURE 

2-62 

0.007 
60r-------~--------------~------~------~------~ 

COlUMBIA-G CHARCOAL. 8-14 MESH 

KRYPTON-85 WITH OXYGEN SWEEP GAS (25°C) 

10~------------------------------__ ------~--____ ~ 

O~ ______ ~ ______ ~ ______ ~ ______ ~ ______ ~ ______ -J 

o 2 3 4 5 

MOISTURE CONTENT OF CHARCOAL (wt %1 

FIGURE 2-4 EFFECT OF MOISTURE CONTENT ON THE DYNAMIC 

ADSORPTION COEFFICIENT 

2-63 

6 
45 

40 

35 ~ 

I 

C) 

UJ 30 

s: 

0 

w 25 c:I 

c:: 

0 

Vl 20 0 

< 

c:: 

UJ 15 ~ 

< 

s: 

10 

5 

0 

0 

;' .- ./ 

/ 

, 

I WATER ADSORPTION DESORPTION I AND DESORPTION 

I ISOTHERMS ON MACAR 

I G210 AFTER REPEATED 

ADSORPTION AND I DESORPTION 

I (EOUILIBRIUM) 

r DYNAMIC 

DETERMINATION 

15-40o C RANGE 

MACAR G210 

-1-

10 20 30 40 50 60 70 80 90 100 110 

AIR RELATIVE HUMIDITY 

FIGURE 2-5 CHARCOAL MOISTURE AS A FUNCTION OF 

RELATIVE HUMIDITY 

2-64 

120 
N 

I 

0'1 

U1 

SOURCE 

1. REACTOR CONTAINMENT 

a. Primary coolant pump seal 

leakage 

b. Primary coolant leakage, 

miscellaneous sources 

c. Primary coolant equipment 

2. PRIMARY COOLANT SYSTEMS 

(OUTSIDE OF CONTAINMENT) 

a. Primary coolant system 

equipment drains 

b. Spent fuel pit liner drains 

c. Primary coolant sampling 

system drains 

d. Auxiliary building floor 

drains 

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) 

Deep-bed condo 

demi nera 1 i zers 

with ultrasoni c 

resin cleaner 

20 

10 

500 

80 

700 

200 

200 

Deep-bed condo 

demineralizers 

without 

ultrasonic 

resin cleaner 

20 

10 

500 

80 

700 

200 

200 

Filterdemi neral i zer 

20 

10 

500 

80 

700 

200 

200 

Plant with 

b lowdown t reatment. Product 

not recycled to 

condenser or 

secondary coolant 

system 

20 

10 

500 

80 

700 

200 

200 

FRACTION OF 

PRIMARY COOLANT 

ACTI V lTY (PCA) 

0.1 

1.67* 

0.001 

1.0 

0.001 

0.05 

0.1 
TABLE 2-26 (Continued) 

3. SECONDARY COOLANT 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 

N f. Steam generator blowdown Plant dependenttt Plant dependent tt I 

0> 

0'\ g. Turbine building floor 7200 7200 7200 7200 Calculated in 

drains GALE Code 

4. DETERGENT AND DECONTAMINATION 

SYSTEMS 

a. On-site laundry facility 300 300 300 300 ** 

b. Hot showers Negligible Negligible Negligible Negl i gi b 1 e 

c. Hand wash sink drains 200 200 200 200 ** 

d. Equipment and area 40 40 40 40 ** 

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 RADIOACTIVE MATERIALS 

IN UNTREATED DETERGENT WASTE 

Nuclide Ci /y"r/reactor 

P-32 1.8(-4) 

Cr-51 4.7(-3) 

Mn-54 3.8( -3) 

Fe-55 7 .2( -3) 

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) 

1-1 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 
in Table 2-27 should be added to the adjusted liquid source term. 

Detergent waste releases should be reduced, using appropriate decontamination 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 radionuclide 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 Parameter 

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 of 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 all material collected since the previous 

regeneration, leaving a constant quantity of material on the resins after 

regeneration). Regeneration cycles are normally controlled by particulate 

buildup on resin beds, resulting in high pressure drops across the bed. 

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 systerTi wastes, with a primary coolant tritium 

concentration of 1.0 ~Ci/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 
Nuclide 

P-32 

Cr-51 

Mn-54 

Fe-55 

Fe-59 

Co-58 

Co-60 

Ni-63 

Sr-89 

Sr-90 

y- 91 

Zr-95 

Nb-95 

Mo-99 

Ru-103 

Ru-106 

Ag-ll Om 

Sb-124 

1-131 

Cs-134 

Cs-136 

Cs-137 

Ba-140 

Ce-141 

Ce-144 

TOTAL 

Note: NA = 

TABLE 2-28 

RADIONUCLIDE DISTRIBUTION OF DETERGENT WASTE 

(millicuries/month) 

Oyster 

Creek Ginna Zion* 

(1971-1973) (1972-1973) (1977) 

(Ref. 41) (Ref. 8) (Ref. 51 

1.5(-2) NA NA 

2.3(-1) NA 9.4(-1) 

1.3 1.2(-1) 1.6(-1) 

3.5(-1) NA 1.9 

2.9(-1) NA 2.6(-1) 

3.5(-1) 4.1(-1) 2.4 

3.8 9( -1) 9.8(-1) 

NA NA 3.5(-1) 

2.1(-2) NA 7 (- 3) 

2.5(-3) NA 7.6(-4) 

NA NA 1.4(-2) 

8.3(-2) 1.6(-1) 1.4(-1) 

1.6(-1) 2( -1) 2.7(-1) 

NA 5( -3) NA 

1.3(-2) 3.2(-2) 5.2(-2) 

NA 7.4(-1) NA 

NA 1 (-1 ) NA 

6.1(-2) NA 4.7(-2) 

4.3(-1) 5.5(-2) 1.7(-1) 

1.7(-1) 1.4 1.5 

NA NA 6.2(-2) 

2.9(-1) 2.5 2.1 

7 .6( -2) NA NA 

3.3(-2) 5( -3) NA 

7 .3( -2) 5.8(-1) NA 

7.7 7.2 11.4 

radionuclides were not analyzed. 

* For two un its. 

2-69 

Fort 

Calhoun 

(1977) 

(Ref. 4) 

NA 

NA 

1.9(-2) 

1.6(-1) 

NA 

1.5(-1) 

3(-2) 

7.1(-2) 

1.4(-3) 

NA 

NA 

NA 

NA 

NA 

NA 

NA 

NA 

NA 

1.7(-2) 

1.4 

NA 

1.7 

NA 

NA 

NA 

3.5 
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. 

2.2.17.2 Bases 

The release rate of 0.4 Ci/yr/MWt is based on a review of the tritium 

release rates at a number of PWRls 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 Ci. 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 

1410 Ci (0.8)(365 days/yr) e-0•693 (l)/12.3 = 0.43 Ci/yr per MWt 910,000 MWd 

Table 2-29 gives the reported liquid and gaseous tritium releases 

for 1972-1978 for thirty-five operating PWRls that use zirca10y 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 approximately 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 ~rimary coolant that is released in the nonrecyc1ab1e 

waste streams for the boron recovery, clean waste, and dirty waste systems. 

The concentration of tritium in wastes originating from primary coolant 

is assumed to be 1 pCi/m1, 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 re1p~sed in liquid 

effluents is up to a maximum of 90% of the total quantity of tritium 

calculated to be available for release. 

2-70 
TABLE 2-29 

TRITIUM RELEASE DATA FROM OPERATING PWR's WITH ZIRCALOY-CLAD FUELS* 

Nuclear Thermal Output 

per unit 

Power 106 MWDt 

per unit Startup 

Reactor Name MWt Date 1972 1973 1974 1975 1976 1977 

R. E. Gi nna 1520 1969 0.32 0.45 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 Poi nt 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 

Ca 1 vert Cl iffs 1/2 2700 1974/76 0.58 0.84 1.24 

Cook 1 3250 1975 0.90 0.64 

Mi 11 stone 2 2560 1975 0.63 0.59 

Trojan 3411 1975 0.31 0.88 

St. Lucie 1 2560 1976 0.73 

Beaver Va 11 ey 2652 1976 0.42 

Salem 1 3338 1976 0.28 

* Data from semiannual reports of reactors listed. 

2-71 
TABLE 2-29 (continued) 

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 **  ** ** 

Fort 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 

Mi 11 stone 2 2560 1975 21.3 47 

Trojan 3411 1975 1.5 2.9 

St. Lucie 1 2560 1976 320 

Beaver Valley 2652 1976 213 

Salem 1 3338 1976 51 

* Data from semiannual reports of reactors listed. 

** No reported data. 

2-72 
TABLE 2-29 (continued) 

TRITIUM RELEASE DATA FROM OPERATING PWR's WITH ZIRCALOY-CLAD FUELS* 

Power Tritium Released (Ci/Yr) Per Site 

per unit Startup Liquid 

Reactor Name 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 210 185 8.3 41.3 9.6 56 

Maine Yankee 2440 1972 219 177 368 153 

Indian Point 2/3 2758 1973/76 48 366 332 371 

Surry 1/2 2441 1972/73 246 442 782 408 

Tu rkey Poi nt 3/4 2200 1972/73 580 793 771 924 

Oconee 1/2/3 2568 1973/74/74 124 3550t 2192t 1918t Zion 1/2 3250 1973/73 39.4 1.1 727 

Fort Ca 1 houn 1420 1973 111 122 157 

Prai r;e Island 1/2 1650 1973/74 763 1925 1349 

Kewaunee 1650 1974 277 213 295 

Three Mile Island 1 2535 1974 463 189tt 192 tt 

Rancho Seco 2772 1974 132 0.0 0.09 

Arkansas 1 2568 1974 460 212 245 

Ca 1 vert Cl iffs 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 2652 1976 108 

Salem 1 3338 1976 296 

* Data from semiannual reports of reactors listed. 

t 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. 

tt 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* 

Power 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 O. 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 

Arkansas 1 2568 1974 0.21 0.13 0.19 

Ca 1 vert C1 iffs 1/2 2700 1974/76 0.13 O. 11 0.16 

Cook 1 3250 1975 0.06 0.13 

Mi 11stone 2 2560 1975 0.14 0.13 

Trojan 3411 1975 0.04 0.10 

St. Lucie 1 2560 1976 0.22 

Beaver Valley 2652 1976 0.22 

Salem 1 3338 1976 0.36 

* Data from semiannual reports of reactors listed. 

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) 

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 0.39 0.22 0.59 0.36 0.35 0.37 0.51 

Maine Yankee 0.14 0.09 0.13 0.07 0.18 

Indi an Pte 2/3 0.04 0.17 0.19 0.08 

Surry 1/2 O. 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 

TM! 1 0.20 0.46 0.13 0.17 

Arkansas 0.21 0.13 0.19 

Calvert Cl iffs 1/2 0.13 0.11 0.16 

Cook 0.06 0.13 0.31 

Mi 11 stone 0.14 0.13 

Trojan 0.04 0.10 

St. Lucie 0.22 

Beaver Valley 0.22 0.51 

Salem 0.36 0.41 

Average 0.16 0.29 0.30 0.25 0.26 0.31 0.40 

2-75 
TABLE 2-31 

TRITIUM RELEASE DATA FROM OPERATING PWR's 

PERCENT OF TOTAL TRITIUM RELEASED IN LIQUID EFFLUENTS 

Reactor 

R. E. Ginna 

H. B. Robinson 

Point Beach 1/2 

Palisades 

Maine Yankee 

Indian Point 2/3 

Surry 1/2 

Turkey Point 3/4 

Oconee 1/3 

Zion 1/2 

Fort Calhoun 

Prairie Island 1/2 

Kewaunee 

Three Mile Island 

Rancho Seco 

Arkansas 1 

Calvert Cliffs 1/2 

Cook 1 

Mi 11 stone 2 

Trojan 

St. Lucie 

Beaver Va 11 ey 

Salem 1 

Weighted Average* 

1972 

100.0 

99.8 

98.6 

97.7 

99.2 

1973 

99.6 

99.1 

95.7 

99.8 

98.0 

1974 

99.8 

90.1 

95.1 

** 

96.7 

70.6 

80.4 

98.4 

99.4 

91.1 

1975 

97.8 

76.4 

83.3 

** 

97.4 

93.7 

93.2 

99.6 

68.9 

** 

97.9 

98.7 

88.1 

92.0 

94.5 

99.9 

99.5 

89.5 

* Average weighted by nuclear thermal output per unit. 

** No reported data. 

1976 

91.1 

86.1 

63.7 

** 

99.0 

93.3 

67.8 

99.3 

81.4 

** 

98.0 

98.3 

99.7 

20.9 

O.ott 

96.9 

87.0 

100.0 

92.9 

96.0 

83.5 

1977 

70.4 

91.8 

83.8 

96.2 

98.6 

96.8 

31. 7 

99.6 

96.8 

** 

98.1 

93.9 

98.7 

59.8 

0.43tt 

56.3 

83.1 

100.0 

81.8 

99.1 

43.1 

33.6 

85.3 

78.5 

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

2-76 
The difference between the tritium calculated to be available for 

release from the primary coolant and the tritium calculated to be released 

in liquid effluents is considered to be released as a vapor through 

building ventilation exhaust systems. Based on measurements taken in 

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 refueling 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 

Anion 

Mixed bed purification 100 

system (L iB03) 

Boron recycle system 10 

Evaporator condensate (H+OH-) 5 

Radwaste (H+ OH-) 102(10) 

Steam Generator Blowdown 102(10) 

Ca t i on bed ( H+ ) (a ny system) 1 (1 ) 

Anion bed (OH-) (any system) 1 02 (1 0) 

Powdex (any system) 10(10) 

Cs, Rb 

2 

2 

2( 10) 

1 O( 10) 

10(10) 

1 (1) 

2( 10) 

Other 

Nuclides 

50 

10 

10 

102(10) 

102 (1 0) 

10(10) 

1 (l ) 

10(10) 

Note: For two demineralizers in series, the DF f6r 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* 

DISTRIBUTION OF TRITIUM RELEASE IN GASEOUS EFFLUENTS 

Source of Gaseous Tritium Release (% of Total) 

Plant 

Ginna (Ref. 7) 

Calvert Cliffs (Ref. 7) 

Three Mile Is. (Ref. 7) 

Zion 1/2 (Ref. 5) 

Turkey Point 3/4 (Ref. 6) 

Rancho Seco (Ref. 43) 

Prairie Island 1/2 (Ref. 43) 

Average 

NM - Not measured. 

Auxil i ary 

Building 

31 

38 

5 

79 

75 

92 

7.2 

32 

Refueling 

Area 

69 

46 

43 

WA 

17 

WA 

91.8 

50 

Containment 

Building 

NM 

16 

52 

21 

8 

8 

1.0 

18 

WA - Release from refueling area combined with auxiliary building release. 

* The following method is used to determine the 3H release in this table. 

Containment Building operation average % of total release 

(16 + 52 + 21 + 8+8 + 1)% i (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)% i (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 Auxiliary Building release which is equal to 

(31 + 38 + 5 + 29 + 75 + 42 + 7.2)% i (7) = 32% 

2-78 
2. When two demineralizers are used in series, the first demineralizer will have a higher OF than the second. However, the 

data in Reference 32 indicate that Cs and Rb will be more 

strongly exchanged in the second demineralizer in series than 

the first as the concentration of preferentially exchanged 

competing nuclides is reduced. 

3. As indicated in Reference 32, compounds of Y, Mo, and Tc form 

colloidal particles that tend to plate out on solid surfaces. 

Mechanisms such as plateout on the relatively large surface 

areas provided by demineralizer resin beds result in removal 

of these nuclides to the degree stated above. An analysis 

of effluent release data indicates that these nuclides, although 

present in the primary coolant, are not found in the effluent 

streams. 

2.2.18.2 Bases 

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 

(Refs. 4, 5, 6, 42, and 43); (2) the findings of a generic review in the 

nuclear industry by the Oak Ridge National Laboratory (ORNL) (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 

which provides a basis for the numerical values assigned. The ORNL data 

were projected to obtain a performance value expected over an extended 

period of operation. It is considered that attempts to extend the service 

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 

radionuclide concentration of samples by that of the average outlet 

concentrating for each nuclide. 

Based on the data in References 4, 5, 6, 32, 42, and 43, the OF 

used for the parameter was that considered to be representative of the 

data. 

2.2.19 DECONTAMINATION FACTORS FOR EVAPORATORS 

2.2.19.1 Parameter 

Decontamination Factors 

All Nuclides 

Except Iodine 

Miscellaneous radwaste evaporators 103 

Boric acid evaporators 103 

Separate evaporator for detergent wastes 102 

2-79 

Iodine 
2.2.19.2 Bases 

The decontamination factors for evaporators 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 

(Ref. 4, 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). 

Average OF's for Zion, Ft. Calhoun, Turkey Point, Rancho Seco, and 

Prairie Island, were obtained by dividing the average inlet radioactivity 

of samples by the average outlet radioactivity of samples for each 

radionuclide. 

Based on the data given in References 4, 5, 6, 33, 42, and 43, the 

DF used for the parameter was that considered to be the most representative 

of the data. 

2.2.20 DECONTAMINATION FACTORS FOR LIQUID RAOWASTE FILTERS 

2.2.20.1 Parameter 

A OF of 1 for liquid radwaste filters is assigned for all radionuclides. 

2.2.20.2 Bases 

Reference 34 contains findings 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 ORNL report is that no credit should be assigned to liquid radwaste 

filters (OF 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 1/2 

(Ref. 42) indicate that decontamination factors in liquid radwaste 

filters vary widely from less than 1 to greater than 50 (with a mean value 

of 1.3). Therefore a OF of 1 for liquid radwaste filters is used. 

2.2.21 DECONTAMINATION FACTORS FOR REVERSE OSMOSIS 

2.2.21.1 Parameter 

Overall OF of 30 for laundry wastes and OF of 10 for other liquid 

radwastes. 

2.2.21.2 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 volume processed through the membrane to the inlet batch 

volume is the percent recovery. The OF normally specified for the 

2-80 
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 OF (OF). For source term calculations, the 

system OF (OF) should be use~. The system OF is the ratio of the 

nuclide conce~trations in the feed stream to those in the effluent 

stream. The relationship between the system OF and the membrane OF is 

nonlinear and is a function of the percent recovery. This relationship 

can be expressed as follows: 

where 

OF is the membrane OF; m 

DFs is the system OF; and 

F is the ratio of effluent volume to inlet volume (fractional 

recovery) • 

Tables 2-33 through 2-36 give membrane OF's derived from operating 

data at Point Beach, Ginna and Robinson (Ref. 35) and laboratory data on 

simulated radwaste liquids (Ref. 36). These data indicate that the 

overall membrane OF 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 OF is approximately 30. 

OF s = ___ 0-,-._9_5 --.--:-:;~ = 30 

1 - (1 - 0.95) 1/100 

The data used were derived mainly from tests on laundry wastes. 

The OF 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 OF for these isotopes is lower than the average membrane OF 

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 
TABLE 2-33 

REVERSE OSMOSIS DECONTAMINATION FACTORS, GINNA STATION 

Concentrate 

Act i vity Product Activity 

Nuclide (llCi/cm3) (l1Ci /cm3

) Membrane OF 

Ce-144 2.68 (-4) <2.2 (-7) 1200 

Co-58 8.55 (-5) <3.4 (-8) 2500 

Ru-l03 5.83 (- 5) <5.5 (-8) 11 00 

Cs-137 4.09 (-4) 6.6 (-6) 60 

Cs-134 2.02 (-4) 3.2 (- 6) 60 

Nb-95 5.35 (- 5) <5.3 (-8) 1000 

Zr-95 2.36(-5) <3.7 (-8) 640 

~ln-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) 219 

Gross beta 1.63 (-3) 1.86 (- 5) 88 

TOTAL 3.78(-3) 2.84(-5) 

Average 133 

2-82 
TABLE 2-34 

REVERSE OSMOSIS DECONTAMINATION FACTORS, POINT BEACH 

Concentrate 

Activity Product Activity 

Date Time {JJCi/m1} {l!Ci/m1} 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) 213 

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 

1530 1.8 (-4) 5.7 ( -7) 316 

TOTAL 9.8 (-4) 7.0 ( -6) 

Average 140 

2-83 
TABLE 2-35 

REVERSE OSMOSIS DECONTAMINATION FACTORS, 

H. B. ROBINSON NO. 2 STATION 

Co-60 Co-58 

264 29 

382 

436 

107 229 

76 490 

94 131 

Average 227 220 

2-84 

1-131 

14 

20 

39 

26 

96 

11 

34 
TABLE 2-36 

EXPECTED REVERSE OSMOSIS DECONTAMINATION FACTORS 

FOR SPECIFIC NUCLIDES 

Concentrate 

Activity Product Activity Nuclide {llCi Im1 } {ll Ci Iml} Membrane OF 

Co-60 2.5 (-4) 5 ( -7) 500 

Mo-99 3.8 ( -2) ( -3) 40 

1-131,132, 1.2 ( -1 ) 4 ( -3) 30 

133, 134, 135 

C5-134, 137 4.3 { - 2} 2 {-4} 200 

TOTAL 2 (-1 ) 5 ( -3) 

AVERAGE 40 

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. 

3. The discharge time is one-half the time required to empty 

the final liquid waste sample (test) tank to the environment. 

This value is based on the maximum rate of the discharge 

pumps and the nominal tank volume. 

The calculated values in 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 

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. urit waste - Processed through an ~~~~~------~~~~~~ if available) using a discharge 

fraction 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 

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 
For example, a OF of 105t was calculated for a radwaste deminera1izer 

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 OF can be calculated 

as follows: 

1. For 6 days (7 - 2 + 1) out of 7 the system OF would be 105

• 

2. For the remaining one day, the system OF would be 102 (only 

the deminera1izer OF is considered). The effective OF is: 

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. 

t 103 (Evap.) x 102 (demin) = 105 is obtained using OF's from Section 

2.2.19.1. 

2-87 
TABLE 2-37 

FREQUENCY AND EXTENT OF UNPLANNED LIQUID RADWASTE 

RELEASES FROM OPERATING PLANTS* 

Unplanned Liquid Releases 

Total number (unplanned releases) 

Fraction due to personnel error 

Fraction due to component failure 

Fraction due to inadequate procedures or failure 

to follow procedures 

Fraction due to other causes 

Approximate activity (Ci) 

Fraction of cumulative occurrence per reactor 

year (plants reporting releases <5 gals of 

liquid waste/reactor year) 

Fraction of cumulative occurrences per reactor 

year (plants reporting activity released 

)0.01 Ci/reactor year) 

Activity per release (Ci/re1ease) 

Activity released per reactor year (Ci/reactor year) 

Volume of release per reactor year (gal/reactor year) 

62 

0.37 

0.42 

0.08 

0.13 

24. 

0.16 

0.28 

0.39 

0.16 

633. 

* Values in this table are based on reported values in 1970-1977 Licensee 

Event Reports representing 154 reactor years of operation. 

2-88 
2.2.24 ATMOSPHERIC STEAM DUMP 

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 

noble gases and radioiodine released to the atmosphere from steam 

dumps because of low-power physics testing and turbine trips from full 

power. The evaluation indicates that the iodine-13l 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 control rods are inserted and 

during the time required to recirculate one primary coolant volume. After 

this time, the turbine bypass will be adequate to handle steam generated 

from decay heat. The quantity of steam released 

= (1.5 x 107 lb/hr)(60%)(14 sec)(2 trips/year)(454 9/lb)(360~rsec) 

= 3 x 107 g-steam/yr 

The iodine-131 concentration i~Bthe main steam for a U-tube steam 

generator is approximately 1.B x 10 ~Ci/g-steam from Table 2-2. 

Based on the steam release calculated above, the associated iodine131 release is approximately 6.0 x 10- 7 Ci/yr. 

I-13l/yr = (3.2 x 107 g-~~eam)(l.B x 10-B ~Ci/g-steam)(10-6 fit;) 

= 5.B x 10-7 Ci/yr 

2-B9 
Releases due to low-power physics testing are calculated based on 

one 10-hour release of steam each year following a refueling. For a 

RESAR-3 reactor, low-power physics testing is conducted at 5% 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 Ci/yr using the following equation: 

7 I-131/yr = (1.5 x 10 lb/hr steam)(0.05)(454 g/lb)(lO hr/yr) 

(1.8 x 10-8 l1Ci /g-steam) exp [_-~~;d:->--:--~'-'S'-L..] 10-6 Ci /l1Ci 

I-131/yr -6 = 4.6 x 10 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 envi ronment. 

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 ale (Ref. 37) 

found that the carbon-14 reacts to form volatile compounds (principally CH 4, C2H6, and CO2) that are collected in the waste gas processing system through 

degassing of the primary coolant and released to the environment via the 

plant vent. Data 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 

atmospheres. 

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 

Rancho Seco indicates that the release of carbon-14 breaks down to 22.6% 

from 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 
TABLE 2-38 

CARBON-14 RELEASE DATA FROM OPERATING PWR's 

Plant* 1975 

Conn. Yankee 44 

Yankee Rowe 1.6 

Plant** 

Tu rkey Poi nt 3/4 

Fort Calhoun 

Zi on 1/2 

Prairie Island 1/2 

Rancho Seco 

1976 1977 

40 30 

0.13 0.24 

Area 

Aux. Bldg. 

Containment 

WGPSt 

Spent Fuel Area 

Total 

Fuel Pool and Aux. Bldg. WGPS 

Containment Bldg. 

Total 

Cont. Bldg. 

1978 

70 

0.33 

Fuel Handling and Aux. Bldg. 

WGPS 

Total 

Cont. Building Fuel Handling and Aux. Bldg. WGPS 

Total 

Cant. Bu il di ng 

Fuel Handling and Aux. Bldg. 

WGPS 

Total 

Average 

* Based on semi-annual release reports. 

** Based on In-Plant Source Term Measurements. 

t Waste gas processing system. 

2-91 

Annual Average 

Ci /yr-unit 

46 

0.58 

Annual Release 

Ci /yr-unit 

2.4 

0.075 

0.82 

0.38 

3.7 

0.30 

0.81 

0.78 

1.9 

1.8 

1.4 

0.062 

3.3 

0.016 

3.3 

0.25 

3.6 

0.9 

1.85 

0.85 

3.6 

7.3 
TABLE 2-39 

DISTRIBUTION OF CARBON-14 RELEASED IN GASEOUS EFFLUENTS 

Plant Areas: Aux. B1 dg. and 

Plant Containment Fuel Handling WGPS 

Turkey Point 3/4 2% 75% 23% 

Fort Ca 1 houn 41% 16% 43% 

Zion 1/2 55% 43% 2% 

Rancho Seco 25% 51% 24% 

Prai rie Island 1/2 0.5% 92.5% 7% 

Average: 22.6% 61.0% 16.4% 

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 PWRls. 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 (1/2yr) 0.47 

Haddam Neck 1973 0.044 

1977 0.08 

1978 (1/2yr) 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 (1/2yr) 16.5 

D. C. Cook 1978 (1/2 yr) 19.7 

Tu rkey 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 (1/2yr) 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 

Fort Calhoun 1975 

1976 

1977 

1978 (1/2yr) 

Palisades 1978 (1/2yr) 

Zion 1/2 1978 (1/2yr) 

Prairie Island 1/2 1975 

1976 

1977 

1978 

Kewaunee 1976 (1/2 yr) 

1978 (1/2 yr) 

Three Mile Island 1 1975 (1/2yr) 

1976 

1977 

1978 (1/2 yr) 

Calvert Cliffs 1976 (1/2yr) 

1977 (1/2yr) 

Rancho Seco 1977 

1978 (1/2 yr) 

Release 

8.2 

2.2 

2.3 

0.27 

0.01 

24.8 

1.3 

21 

31.8 

13.5 

30 

5.9 

50 

12 

66 

46.5 

2 

3.1 

9.8 

1.8 

* All data provided by the semiannual effluent release reports and the 

annual operating reports for each PWR listed. 

2-95 

CHAPTER 3. INPUT FORMAT, SAMPLE PROBLEM, AND FORTRAN 

LISTING OF 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 problelil 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 

Shim Bleed 

Equipment Drains 

Pump seal leakage 

Pump seal leakage (Table 2-26) 

3-1 

Gal/day 

1440 

300 

20 

PCA Fraction 

Code applies the 

CVCS DFs internally 

1.0 

0.1 
A. Waste Generation Rates and Effective PCA Fractions (Cont'd) 

Waste Type Gal/day 

Other primary coolant leakage 

from miscellaneous sources inside 

the containment (Table 2-26) 10 

Total equipment drain wastes 330 

Clean Wastes 

Primary coolant equipment drains 

(Outside containment) 

Spent fuel pit liner drains 

Primary coolant sampling system 

drains (segregated from 

secondary coolant samples) 

Total Clean Wastes 

Di rty Wastes 

Primary coolant equipment 

80 

700 

200 

980 

Reactor containment cooling system 500 

Auxiliary building floor drains 200 

Secondary coolant sampling system 

drains 1400 

Total Dirty Wastes 2100 

Regenerant Wastes 3400 

Condensate demineralizer rinse 

and transfer solution (secondary 

system wastes) 12000 

B. Available Equipment for Liquid Wastes Processing 

Equipment 

Recycle holup tank 

(To collect shim bleed and equipment drains) 

Clean waste holdup tank 

Dirty waste holdup tank 

Regenerant solution receiving tank 

3-2 

PCA Fraction 

1. 67 

0.97 effective 

1.0 

0.001 

0.05 

0.093 effective 

0.001 

0.1 

0.0001 

0.01 effective 

Code i nterna lly 

calculates the 

buildup on the beds 

Number 

Capacity 

(Each) 

2 

2 

2 

2 

50,000 gal 

7,000 gal 

10,000 gal 

20,000 gal 
B. Available Equipment for Liquid Wastes Processing (Cont'd) 

__________________ E_q~u_i~p_m_e_n_t ________________ _ 

Resin and transfer solution recelvlng tank 

(To collect secondary system condensate 

demineralizer resin and transfer solution) 

Clean waste monitor tank 

(For processed shim bleed, equipment drains 

and clean wastes) 

Dirty waste monitor tank 

Secondary waste monitor tank 

(For processed regenerant wastes and 

secondary system condensate demineralizer 

resin and transfer solution) 

Recycle feed demineralizer 

(To process shim bleed and equipment drains 

and located upstream of the recycle holdup 

tank) 

Recycle evaporator condensate demineralizer 

Evaporator condensate demineralizer A 

(For clean wastes) 

Evaporator condensate demineralizer B 

(For dirty wastes) 

Secondary waste evaporator condensate 

demineralizer (To process regenerant wastes) 

Secondary waste demineralizer 

(To process secondary system condensate 

demineralizer resin and transfer solution) 

Steam generator blowdown demineralizer 

(To process steam generator blowdown) 

Recycle evaporator 

(For processing shim bleed and equipment drains) 

Radwaste evaporator 

(For processing dirty wastes and clean wastes) 

Secondary waste evaporator 

(For processing regenerate wastes) 

C. Additional Notes about Liquid Wastes 

Number 

2 

1 

2 

2 

Capacity 

(Each) 

20,000 gal 

10,000 gal 

10,000 gal 

10,000 gal 

50 GPM 

50 GPM 

50 GPM 

50 GPM 

50 GPM 

50 GPM 

2 in 300 GPM 

series 

30 GPM 

30 GPM 

30 GPM 

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 demineralizers 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 0-150 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 vie'; 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 
Number of pressurized storage tanks - 4 

Volume of each tank at STP - 650 CF 

Design pressure for each tank - 150 psig 

No recombiners 

3. Containment has small diameter (8 inches) purge line and the 

low 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 - O. 

6. Fuel, auxiliary and containment buildinqs 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. 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 OF 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: 

1 • Fi 1 e 1: 

2. File 2: 

A FORTRAN listing of the liquid effluent code. 

Nuclear data library for corrosion and activation 

products for use with the liquid effluent code. 

3-5 
w 

I 

0"1 

FORTRAN CODING FORM I -----------.. ----._-_. ,------

CODER I DATE jADDRESS 

PROBLEM TITLE PWR-GALE CODE PROGRAM NO I CHG NO_ 

PHONE 

SHEET O~ 

STATE MENT 

NUMBER 

~ 

C, AI RI 0, 

CIAi!sIDI 

CIA I R I 01 

C, A, R, QI 

~ 

WiRill 

0' ZERO 

~:ALPHAO 

I : ONE 

I: ALPHA I 

2; TWO 

l: ALPHA ISE!l.TIFICAT'ON 

21 ,P,OIW,T,H, , L1U1LE~IM,A,L, ,P,O,W,E,R, II,E,V'E'L! '('M'E'G:~:~:~:~:~:S')"::"':::::::"::::::::: I~::'::: :~ 

I!~ :L:~:~:~:~:N:: :~:::~:~:A:~:;: :~:~:~:~~::L:E~T~~~:W~N::R:~:;:::O:~:~~~~~~~ ;L:BISill I I ~~~lll Llll'I'lll') 

5, IC,B,E,I,R, ,L,EIT,D,O,W,N, IC,AiI, l,c) til IDIE,M'],N,E,RiA,L, I,Z,E, Rl IE,L10,Wl ,iLGtl'd1 l" ,.1 1 1 I' , , 1 .. 1 , , , , , iCt 11' , , ,) 

1611NIOIGIEINII ,N,UIM,BIE,RIIOIEI ISIIIEIAIMIIGIEINEI~A1QR3111111111 11I1I 1II11 I II :(,11 

71IT,OISITIEIU IIIOITIAIII ISIIIEIAIMI IFIIIOIW! 1(IMIIILIL,IIOINIILIBIS,/IH,!llll II I I I I I I I I II I I I I I II I' :(, I I I' Jl 

8, ,W,Ld, , , , , ,M,A,S,S, ,O,E, ,L']IO,U,],D, !I,N, ,E,A.r~If' ,'l TI1:; AM ,G,E,N,E,R,A1T,01R, '(,T,H,O,U,S,A.N,[j ,U B S "1("",, Ll 

1911BILIWIDIWINI IBILIOIWIDIO,WINI-ITIHIOil.iSIILIBI/IHIRI(1 II I I I)IBILIOIWIDI~WNI ITliSIEIAITIMIEINIT, iIINIP,lJITI IO""ldO,R,21 () 

, 0, ,R,E,G,E,NLTI I IC,O,N,D,E1N,S,A,T, fJ ,D,E,M'],N,E,R,A,L.I,Z.fJ Rl ,D,",r."'''''"'D,AlTIl'I"\''" ,T,I,M,E, ,( ,D,A,Y,S,), , , , , , , (,'" ,l 

CAR 

h&&Q 

b.fufu.Q 

,,:~: 1,'12: ~~~~~: : : :~:~:~:~:E:~:~:!:::~: :Dt~M~IE:N:::R:A:L:I:Z:E:i :~:I :O:W:_:F:R.:;:C:~:;:::N::: :": " " " : I I : I I'::: ::: 'j 

'131 I I 1 I I I I I ID,Fdl=d' I I I I I I),D,FIC,S,-d, I I I . I Il,D,FIOI 1-.d.Ll l.~ L 1 LL:~"---: L..: L..: L..: L..: L..: +-: .L.: .L.:.L.:.L.:.l.-:.L.:.l.-_ -'-: -'-: +-: -'-: +-. -L.: -L.: -L.:...L.:...L.:...L.:-'-l:j 

CIAI iSl DI 14"", I, " IClJILILIEIC,IilIO,N''(1 I I I,ll ,D,A'YISI ,PIR,O,r.,FISISI ,(", I ill ID,A,Y,S, IF,R,AIr.,I, ,DrI,S,C,H, ,( ,) 

C,A,R,D '~' 'I' I I I IEIQIUI~IPIMIEINITI IDIRIAdINISI_,IINIPil.hTI 111(" "" I)' IG,PIDI" (, I') ,P,C,A, "I' 

C,A,R,D, 1'_LLL I 1 ~ L lD1FlI1-'~_LLl l)IDIFIC,S'-~(1 1~.-,-_IlLD1F,O, ,=11, , , , , , ,) 

l.....J.1:U.tI..ill'"'-~f-4--L--L+...L-L-L-J--L-'-'''-'-''UJ.J..I..~L\<l-ll-.ll-'olL!li--'-LL--'--+--L-U'-L-liLLill-ll-->lL-L!+''I!....l.!u.<-'-'---'''--'''--'LL-'-+-'--"-'--L--'-'<-D liAll2Y-,-"S-'--L'F-t-"RULlAL' J"LlT ~~ HI ,( I -'-" ), , , -'<.L!:l.L"'-'''-+-Yf'-4--'--L+--L--'--l-J--L-"''--''''-'-=-'-'-'-l-'-'-'L-l-''-'-'-'-'-'''-L1-'--''~Lli-,+-'-L,",-''---'--L--LLJ I I ,(, I , , I , I) , I , '" , I I I) I ,P I C ,A, I ' I I ' I 

l-"-'-n.LlllJ"-'- +-'-l-"-I---'--'-+--"--'--L-L--'--L-"L-'-'--"-+'-L-'--L-'--'----"--'-L-'--"'JLCf-"-L-"'L--'L-'--L-L-l_·_·L-j hili F,O I I = I (, I I I I I I I 1 I , : .--LL I: I' i I 

C,A,R,D, 12 :01 I I I' I I ICIOILilIEICIIII,O,NI 1(1 I 1,,)1 IDIA,y,S, ,PIRIO,OE,SiS, 1(' I I I ,), ,D,A,Y,S, ,EIRIA,C,T, ID'I:SiC,H, d!", :)'" ) I 

I

C: A: R: D: 121": I' ,D,I,R,TIV' ,W,A,S,T,E, 1l1t11Pd1T", I I:' i,l"" ,), '('UP~.L...l...l....I..J...(~L..L.I-LJLJ'..L:..A ";, I ." '. 

CARD ,22 II IDIFdl=I(,IIIIII)IDIEIC,SI-I(llllllll,DIFIOII=I(111 1 ')'" I II 1'1'" I' II ii' 't 

C,A,R,D, "lll I I I I..L1C..LQ1llI IEICITIIIOINi 1(1 III ,D,A,Y,S IPIRIOICIEISISI 1(1 I 1-'---!)·lDIA1YS .FR.A.JC1.T11D1IL5.lCLHL-J(+J i '; :)::: I :~~~.r' 1.l....l._LLLI I I I I I I I I I I LLI I I I , I I I I I I , I I I I I I I I I I I I I I. L.L.L.L_LLL_LJ_-f I I' _____ . _"_ 1 ~J 1 Ll._.l.. ___ L-.l_L I I , , I I 'I I' I I 1 __ L.L..+-_L_L.L.l..L.L_c LL I I I , I'Ll 11 +.Lr

·L..c_LLi...L1-

-'---'.J.J.4_t-t-.l.._l.+-LLLL_Ll. ".- 1...L.1 L1.1 1 l.~ _1.L.L . .LL--LL .. _1 I' , , I I ',1 '~_LL1<_L<1< 1<,1..1<_1.1. I. I I.' ' .. 1._'. I ' •• I.' . .!.,I __ ' I •. '.p ._. :: i 3 I .: I: , ~ ! -. !:- ,:::l I j ~ I ;1 ~ I :1 6] I 5, 912vl~ [22!23j24j25126,r'j2c!J29)3vJ3 ,32!~3j34!35,36!37!38j39!40!4: ,42,4-,44145146,4 !4~4;ry5-'j5 ,52(:3!~"5-'1 .... 6jv ,58,-'S'o .... jt jt2,c;'5 ... ,cS!ct,1:: r:;8,69j. '-I ~j -I 4!~! 01 .! 81 9J1j~ 
w 

I 

-.....J 

CODER 

PROBLEM TITLE PWR-GAlE CODE 

STATEMENT 

NUMBER 

0' ZERO 

i;l'ALPHAO 

FORTRAN CODING FORM 

1 DATE IADDRESS 1 PHONE I PROGRAM NO. I CHG. NO. I SHEET 01'" 

IDENTIFtCATION ONE TWO 

ALPHA I ~: ALPHA Z 

I 12 I 3 I 41 516171 8 I 9110 I' '1'21'31'41'5 11611711811912~2' 122'23124125126127128129t301~ I i3213313"35136137(38139t4~4'142143144t45146(4714S!4915015' 152,53154155156157158159160161 ,6963164165166167,6816917017' 172173174175176177178,79180 

~ 

~ 

~ 

~ 

~ 

~ 

CIAI RI D 

~ 

CIAI RI DI 

~:R:DI CAR D 

C IAI RI D' 

Lfufu..Q 

~ 

C,A,R, P 

2141 I 'BllIOI~IDIDIWINI I IF,RIAICITdlO,NI 'P...:.R..ill..1C..I.E.jS1S..I.ELD, 1'1'1 i I 1'1"'" I" 1'1'1" I I 1(, I I 1)1 

2 51 L.t..!? iF I I I -I (I , , , I , tll D, F I C, S I = ,(, I , , , I I). P , FlO I I =.r, , , I ' I d 

2\61 I ICIOliliIEICITII,OIN! 1(1 I I I 1)1 IDIAIYISI IP~RIO'C'E,SISLilL~, ll.L~-' IFIRIAIC,TI ID,IISICIHI ,(i, I I I) 

2171 IRIEIGIEINIEIRIAINITI IEILlolViI IRIAITIEI I(IGIPIDI)I I I I I I I I LL.L.J. I I I I I I( 

2 8, , I , , , , , , ,p I F tl ,=", , I I I I ,) ,D IF IC 15 1= d I I 1 1.-' I .1 ID I E 101 I = Ii I I I tl 

12191 LL.LL.Li L L LCIO I l I II ELC1ILll ClLfu_..l(1 I I I l) I lOlA IY IS I IP IRlllLCl EI 515, dLLL I I) 1 illl1l.j'L15.J IF IR I A Ie I II 101 T, SL~ L 

3101 1111111 I illSI iTIHIEIRIEI IC/OIMTIqNIIOI!9 ISIIiR1TIPlPlIiNlr., IOU:Lt1'1111111 IIlc:t"D.r1QWII tFIl,DilI,?, IQj~LlLJnIRI21 

I I I I 

~-""-L:.J."'+'!.LLL-L..LUJ..LilLJ"-L-.LD-'4-"'---'I--,K!",-"Il!LlJ.'tl--,AL..l]LlJO N I ( I D I A I Y , 5 I ) I I I I I I I 

I 51 

I I 

) 

10 

1 

.Ll 

it 

~~~~~-L--'-~~~~~~~~~L-LW~UU~~-L-L-L..~~~~~Ud1L~7~ .. J(~I--'I~)~1~IHul~E~IP~~IAul~7~1(~ILjI~)1.-L~-L..L--'-L-L--'-L-L--'-+-L+-~~-L~L-L1 

~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~E~T~.llL~+~~L-L--'-L--'-~~-L--'-LJ-L~L-L+-L~L--'~-L--'-L--'+ 

10 II R A ti II IC,H,AIRIC,O,A,l,?dl 1)1 IHIEIPIAI?"I 1)1 IRIAITIEI(llIOIOIO,C.l..F~M..l)~L. I I I I I 

C,A,R,D, 139, , , , , , , , , ICIN,T,M,T,-,H,T,r.:,HI IVlnt II ,PI II a Q 0 Ie HIAIRIC10IAlil?ifL111ili IEI P,A1?lCL..ll, , INllllMiBiELR,/,Y,E,AiR, 1 I 11 LL ,) 

C,A,R,D, i4llli , , , , , , , , ,C,N,T,M,T,-,L,O,W, IJJ IQ.lll ,P1UdhJil.L_l jCLHL'~.lB~IOIA,L,?dl 1)1 IHIE,P,AI?'(I IL ,R,AIT,EI '(,C1FlMdL LLh, 1 i' , , 

C,AIRIPI [41111FIVINI II IF.JR.1AICITtIIOINt IIIOIDillNIEI IRIEILIEIAI51Ei1lliFIRIOIMI IBIIIOIWjOO\jf:1 ITIAINIKI IVIEINITI I I I I I i I 1(, I) 

C,A,RlDl 42, IF,ElJl 1..1 l.l IP,E,RICIE1NLli ,OIFI il,OO lt~&, R HM,O.VIEdJ' ,F!I~j(AM lAltlt tEW,EIC,T,OLRI IR,Ell,EIAIS,E, , , , I I i(, , , , , , ,) 

CIAIRID I 14131 IPIFlllAIUINI IDIEITIEIRIGIEINIII IViIAlSiljEI IPIFI I I J~l-LLl...1 II I I I I I I I I I I I I I I 1(1 

i 

I L I I j I -'--1 I LLi...L.L.J.. I L.J. I I I 1 I I i , I l I 

I 

1::::11:::::11 III1II LLL+i.IIII-'--L.;IIIIIIIIIIIIIIIIIIII+--,-111111 II :.i. I I I I I I I I I I I 1 I I I I I '--'--J--LLLL LLLL.LLL I I I I I I I 1 I I I I I I I . I I I ' I I I I I 

I 

t ~l..~J....L~-+-+~~~~L-'LL_1 I I I I I I I I I ..L-LI-LI~~_~ __ L..L..L..~~J..-L-L-L4--L~-L-L....I.~~ __ L-L-.j 

I I I I I ' I I I I I I : I I I ' I I I ' , i I I I I I I I I I "1.1 I I I I I I I I I I I 

2 3 4 :. 6 .., 8 9 -0, '21314 15 d~ 17,8192021222324252627282 

CARD 1 NAME 

CARD 2 POWTH 

CARD 3 PCVOL 

CARD 4 LETDWN 

CARD 5 CBFLR 

CARD 6 NOGEN 

CARD 7 TOSTFL 

CARD 8 WLI 

CARD 9 BLWDWN 

CARD 10 REGENT 

CARD 11 FFCDM 

CARD 12 

CARD 13 

CARD 14 

CARD 15 

CARD 16 

CARD 17 

CARD 18 

CARD 19 

CARD 20 

CARD 21 

CARD 22 

CARD 23 

CARD 24 

CARD 25 

CARD 26 

CARD 27 

CARD 28 

CARD 29 

CARD 30 

CARD 31 TAU1 

CARD 32 TAU2 

CARD 33 TAU3 

CARD 34 

CARD 35 

CARD 36 

CARD 37 CONVOL 

CARD 38 

CARD 39 

CARD 40 

CARD 41 FVN 

CARD 42 FEJ 

CARD 43 PFLAUN 

NAME OF REACTOR SAMPLE PWR REV 1 TYPE = PWR 

THERMAL POWER LEVEL (MEGAWATTS) 3400. 

MASS OF PRIMARY COOLANT (THOUSAND LBS) 550. 

PRIMARY SYSTEM LETDOWN RATE (GPM) 75. 

LETDOWN CATION DEMINERALIZER FLOW (GPM) 7.5 

NUMBER OF STEAM GENERATORS 4. 

TOTAL STEAM FLOW (MILLION LBS/HR) 15. 

MASS OF LIQUID IN EACH STEAM GENERATOR (THOUSAND LBS) 112.5 

BLOWDOWN-THOUS LB/HR 75.0 BLOWDOWN TREATMENT-INPUT 0,1, OR 2 0 

CONDENSATE DEMINERALIZER REGENERATION TIME (DAYS) 8.4 

CONDENSATE DEMINERALIZER FLOW FRACTION 0.65 

SHIM BLEED RATE 1440. GPO 

DFI= 5.0E03DFCS= 2.0E03DFO = 1.0E05 

COLLECTION 22.6 DAYS PROCESS 0.93 DAYS FRACT DISCH .1 

EQUIPMENT DRAINS INPUT 330.0 GPO AT 0.97 PCA 

DFI= 5.0E03DFCS= 2.0E03DFO = 1.0E05 

COLLECTION 22.6 DAYS PROCESS 0.93 DAYS FRACT DISCH .1 

CLEAN WASTE INPUT 980. GPO AT .093 PCA 

DFI= 5.0E02DFCS= 1.0E03DFO = 1.0E04 

COLLECTION 5.7 DAYS PROCESS 0.13 DAYS FRACT DISCH 0.1 

DIRTY WASTES 2100. GPD AT 0.01 PCA 

DFI= 5.0E02DFCS= 1.0E03DFO = 1.0E04 

COLLECTION 3.8 DAYS PROCESS 0.19 DAYS FRACT DISCH 1.0 

BLOWDOWN FRACTION PROCESSED 1. 

DFI= 1.00E03DFCS= 1.00E02DFO = 1.00E03 

COLLECTION 0.0 DAYS PROCESS 0.0 DAYS FRACT DISCH 0.0 

REGENERANT FLOW RATE (GPD) 3400. 

DFI= 5.0E02DFCS= 1.0E03DFO = 1.0E04 

COLLECTION 4.7 DAYS PROCESS 0.37 DAYS FRACT DISCH 0.1 

IS THERE CONTINUOUS STRIPPING OF FULL LETDOWN FLOW? O,l,OR 2 0 

HOLDUP TIME FOR XENON (DAYS) 60. 

HOLDUP TIME FOR KRYPTON (DAYS) 60. 

FILL TIME OF DECAY TANKS FOR THE GAS STRIPPER (DAYS) 30. 

GAS WASTE SYSTEM HEPA?99. 

FUEL HANDLG BLDG CHARCOAL?90. HEPA?99. 

AUXILIARY BLDG CHARCOAL?90. HEPA?99. 

CONTAINMENT VOLUME (MILLION FT3) CNTMT ATM. CLEANUP CHARCOAL?O.O HEPA?O.O RATE(1000CFM) 

CNTMT-HIGH VOL PURGE CHARCOAL?90. HEPA?99. NUMBER/YEAR 

CNTMT LOW VOL PURGE CHARCOAL?90. HEPA?99. RATE (CFM) 

FRACTION IODINE RELEASED FROM BLOWDOWN TANK VENT 

PERCENT OF IODINE REMOVED FROM AIR EJECTOR RELEASE 

DETERGENT WASTE PF 

3-7a 

2.45 

0.0 

1000. 

0.0 

0.0 

1. 
IjAMPLE PWR REV 1 

THERMAL POWE~ LEWEL (MEGAWATTS) 

PLANT CAPAClly FACTOR MA~S Of PRIH~RY COULANT (THOUSAN~ La~) 

PRlMARY SYSltM LbTOowN RATE (GPH) 

LlTDUWN CATIUN ObMINERALIZER fLO. Ce~H) NUHAER Of STtAM bENfRATORS 

!UIAL STEAM tLOW (MILLION LRS/HR) MA~S Of LIQUlO IN EACH STEAM GEN[RAT~R (THOUSANO LSS) MA~S Of WAT[~ IN STEAM GENERATOA~ (T~OUSANO LaS) ~LUwOOWN RAT~ CT~OUSANO LRS/HR) 

PRIMARY TO S~CONPARy LEAK RATE fLSS/~AY) 

t;U/jOENSATE O~MIN~RALIZER REGENERATION TIME (DAYS) 

FISSION PRO~UCT tARRY-OVER fRACTION 

HALOOEN CARRJ.OVfR fRACTION 

t;UNOENSATE U~MlNbRALIZlR fLOW fAACTIUN 

LlijUIU wASTE INPUTS 

f"ACTlON FRACTION CI,/LLECTION DEcAy 

STREAM fLOw RAT!! Uf PCA OISCtjARr,£O TIME UME 

(GAL/OUl (OAYS) (DAyS) 

ShU'" tlLEEP RATE 1.44E+O~ 1.0000 .1000 ai.bOOO .'300 

EQIlIPMENT f}PAINS 3.30E+oi .9700 .10011 i4!!.6000 .9300 

CLEAN WASTE INPUT 9.~OE.0~ .0930 .1000 !I.7000 .1300 

DIRTY wASTES 2.IOf+Oj .0100 1. 0000 ij.8000 .1900 

SLOWDOWN 2.16E+Oij 

y> UNTREATEO SlowDUWN O. 

0.000 0.000 0.000 

1.000 0.000 0'000 co REGENERANT SOLS 3.40E+0;' .100 4.700 .310 

GA5EOUS WASTE INPUTs 

T"~RE IS NOT cONTINOUS STRIPPINr, Of tULL LETDNN fLOW 

HOLDUP TIME tOM AENON (DAYS) - HOLDUP TI~E toR "RypTON (DAyS) 

fiLL TIME OF DEC~y T~NKS fOR THE GAS STRipPER (OAYS) uA~ WASTE SY~TEM PARTICULATE ~ELE~sE FRACTION 

AUXILIARY BLUG IODINE R~LEASE FHACTION • PA~TICUlATE ~ELE~sE FRACTION 

CONTAIN"'ENf YOlUHE (MILLION ~T3) - FHEaUENCy Of C~T~T BLDG HIBH VOL PUR!E (TIMES/YR) 

t;NfMT_HIGH V~l PWRGEIODINE RELEAbE F~ACTION 

PAHTICULATE ¥ELEAsE fRACTION 

tNIMT lOw VO~ PUMGE RATEICFM; - CNTMT LUw VO~ PURGE IODINE RELEASE F¥ACTION 

PARTICULATE RELEASE fRACTION 

STEAM LFAK TU TUMSINl ALOG (LBS/HR) - FHACTION IODINE ~ELEASEO FROM BLOwOuwN TANK VENT 

PERCENT Of IUolN! REMOVEO fROM AlR E~ECTOR RELEASE 

PWR 

340ri.oOOO 

o.flOOO 

550.0noo 

70;.0000 

7.5000 

4.0000 

10;.0000 

11;>.5(100 

45;;.0000 

'5.0000 

1<;.0000 

1'1.4000 

.0050 

.0100 

.6500 

DECONTAMINATION fACTORS 

I 

5.0nEoo1 

5.00E+03 

5.0nEo02 

5.00E o02 

1.00£003 

1.00E+oo 

5.0nEo02 

60.0000 

60.0000 

30.0000 

.0100 

.1000 

.0100 

;>.~500 

;>.0000 

.1000 

.0100 

1000.0000 

.1000 

.0100 

170';.0000 

Ii.oooo 

1.0000 

Cs OTHERS 

2.UOE+0", I.OOE+05 2.00£+0~ 1.0OE+00; 

1.0UE+OJ I.OOEo04 

1.00E+Oi 1.00£+04 

1.00Eool I.OOE+O] 

1.00E+01I 1.00Eoon 

1.00E+O. 1.00E+04 

FIGURE 3-2 

PRINTOUT OF INPUT AND OUTPUT FOR THE SAMPLE PROBLEM 
SAMPLE ~WR 'EV 1 L'~lIID EfFLUt::NlS 

ANNUAL R~LEASES TO DISCHARGE CANAL COOLANT ~ONC~NTRATIONS-_- ___ ~_--~ _________ - ___________ : _______ ------~------ ____ ADJUSTED DETERGENT TOTAL NUCLIO~ HALF._LlfE PRlfoOAHY SECONDARy BOI~UN HS HI SC, WASTES SECONUARY TURB BLDG TOTAL LWS TOTAL WASTES 

(DAyS) (MICRO ,I/ML) (HICRO CI/ML) (CURIE!!) (CURIES) (CU~lE.S' (CURIESI (CURIES, lel/yRI (CI/YAI ICI/VR) 

CORROsiON AND ACTIVATION pHuoUCTS 

NA 24 6,20:;F-01 4,70E"0~ 1.~2E-06 ,00000 .00003 ,00000 .00005 .000n9 .00114 0.00000 ."0110 

P 32 1,~3E.Ol O. 0, a.ooooo 0,00000 0,00000 0.00000 0.00000 0.00000 ,00018 .0001" 

CR 51 2.7~E.Ol 3.10E .. 0;J 1.22E-07 .00000 .00001 ,OOOnO ,00001 .00002 ,00025 .00470 .00500 

MN 54 3,OlF..o2 1,60E"03 6,08E_08 '00000 .00001 ,00000 ,00000 .00001 ,00014 ,003110 .00390 

FE 55 9,50F.o2 1,20E"03 4,59E-08 .OOQUO .00000 .00000 ,00000 ,00001 .00nl0 .00720 .nono FE 59 4.50E,01 3.00E"0' l,12E-08 .00000 .00000 .00000 ,00000 .00000 .00002 .00720 .002211 

CO 58 7 .llF. 0 1 4,60E"OJ 1.78E_07 .000uo ,00002 .00000 ,OOonl .00003 ,00038 ,00790 .001130 

CO ~o 1.92E'03 5.30E"Oti 2,061:.-08 '001100 ,OOOOU .000no ,00000 .00000 .00005 .01400 .01400 

I'll 63 3.36E.04 0, O. a,OOOUo 0,00000 o,OOOnO 0,00000 0.00000 p,OOOOO .OOPO ,00170 

ZN 65 2.IIO:;E.02 5.10E"04 1.'nE-08 .00000 .00000 .00000 ,00000 .00000 .00004 0.00000 ,00004 

11181 9,Y6r-Ol 2.50E"0" B,221:.-08 ,OOOUo ,00000 ,00000 .00000 .00001 .00008 0.00000 .ooooe 

NP239 2.30:;E'00 2.?OE-03 1.90E-01I .OOUOO ,00000 .00000 ,00000 .onool • I!OO 11 o.ooono .00011 

FIsSION PkOUUCT~ 

SR 89 5.20E+Ol i.40E~01t 5,34E-09 .00000 .00000 .ooolio .00000 .00000 .00001 ,ooon9 .000)0 

SR 90 1.03f+Olo 1.20E .. 05 4,59E-I0 .ooouo ,00000 .000nO ,00000 .oonno .0noOo .00001 ,nonol 

SR 91 4.03E-01 9.bOE-0'l 2.1>7E_08 .ooouo .00000 ,OOOno ,OOOno .Oooon .00002 o.oonno .lInno? 

V .114 3.~1E-02 4.60E"01\ 3.16E-09 .001lUO .00000 .00000 .000no ,onooo .(10001 0.00000 .oonnl 

'( 91 5.iRE+ol 5.20E"Ot 1.'HE-I0 .noOuo .00000 ,00000 .00000 .Ooonn .onooo .00008 .00nn9 

'( 93 Io.2o;E-ol 4.?OE .. 03 1.14E_07 .00000 .000011 ,OOOno .00000 .00001 .00007 0.00000 .noo01 

lH .5 6.50E+Ol 3,90E"0' 1.50t:-08 .OOOUO .00000 .000no ,00000 .00000 .00003 .00110 .00110 w Nil 95 3.5liE.ol 2. 8 OE-01t 1.03E-08 .00000 ,00000 ,00000 .00000 .00000 .(10002 .00190 .001'10 o MO 99 2. 7qE'00 6.40E~0~ 2,35E-01 .00800 .ooool .00000 .00001 .000n3 .(10035 .00006 .00n41 

TC 99H 2.Slif-Ol 4.10E"0;' 1.06E-01 .00000 .ooool .00000 .00001 .00002 .00030 o.ooono .00n30 

RUIO] 3.YfoE.ul 1.50E,,03 2.90E-01 .00000 ,ooooJ ,00000 ,00001 .on005 .00062 ,OOn29 .000'11 

RHI03M 3.9"E-02 O. 0, .00800 .00001 .00000 ,OOOnl .00000:; .00061 0.00000 .000"1 

RUI06 3. 'ITE. 02 9,OOE"0~ 3. 46E_06 .000114 .000311 .00000 .00011 .000<;9 .00166 ,00'190 .01700 

RHI06 3.47E-04 O. O. .000U4 .00037 ,oooiio .00011 ,00059 .00166 o.OOnno .00770 

Ab l10M 2.5'lE,Oi 1.30E·0;J 4,96E_08 .000 0 0 .00001 ,OOOnO .00000 .onool .&0011 .00120 .00130 

AGllO 2.i;>E-04 o. O. .00000 .00000 .00000 ,00000 .ooono .00001 0.00000 .0nOOI 

SI:I124 b.Onf.ol !I. O. O.oOIlUo 0.00000 o.OOOnO 0,00000 o.oonoo 0.00000 .0004 3 .00n4:1 

TE.129M 3.~IiE.Ol 1.90E4 01t 7.30E-09 .00000 .00000 .000nO ,00000 .00000 .110002 0.00000 .00002 

TE.129 4.19E-02 2.40E~0~ 2.11E-01 .00800 .00000 ,00000 .00000 .001100 .eno02 O.Ooono .oonn? 

TE.)3)!! I.Z<;E+OO 1.50E.0.1 5.09E_08 .00000 .0000 0 .000no .000no .Oollno .00n06 o.OOnno .oono" 

1131 8.0<;[.00 4.'50E"0~ 1.36E-0 6 .000 11 .0030' .00Onl ,00013 .on34n .0442 9 .0011'>0 .04foOO 

TE1l2 3.25E+OO 1. 7 OE 4 0J 6.20E-08 .00000 .00000 ,Ooogo ,00000 .oono) .00010 O.OOono .00nl0 

1132 9.5I1E-02 2.10E,,01 2,71E._0l> .001100 .00019 .00000 .00005 .00024 .00313 0.00000 .OO~IO 

1133 8.1o;E-01 1. 4 OE"0' 3.}sl:.-06 .000U3 ,00290 .00000 ,00031 .0032'5 .04226 0.00000 .042no 

1134 3.IHE-02 3.40E .. 01 2.21E-06 .00000 .00001 o.ooono .00000 .00002 .000 2 1 o.ooono .no021 

CSIH 7."qE.02 1,10E .. OJ 2. 8 41::-01 ,000:'0 .0002Y .00000 .00001 .000111 .0106 0 ,Ollno .0;'200 

1135 2.T9E-OI 2.60E .. 01 5. 4 7E-0 6 .00000 .0013a .00000 ,00029 .00163 .02122 0,00000 .02111n 

CSI3" 1,;1oE.01 8.70E~0' 3. 45E_01I .00803 ,ooool ,00000 .00000 .onn01 .0008 9 .00037 .00130 

CS137 1.1iiE'0. 9.40E"0~ 3.79E-01 ,00'~8 .00039 ,OOono .00002 .onl09 .01413 .01600 • .,3000 

BAI31M 1,71E-03 O. O. .000113 ,00037 .000~0 .00002 .001n2 ,0)322 0.00000 .01300 

IUl40 l.lRE'OI 1.30E-0~ 4.81E-07 .00000 .OOOO~ ,OoonO ,00002 .OOOOR .0009 8 .00n9 1 .001'10 

LA140 1.6RE.00 2.50E·ol 8.'''1:.-01 .000uo .0000 1 ,OOOno .00004 ,OOOI? .00)55 0.0001'0 .00160 

CEI41 3.24E,01 ).50E-0" 5.'IE-09 .00000 .00000 ,nooon .000no .onoon ,onool ,oon23 ,00024 

Cl143 1.38E·00 2.ROE~Ol 9.1031::-08 .000uo .00000 ,00000 ,00000 .oonlll .00011 o.ooono .00011 

PHU3 l.l1E,01 O. O. .00000 .00000 .000no ,000(10 .onono .00nOl 0.00000 .nonni 

CI::144 Z.94E.02 3. QOE"OJ 1.50E-07 ,00000 .00002 ,00000 ,00001 .000n3 .00033 .00390 .OO4?1I 

PR144 I.ZiiE-02 O. O. .00000 ,00002 .00000 ,00001 ,00003 .nn033 0,00000 .00113:1 

ALL OTtilHS 2.14E~0' 6.26E-07 .nOOuo ,00000 ,OOUnO ,ooono .oonnn .nooOl O,oonno .nonol 

TOTAL 

(EXCEPT TR IT IUM) 1.4BE.00 2.52E-05 .00216 .00972 ,oOOnl ,00141 .01111 ,17331 .n8915 .:o60no 
III 

I 

"" C 

Z 

c a:: 1: c 

"" >- III 

III 

X "" !oJ ..J CL III 

III 

W 

... W a:: :::l :::l ..J U C 

:> 

WI Q l: II) ~ 

N 

-c 

% 

~ 

III 

"" ~ 

C 

U -:l W Z III ... C 

"" C 

~ Q 

"" Q 

a:: Q 

C 

% 

:::l -~ 

... "" ~ 

IX 0 ~ z 

3-10 
W 

I 

Lltmll> iljAS1F 

5rl-ltAM 

!J""Pl.E ~''''n Rl v 1 

lHt.IIMIIL powEll lt~[l (11UM\oIlTT!') 

PllltH (;t. ... IIClly t'ACTnJ( 

,"/\::,,, of pprl~iI ... 'I' l·OuLAIIT ITliOI.lSIlfHl Lu:', 

.... tllMAlly c:;y~II:.~1 I.bTIJo.,tI ~ATf (.,PM' 

llTIlU\J1II CArlul' LJttlltllHAt.llll' nllilj (b'.' ... ) 

NlJt~Htp ('If SH·A" vHIEH,\ r.)HS 

filiAL S'£I,~~ tt U ... (MflLIorl UIS/HIO 

, ... ~",<; Of L I(JU~O IN [AI..H 5 Tr:1\l1 r;trl~~A run (TlIUtJ5ANfl LtJS) 

o1LOwlJ01I'1 r!A It: (TtIlIIl<;MIU lIl;;/tlR) 

LU,lI)ErlSIITf II,:MJNt.kI\Lllt::~ ~t.6!'..'It.M"Tl',"~ TIME (nAYS) 

(,.UI")f'~S/l Tf: ,)~ ... Flt-R/,L 17l:.11 FLow f II"I';T l"'tl 

INPIJTS 

FHACT I'm FRller lUI. COLLEcT IOfI IJf~CAY 

Fl\.)w ftAl~ uF peA IJIC;CtIAP(;[I' TI"'£ lIME 

(.,Al./I'IIIYl IIJII Y5) ("US) 

<;IIIIM IiLI:.E-V \.IlIrt. 1."4f.+04 1.0000 .1(01) U.bOOO .'#300 

£UlJlPHtYT np.IlIlS ~.:\OF+O~ .1700 .100U a.bOOO .\1]0(1 

CLFAN wllsrF 1' .... UT Q."'or.+o.( .09JO .1/l00 '.7000 .1300 

IIHHY wJI·SIF" ~.' Of+G.1 .oillo 1.11000 J.tlOOO .1900 

IJLOwN)IoIN ?.lM+O~ O.nOO 0.000 0.000 UNH~FATt::1I IiL(hHJlIWj !i. 1·000 0.000 0.000 

kF.6ENlHMI r C;Ol'> :4. 4 OE+O.J .100 4.700 .370 

(jllii£OlJ~ w/l~H. INPUTS 

ltiLI/£: 1<' 1~l)r cOi~TIlmlll'" STHI"'PIN'j OF FULL LE.TOWN FLOW 

~LlJ'" 1'1"'[ Ttl~OU(,H I;I\S Sl~tPp~fI ("PH) 

tl'.llf'I)P TI"'t t('\~ XE'lO" (1l1ly51 

tllILlIlJP T1"-E ~()H II.HYPTO'l (PAt'i) 

~ h.L TJ'~E of oECAy Tldk.S F.'OP Tl"l GAS ~TRIPPlR (IlAyS) 

t"'lll'4.1RY c:"ol.ktH LtAK TU AtljllLlA~Y flLiolll (LU/I1AY) 

v~'S WASTf SY~TEt. PAHrICIJLAH: HELt."S!'.. fRAt;TlON 

t"ll HANDL6 ~LDG !f'llll"E ~~UA:;f. F/:(ACTION 

AUI\lllAqy ttLUr, 

"'A~T'C~LA1~ HELtAst:: F~AtTION 

1')11 1"'£ .,E.UASF.: ft&iAC TI ON 

PA~TJ~UlAT~ hELE.~SE f~ACTIUN 

ClJlqIQN.1E'lt 1I0LU.~l (HILLln:j t T31 

1-1l1:,f)IJF"JCy 0" PHH1A~y Ct)OL"NT Of\jIoSSl'~G (TI~£S/YR) 

"'tU"4RY Til S~COI,lIAl(Y LEAK HArE (LA/USy) 

fHlH£ I" Nor· II I(jnNFY FILTf.R . 

I-I<"CT (,),'l IOU.l.flt by'PA~'ilfll(; CtlNOF.~!)Ale: [)Et1II~EHALIZER 

P\IIR 

3400'.00000 

O.AO ~SO.OOOOO 

75'.00000 

7".50000 

4'.00000 

15'.00000 

112.50000 

75'.00000 

8.401i00 

.65000 

OECONTAMINA flo" 

, 

5.00E.03 

5.0nE+03 

~.OnE+02 

5.00E+02 

1.00E+03 

1.00E+00 

0;.0';E+02 

1 ~22917 60'.00000 

60.00000 

JO~OOOOO 

160.00000 

'.01000 

'.10000 

'.01000 

'.10000 

.01000 

2'.45000 

2~0001)0 

75.00000 

CS 

2.00E.0' 

2.00f+01 

1.00'.0 

1.00E+0' 

1.00£+0. 

1.00E+O' 

1.00E+0' 

(IJUIM:: PAlHI'IO.~ fAcTo,1 (I;A~/lIIWI'" IN STE.At4 C,EN[RATOR 

H't;WII::NCY OF' CI.,rMT qLIl'; Iilli'l VOL PUI("F.( TIME.5/YH) 

.35000 

'.Olono 

2.00000 

'.1001)0 

'.01000 

1000.00000 

(,;11) oil _,HI,t1 VUL pU'(IJf 11),'INr PtLF.:A~[ F 1'\ 1\(; TI oN 

PA,qICIILIITl t:tfl.E.I'sl fHACTI()N 

Vllllr LOw Vl;L "l.m\j[ 'II\H Ict·.,) -

(;Il I tIT 1Ih' VIIL PU~ljl lOll [ijJ: Ht..U.ASF. FHACT IoN 

PI\~fICUL~rl ~ELlASE FHACTIO~ 

~ltl\l~ I.F..A" lU TUt'lttIM !ill)(; (u·:s/.1 'l) 

~hllf:ll')" 1(j(,HjE ~ELFI'5~" FhO'1 tllUIo/DUt'N TANKllflH 

I'I:.H(I:.Nl Ilf !\J.ll t,t, ilf.IOllt-ll FHJt, All t.':I(CT(l1l HH.tASf 

'.10000 

.01000 

1100'.00000 

0.00000 

1 '. noooo 

'ACTORS 

OTHERS 

1.00E+05 

1.00E+05 

1.00E+04 

1.00E+04 

1.00E+03 

1.00E+OO 

1.00E+04 
W 

I 

N 

!>AMflLf P"I~ n~v 

f'lll",MiY ~t CONUAHy 

Cn0lANI COOLANr 

PH C~I\JC J /Gr·\) (t') CIW~ I /(;~.) 

GASEOUS RELEASE RATE - CURIES PEA yEAR 

AUILOING VENTILATION 

F ------------------~------------------------. rrl:.l t1l1t,ULG REACToR AUXILIARy lURBINE 

RLOWOOWN 

VENT OFFal! 

AIR EJECTOR 

EXHAUST 

TOTAl 

-----------------------------------------------------~----------.--------.--------.-... -----------.---------------------.. -------- I I. I -i 1 4.500l-02 

).:.,"3 1.40nl-O' 

I.J97E-U& 

1. f17E-Of, 

5. ~f~-04 

1 • 4)~: - 11.3 

3.7[-03 

7.9f-03 

i.4E-02 o. 

4.5E-02 1.!f-04 

O. 

o. 

o. 

o. 

•• f'f':''' 

1.ll';' .. 

------------.---------------.-----.-------------.---------------------------------------------------.-.---------------------~-~~--

TUTAL H-l HELEASt~ Vl~ ~ASEOUS PATHWAy D 1100 CI/YR 

C-l~ HELtA~EO vIA G~S~OUS PATHwAy. 7.1 Cl/yR 

AN_41 nELEASEU vI~ CU~TAINME~T·VENT. 34 CI/YR 
5M;t-L!:. ;,w'( ;:~V 

GASEOUS RELEASE RATE. CURIES 'iR YEAR 

PQT(~MIY C;Er:(Jt-JI:'''~'{ GAC\ C;TFU ...... tNG BUILDING VENTILATION 

COOL"IIT CtH)L"NT -------.-.---.. --.--- ---------------------------.~- .. LOWDOWN AIR EJECTOR TOTAL O~lcuUCl/c;r1' (MICRtJl:(/tit", !;tlt.1 rI,(I"1:~ CIII"T I NUoUS RE.ACTOR AUXILIARY TURBI~E VENT 0"1" EXHAUST 

----------.-.-----________ --.----_.-----__ .-____ . __ -__ ----------.--------. ________ -____ -_________ w_-________ ----------------~-----

KR-Jl!j~1 1 • (;)OOI:.-t' 1 "l.4CJOE-1I1! I) • o. it.dEoOI '.OEoOO O. O. 2.0E.00 5,:11.01 

K~_R~ 4.3UIIl-01 R.QU(1l-1/8 2.1HO? 8. J r+o2 9."Eo02 CJ.OEoOO O. O. 4.0£.00 2,.ltU 

1C~_R7 I.SOOL-OJ l.OOllt-(l8 \1. O. 1.~F..01 :\.OE+OO t. 0, 1.0E.00 • •• ,.oi 

KH_RIi ?.8Jut._Ol S.9110f-08 O. O. S.flE+01 6.0E+00 II. •• :I.OE.OO .,Sf,oi 

Xl-IJ1M 7.Jfl Ol-01 1. 5U IjE-07 1.1E+lll 2. Ur-Ol 1.!iF+03 i.5E+Ol •• O. T.OE.OO 1,6'.03 

XE-1JJt1 7.IjIlOt._O? I.SCJOE-OEI II. O. 1.UE+02 i.OE+OO O. I, O. 1.01.or 

Xl-IJJ ;>. hllOI:. + 0 () o:;.4110E-1I7 o. o. 4.7F.:+03 5.5E+Ol O. t. 2.6E.Ol 4.81.03 

Xl-lj~M 1. HIUI:.-O 1 ;>.7CJOE.-1/6 O. n. l.Of+OO l.OF+OO O. I. 1.0E.00 ',0' .... n-1J~ FI.S!I'JI:.-Ol 1.I:IIIoE-07 O. O. 4.61:+02 i .8F.:-01 O. ,. •• OE.O' 4,9ItOi 

w Xf:.-l.H l." 'IOt,-O;;> l.llloE-U9 o. O. o. n. I. I, O. I. I 

W XE_131:i 1.~O(ll-Ol ;>."\lof.-Ol'l lI. O. 2.0E+OO :1.01::+00 •• '. 1.0E+00 '.Oltot 

TOTAL 'W'lU" t;r'':>t.5 'tU.U 

------------.---------------... ---------.------------------------------------------~---.----------------------------------..... ~.-

0.0 APPI:.ANI~~ l~ THt T~HLE !WU'LATE~ ~lLfASE. I~ LEi~ THAN 1.0 ~I/YR FOM NOBLE GAS. 0.0801 Cl/YR fOR I 
w 

I 

~ 

;,At,PLf u w~ R~ II 

NlJCttUt:. 

",IISTE GAS 

st<,rr:'" 

AIHbOHNE PARTICllLATr RELEASE RAT[-CURltS PER YEAR 

BUILDING VENTILATION 

REACTOR AUXILIARy FUEL HAN'LG TOTAL 

------------------------------------------------------ ---------------------------------.--------~-------------~-------------------

CP'-51 1.4F-07 9.2t:.-05 3.2E.-06 1. 8E_06 9,1£-05 

Mo\I~"4 Z.lE-O~ 5.3E-05 7.8E-07 3.0E';'06 5.1£-05 

CO:..'; 7 o. f3.?E-Of, o. o. 8.2f-06 

C()-5~ 8.7r-ot! 2.~E-04 1.'iE-05 2.1E_04 .,8[-04 

CO"I)I) 1.4E-01 2.bE-05 5.1[-06 R.lE_OIS 1.1[-04 

FF.:'-C)Q 1. tif:':-08 2.7E_05 5.0E-07 o. 2,8l-05 

SI-/:"H9 4.4£-0 1 1.3E-04 7.5£.-06 2.1E-05 1.6[-114 

SR"Q/I 1.7F:-O' 5.2E-05 2.9£-06 A.OE_06 6.3[-05 

lI!'-Q<; 4.~f-Ot! (l. 1.OE-05 3. 6E_08 I.O[-OS 

NH!.QC; ].7E-08 1.8E-05 ~.oE-il7 2. 4E-05 4.:»[-05 

Ii!)!.! "3 3.21':-OB 1.6E.-05 2.JE.-07 J.8E-01 1.7E-OS 

HU"'106 ?1F:-oti o. f..OE-;;1! 6. 9E-07 1.8[-01 

SB"12~ o. CI. 3. 9 E-;;8 S.1E-01 •• \f-01 

CS-l"'l4 J.Jf-0 1 2.5E-OS S.4E-;;6 1. 1E-05 t.8E-U 

CS-)10 -;.3~-Od J.2t-OS 4.t!E-n7 o. S.JE-tlS 

C5"111 7.1[-0 7 5.5£'.05 7 .2f:-06 2.1E~05 ,.of-II! 

ElA:')40 2.3r-ol o. 4.0E-ii6 o. 4.2£ .. 06 

CF.-141 2 .2r -0.1 1.3F .• 05 2.~E~;;7 4.4E_09 l.l[-05 

---________ -_________ -_____ 4_---___ -----___ -____ ._--_-_______ -_-__________ --. _____ -_. _______________ . _____ -______________ • _______ _ 
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 liquid effluent 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 

2 

2 

2 

2 

2 

2 

2 

2 

Column 

1-72 

75 

1-10 

11-20 

21-30 

31-40 

41-46 

48 

50 

52 

Input Data 

Title 

The va 1 ue 2 

The value 0.632 

The value 0.333 

The value 2.0 

The value 1. OE- 25 

The date (month, day, year) 

of the calculation 

The value 

The value 0 

The value 0 

A description of the information contained in the nuclear data library 

can be found in the report ORNL-4628, "ORIGEN - 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 
-UECK 

C 

C 

FIGURE 3-3 

PGALEGs PROGRAM LISTING FOR GASEOUS DETERMINATION 

GALE CODE FOR CALClJL~TIM;' GIISEOUS E.FFLlJEI~TS ~ROM PwRS. M()OIFIE) 

AUG. 1919 TU IMPLEMENf ~PPENDIX I TO 10 C~R ~~. REACToR 

Oo02~O 

C 

C 

C 

000300 

000310 

1974000320 

WATER CONC~NTRATIONS CALCULATE.D USING ~E.Ttl0DS OF DRAFT STANDARD 

ANS 237 "HAUIOACTIVE i1ATt.~IAL~ IN PRINCIPAL FL.LJIO STREAMS OF 

LIGHT \VATER COOLED NU~LEAM POWER PLANTS" DRA~T LJATED MAy 20, 

C 

C 

C 

C 

C 

THE FOLLOWING FIRST STATEMENT IS SPECIFIC FOR THE CDC USEHS. 

FOR THE I~M USER5, OELETE THIS STATEMENT. 

PRuGRAM PGALEGSCINPUT,OUT~UT,TAPE5=INPUT,TAPEb=OUTPUT) 

K[AL NUCLlU(13) 

REAL PPARTCIB) 

DIMENSION ACONT C 13) .C~CP (13) ,CBSP (13) 'ASHIMC C 13) ,ASHIMS C 13) 

OIMENSrON C~pp(13) 

DIMENSION ASHIM(13) .C~NCPCl3) ,CONC5(l3) ,Pt.:CONCI3) ,FHtjLCI3) 

DIMENSION DECOHCl3) .NAMECC!) ,E""T(13)'Tt:lL(l3) ,CPLCI3) ,AXBlCl3) 

Ulf>1ENSION tH/OGCI3) ,TO! (13) ,X2(13) ,X3CI3) ,)\4Cl;3) ,X5(13) 

lJ I MEN S ION CT PRO C 1 3) • X ~ 1 C 1 3) , X P 2 C 1 1 ) ,w U R D C 14) Hi A R 0 ( 5) , W U k D ( 4 ) 

DIMENSION PCBlCIB),PAXBLC 1S),PCt:lPClS) ,PAX~PCl~) ,PGW5CI8) ,~TOTPC1R) 

DIMENSION PGWLCIB),PC~CPC18) ,PCBSPCI81 .PR~ONTCI8) 

ulMENSION PFHBLCI8) ,ptHBP(IB) 

UIMENSrON kNAXCZ) ,RNAXSCZ) ,RNFH(2) ,RNFHS(~) ,RNT(2) ,RNTS(2) .RNS(2) 

DATA NIICLIU/" KR-B5M",1I KR_85"," KR-87"." KR-~8"," XE-131 MII , 

1" X t -1 , 3M" ," X E -1 33" ," X E .. 13 5 t-lI' ," X E -1 35" ,,, X!:: -131" , 

2" XE-13B"'" 1-131"." .I_133"1 

DATA PPART/" CR-51"." "1N·54"." CO-57"," e0 _58"." CO-bO"." FE-59", 

1" 5R-89"'" ::'R-90"," Zi"1_95"," 1;8-95"," Ru-l03"'" RU-I06"," 58-125", 

:::>" C5-134"," C5_136"."·CS-137 11 ,II 8A_I40"." CE-141"1 

C XP1 A~D X~2 ARE THE PHIMA~Y COOLANT AND StCO~~AHY COOLANT 

C CONCE~TRATluNS. RESPt~TIV£LY (MICROCI/GH). 

C 

C 

UATA XP1/1.bE_1,4.3E-l,1.~E-l,2.8E_1.7.3E-l,1.0E-2,Z.6E.0'1.3E-l.8 

1.~E-l,3.4E-2.1.2E-1,4,5[-2.1.~E-11 

OATA ~P2/3.~E_B,8.qE-~,3.0E-8,5.9E_B.1.5E-7,1.5E-8.5.4E-7'2.7E_8.1 

1.tiE-1.7.1E-~.2.5E-81 

C PECAY CONSTANTS FOR T~E CURRESPUNDING NUClID (1/SEC). 

C 

C 

PATA DECON/4.3BE_5.2.03E-~,1.~2E-4.6.88E-~.6.~OE-1,3.95E_b.1.S2E-6 

1,7.41E_4,2.09E_5.2.96~_3,8.14~-4.9.97E-1,9.11~-61 

C NORt-1ALIZEIJ IODINE ANI~'!AL kELEASE CcI/YR/MICRO~I/GM). 

C 

C 

DATA RN5/0.32,0.321 

(JATA RNAX/O.68,O.681 

UATA RNAX~/2.5,2.51 

DATA RNFH/O.038,O.0381 

UATA 8NFHS/O.093,0.09~1 

DATA R~T/3.~E3.3.8E31 

UATA RNTS/4.2E2,4.2E21 

C PAHTICIJLATE AN~UAL RE~EASb RATE CCI/YH) 

C 

PATA PCRP/9.2E~3,5.3E~3,c.2E_4,2.5E-2,2.6t_3.'.7E-3,1.3E_2.5.2E-" 

lO.OE+O.l.HE-3.1.~E_3.0.0E+O.0.OE+O,2.5E-3,3.2~-3,5.5E-3,0.OE+0,1.3 

2E.-31 

DATA PAXdP/3.2E_4,7.8~_5.0.0E+O,1.9t_3,5.1E-4,5.0E-5,1.5E-4.2.9E-4 

1.1.OE-1,3.UE_5.2.3E_5'6.0~-6,3.qE-6,5.4E-4.4.~E-5,7.2E-4,4.0£-4,2. 

2bt.-5/ 

UATA PFHdP/l.8E_4.3.0~_4,O.OE+0,2.1E-2.B.c.E-3,0.oE+o,2.1E-1.8.oE-4 

3-16 
c 

1,3.6E_h,2.~~_3,3.8F..5,~.9t_~,~.7E-~,1.7E-j,0.UE.0,2.7£_3,0.OE.0,4. 

24E.-7/ . OAT A PGWS/l.4E_5,2.1E~6,0.OE.O,8.7E-6,1.4t_5,1.8E-6,4.4E-~,1.7E-~, 

14.8E_6,3.7E_6,3.2E_6,~.7t-6,0.OE..0,3.3E-5,5.3~-6,1.7£-5,2.3E-5,2.2 

2E-b/ 

C 6UIlT-IN PARAMETERS 

C 

C 

c 

c 

c 

OPFRA=O.80 

AUXLR=i60. 

EM=2.0 

GENL=75. 

CLFNG=ii.03 

CLFI=e.(lE-6 

PURTI~=]6. 

TBLII.=1700. 

AFPTEG=o.O 

READ(5,lOOO)NAME,TyPE 

WRlTE(f,,1440) 

WRITE(6,lOOO)NAME,TYP~ 

HEADC5,1010)WORD,POWT~ 

wRITE(6,lOlO)WORD,PO~!H 

wl-\lTEC6,1020) 

READC5.1010)WORD,PRIV~L 

wRITE(f"lOIO)WORD,PRIVOL 

~EADC5.1010)~ORD.DEMl~L 

~RITE(6,IOlO,wORn,DE~lFL 

READ(5.101u)WORD,C~FLR 

wRITECA,lJIO)WORD,CBF~R 

RtAD(5,lOlO)WORD,GEN 

wRITEC6,lOlO)WORD,GEN 

READC5.1010)WURD,TOST~L 

~RITE(6,lOlO)WORD,TosiFL 

~EAD(5,lOlO)WORD,WLI 

WRITE(A,lOlO)WORD,wLI 

wLI=GEN .. IIIL! 

READ(5,1040)TBD,KFNRT 

WRITE(6.10

5 0)TAD IF(KF~~T.EU.O)FNRTSC=U.99 

IFCKFNRT.EQ.l) FNRTSC~O.9 

IF CKFNRT.EQ.2) FNRTS~=}.P 

READC5.1010)WORD,REGENT 

WRITE(6,lOlO)WORD,REG~NT 

HEAD PATA FUR LIQUID ANFO~MATION 

HEAD(5,1010)WORD,FFCDM 

WHITE(f"lOlO)WQRD,FFCDM 

HEAD(5,1060)WARD,SBLU~ 

CWA=l.O 

REAO(5,1070)DFICW,DFC~cw,~FCW 

HEADC5.10

8 0)TC,TSTORC.CWFP 

wRITE(6,1090) 

WkITE(",llOO) 

wRITE(6,1110)WARD,S8L~R,CWA,CWFD,Tc,TSTORC,Df!CW,OFCSCW,DFCW 

RlAD(5.1120)W

ARD,EDFLH,EDA kEAD(S.1010)DFIED,DFC~ED,UfED 

READ(5.10~O)TE,T~,EDf~ 

W~lTE(~,lllO'WARD,EDFLR,EllA,EUFD,TE,TS,DFIED,~FCSED,OFED 

kEAD(5,1120)WARD,DWFL~,Dw~ 

H[AD(5,1070)DFID~.DfC~Dw,UfDW 

3-17 

004140 

00 48 20 
C 

~tAD(5.10~Q)TD,TSTQRO,DwF~ 

WHITE(6.l1l0)WARO.DWf~R.DwA.DWFO,TD.TSTOR~,DF~DW,DFCSDw,OFow 

~EAO(5,1120)WARD.DWFL~.D~i 

~£AD(5.t070)DFID?,OFC~D2,VFD2 

READ(5.)0~0)T2,TSTOR2IDwF2 

wRITE(6,1110)WARD,DWF~2.D.2,DWF2,T2,TSTOR~,DF'D2,DFCSD 2,DFD2 

R~ADC5,1130)BDTFR 

~EADC~.1070)DFICM,DFC~CM,DFCM 

READC5.10BO)TCM,TSTORe,~MtD 

REAOC5.1130)RGWFR 

READ(~.1070)DFIRG,DFC?RG,eFRG 

READ(S.10dO)TRG,TSTORR,RGfD 

IFCTBD~EQ.O.O) GO TO ao 

BOFR=TRD

o l.OE3oBDTFR/O.34f.6 WHITEC6,1140)BDFR,CMF~.TC~,TSTORA,DFICM,DFCSC~.DFCM 

BUFR=TBDol.OE3oABSCl.~8DTtR)/O.3476 

WRITEC6,1150)BDFR 

IFCFFCDM.EQ.o.O) GO TU 50 

30 IFCRE~ENT.EY.o.O) GO !O 40 wRITEC6.ll60)RGWFR.RGtD'T~G,TSTORR.DFIQG,~FCS~G,DFRG 

GO TO 5('1 

40 RGwFR=n.O 

WRITEC6.1160)RGWFR.RGtD,THG.TSTORR.DFIRG.~FCS~G'DFRG 

C READ DATA FOR GAS INF~RMATION 

C 

50 wHITEC6,1170) 

HlAO(5,11~O)KGTRWT 

H (KGTRWT.t::U.O) GO TO 70 GTRW=CO[~lFL-SBLDR/1440.)/D~MIFL 

IFCKGTRWT.EU.2) GO TO 60 

wklTEC6ol1

9 0) 

GO TO AI' 

60 GTRW=O.c~~GTRW 

wHITE C6, 1200) 

GU TO RO 

70 uTRIO=O'.O 

w~ITE(",1210) 

80 SR~=GTRW~ut::MIFL.(S~LO~+EDtLR)/1440. 

wRITEC6,122~)5AB 

READC5.l010)WORD,TAUl 

WRITE(6,lOlO)WQRD,TAUl 

READ(S,lOlO)wORD,TAU2 

WHITE(6,lOlO)wORD,TAU~ 

RtADCS,)OlO)wORD,TAU3 

WRITE(6,lOlO)WORD,TAUJ 

wRITE(6,1230) 

(';WPRF=1.0 

AXIRF=i.o 

AXPRF=i.o 

CHIRF=i.o 

CHPRF:q.O 

CLIRF;t.O 

CLPRF=l.O 

F.HIRF=1.0 

FHI-'RF=l.o 

CAIRF=i.o 

CAPRF=j.O 

~lAD(5.1250)~URD.G~HR~ 

IFCGw~RE.GT.O.O)GWPRf~l.O-GWHHE/IOo. 

wRITE(6,1260)WURD,GWP~F 

kEAD(5.1270)WARD.F~CH~E,FOHRE 

IF(FHCHPE.GT.0.O)FHIRr=1.O-fHCHkE/100. - 3-18 

005070 

0050QO· 

U05?lO 
c 

c 

c 

IF(FHHR[.~T.O.O)FHPRF~l.O~FHHRE/lOO. 

WRITE(6.12dOIWARD.FHI~F,FHPRF 

~EAO(5.1270)WARD.AXCH~E,AXHRE 

IF(AXCHRE.GT.O.OIAXIRt=l.O-AxCHRE/IOO. 

IF(AX~RE.uT.O.O)AXPRF~I.0-AXH~E/I00. 

WRITE(6.1280)WARO.AXI~F,AA~RF 

~EAD(5.1010)WORD,CONV~L 

WRITE(6.10loIWORn,cONXOL 

WRITE(6,12901 

~EAD(5.1370)WARD.CACH~E,CAHRE~CFM 

IF(CACHRE.GT.0.0)CAIRt=I.0-CACHRE/IOO. 

IF(CAHRE.GT.O.0)CAPRF91.0-CAH~E/100. 

IF(CFM~EQ.O.O) GO TO Vo 

KID=1 

wRITE(6.1300)CFM,PURT~M 

60 TO 100 

90 KIO=O 

wRITE(6.1310) 

100 IF(FFCDM.uT.O.O) GO TU lIP 

wRITE (6, 1320) 

GO TO j20 

110 FlaCD=l.O-F~CDM 

WRITE(6.1 33 0)FIBCD 

12U IF(T8D.E'~.().0) GO TO 130 

CUr-.l=o.OI 

wRITE(6,13

4 0)CON 

,GO TO}40 

130 COI'I=1.0 

WkITE(6,13

4 0)CON 

140 READ(S,J350)wARD,CHCHKE,Ct,lHk[,[rJP 

IF (CHCHRE.GT .0.0)CHIRt=1.0~CHCHRE/IOO. 

IF (CHHQE.uT .0.0)CHPRF~ .0~Ct1Ht-<EIl00. 

tN=2.0.ENP 

wl .. <rTE (o.lJbO) EN 

tlHITE (t'Jt 1200) WAR[).CHI~F,CHPRF 

READ(5.1310)~ARD.CLCHKE,CLHRE,PNOVI 

H (CLCHRE..GT .0.0)CL1H~=1.O-CLCHf~E/100. 

If (CLHRf.GT.0.0)CLPRF~1.0-CLH~E/I00. 

If(PNOVl.LT.l.O) GO TU 150 WRITE(6.1380)WARO,PNOVl,~~RD.CLIRF.CLPRF 

GO TO )60 

150 wHITE(6,13

9 0) 

160 WRITE(6,1400ITRLK 

READ(5,lOlO)WURD,FVN 

WRITE(6,lOlO)WORD,FVN 

READ(5,}010)WORD,fEJP 

/fEJP=l.O-FEJP/IOO. 

~WRITE(6,1010)WORn,fEJe 

~EAD(5,1010)~ORD,PFLAUN 

IF(PFLAUN.LE.O.O) WRI!E(~~1430) 

CUNVE~SION UF UNITS 

TOSTFL=TO~TFLolOOOOOO. 

wLl=vlLlolOOO. 

CVNVOL=CUNVOL*lOOOOOO, 

CFH=CF'-1*lOOO. 

Tt;D=rtlD*l£3 

PRIVOL=Prtl

VOL*lE3 JtMIFL=D~MlfL*500.53 

SbLDR~SRLUR*.3476 

EUFLR=[DFLR*.3476 

3-19 

005710 

006210 

006220 

006230 

00

6240 

On6260 

006270 

0062AO 

Op6290 

006300 

006310 

On63?0 

0;'6310 
c 

c 

c 

190 

200 

210 

220 

230 

240 

C 

C 

C 

OwFLR;DWFLHO.347~ 

UWFL2=OwFL2*.3476 

C8FLR=CBFL~*500.53 

006340 

0063'50 

006360 

H3COPW IS THE PWR TRI!IUM PRIMA~Y COOLANT CON~ENTRATION IN UCI/G~ 0063AO 

H3COPW=1.0 006390 

H3PRP.=0.4*POWTH 

TPLRPW=S8LDR*CWA*CWPD~EDF~R*EDAOEDFO.UWFL~*OW~oOWFO.D~FL2*OW2*OWF2006400 

H3RLP~=TPLRPW*H3COPW*j.~1T 006410 IF(H3RLPW.GT.0.90H3~RtW)H~RLPW=0.90H3PRPW 

H3RLG=H3PRPW-H3RLPW 

~IV=10.o0(INT(ALOGI0(~3RL~»-1) 

IDIV=PIV 

IH3RLG=INT(H3RLG/DIV.Q.~)~IDIV 

IF(TAU3.E~.0.)TAU3=.01 

SRS=SRBo500.S3 

PE=365./T AU 3 

Tl=3.1s5 7E7!ENoOPFRA T3=3.1~57E+07!pE 

T4=TAUlo8b4QO. 

T5=TA Ul'0Rb 400. 

00 190 1=1,13 

UECOH(I)=U[CONCI)0360V. 

00 200 1=1.13 

CONCPCI):::.t.PICI) 

If (PO"TH.LT.3000 •• 0R.~owTH.bT.3800.) GO TU 21U 

P (PRIVOL.LT .5.0E5.0R,PRIYOL.GT.6.0E5) GO TO ciO IF(UE~lFL.LT.3.2F.4.0R.DEMIFL.~T.4.2E4) GO TO cI0 

IF (SI3LDP.LT.250 •• 0R.S~Lt)R.GT.I0uO.) GO TO 210 

H (C8FLP.GT.7500.) GO TO cl0 

If(KGTpwT.GI.O) "0 TO 210 

GO TO l'40 

AI- pTEG:::l .0 

~NG2=(SRLDR+DEMIFLo GTRw)!PRIVOL 

RHAL2:::CDEMIFL*0.99+0.Yl~S~LDR)/PRIVOL 

HK2G=161. 7bOPOWTH/PRIVOL 

DO 230 1:::1,13 

IF(I.GT.l11 GO TO 220 CONCP(I)=CUNCP(I)ORK2~*(.Q009.DECOH(I»!(~NG?·DECOH(I)) 

GO TO 230 CONCP(I)=CONCP(I)*RK2~*(0.067.DECOH(I»!(HHALG+DECOHCI» 

CONTINUE 

IF(TBP~E~.O.O) GO TO ~80 

PWTVPE=I.0 IS FOR pWR~ ~ITH U-TUHE STEAM ~ENE~ATORS 

PwTVpE=l.O 

00 250 1=1,11 

006430 

006440 

006450 

006460 

006470· 

0064M 

0064QO 

006500 

006510 

006520 

006!=i30 

OOf>f>60 

250 CONCS(I)=XP2(I) 

c 

CONCS(12)=1.8E-6 

CONCS(13)=4.8E-6 

IF(AFPTEG.EQ.l.O) GO TO 300 

IFIWLI~LT.4.0E5.0R.WLA.GT.5.0~5) GO TU 300 

IfCTOSTFL.LT.l.3E7.0R.TOSTFL.uT.l.7(7) GO TO joO 

If(T8P~LT.5.0E4.nR.TBq.GT.l.OE5) ~O TO 300 

I r' IfF COM. G T • 0 • 0 1) GOT 0 3 0 0 

('0 TO 340 

C pwTYPE=?.O 15 FOR pWR~ WITH ONCE-THROUGH ~TEA~ GE~ERATO~S 

C 

280 PI'TypE=2.0 

UO 290 I=ldl 

3-20 
290 

300 

310 

320 

330 

340 

C 

C 

C 

C 

C 

CONCS(ll=X.P£!(Il 

CUNeS(12)=5.2E-8 

CONCS(13)=1.6E-7 

IF(AFPrEG.EU.l.O) GO TO 300 

IF(TOSTFL.LT.1.3E7.0R;TOSTFL.UT.1.7E7l GO TO JOO 

IF(FFCDM.lT.0.55.0R.FrcD~.GT.O.75l GO TO JOO 

GO TO 340 - ' 

CONTI~UE 

If (fFCDM.GT.O.Ol.AND,FfC~H.LT.l.Ol FFCDM=0.2 

RHAl3=(TBD.FNRTSC •• 9.~ON*rOSTfL.FFCOH)/~Ll 

00 330 I=ltl3 

IF(I.GT.l1l GO TO 310 CONCS(I)=CONCS(I).1.5~7/TUSTFL.(CONCP(I)/XP1(~» 

GO TO 330 

IF(PWTypE.EQ.2.0) GO Io 310 CONCS(I)=CONCS(Il.(4.~ES/.LI).(O.17.DECOH(I»/(RH~L3.DECOH(I»~ 

1 (CONCP( I ) I XP 1 ( I ) 1 

,,0 TO l30 

CONCS(I)=CONCS(I)~(1.UE5/WLI).(27.0.DECOH(I»/(HHAl3.DECOH(Il)~ 

1 (CONCP cI 1 I XP 1 ( I ) ) 

CONTIl'iUE 

PNOV=PNOVI/CONVOL~60. 

THIS PART OF PROGRAM 1S FUR NuBLE GASES 

J=O 

DO 370 1=1 tl3 

X2(II=(DECON(I).PNOV/~600.)~Tl 

IF(X2(JI.bT.30,) X2(1 1 =30, 

)(3(1) = otCON(!) ~ T3 

If(X3(J).C,T.30.1 X3(0.=30. 

X4(1) = tJECUN(I) ~ T4 

IF(X4(II.GT.30.) X41Il=30. 

X5 (I) = Ut:CON (I) ~ T5 

.(UK=X5(I) 

IF(X5(J).GT.30.) XDK=JO. 

IF(I.Gr.ll) GO T~ 350 

007140 

007?40 

007?50 

On7?AO 

007100 

0013AO 

On7370 

IF(I.Gr.4) XDK=X4(Jt CTPRO(I)=(CUNCP(I)~PH~VOL~CLFNG)/(DECOH(II.P~UV)~1.892E-5 O~7470 

ACONr(II=CTPRO(II~(1.~EXP(-X2(Il)l On74RO ASHIM(I)=(eUNCP(I)*SRe)/D~eOH(I)~4.54E-4.{1.~~XP(-X3(II»~ 0074QO 

AXBL(I)=CUNCP(Il~AUKl~~.165700PFHA 007500 

CbCP(I)= EN ~ PNOV ~ (CTPRO(I)~TI/3600 •• ~TP~~(II~(EXP(-X2(I»-1.l0n7S10 

11 WEeOH (Il.PNOV) ) 0075£10 

C8SP(I)=EI~*~eONT(Il 007530 

CBL(I)=CBCP(I).CBSP1Il 007540 

ASHIMC(I)=PE~ASHIM(I)!EXP(-XDK)~OPFRA 

ASHIHS(Il=EM~CONCP(I)!PRIYOL~4.54E~4~EXP{-XDK) 

EJT(I)=CONCS(I)~TOSTF~~3.~77~OPFRA 

T~l(I)=CONCS(Il~TBLK~~.977~OpFRA 

FI1Bl(I)=O.O 

BVOG(I)=O.O 

TEST=l.O 

IF(eaL(I).lT.TESTlCBl'Il=O.O 

IF(ASHIMS(ll.LT.TESTI~SMIMS(Il=O.O 

IF(AS~IMCtI).LT.TEST)~SHIMC(I)=O.O 

IF(EJT(I).lT.TEST)EJT'II=O.O 

IF(TBL(II.LT.TEST)TBl'Il=O.O 

IF(AXBl(I).LT.TESTlAX6l(I)=0.0 

\,,0 TO 370 

THIS PART UF PROGRAM IS FVR IUDINE 

3-21 

007570 

0075RO 

007SQO 

007600 

On7f.lO 

Oii71i?0 

0076~0 

007MO 

007650 

007660 

007f.80 

0076QO 
C 007700 

350 CTPROCII=ICUNCPCI)~PR1VOLOCLFl)/(DECOH(I)+PNovlol.A92E-5 

ACONTCI)=CTPROIII*CI.~EXP(-X2(I)I) 007720 

J=J+l 

AX6lCl)=CRNAX(J)+RNAX~IJ»oCONCPIIloA~IRF 

fH~lCl)=CRNFHCJ)+RNFH~(J»)*CO~CP(II*FHIRF 

ASHIM~(I)=O.O OQ7740 ~SHIMSCI)=O.o 007750 G~CP(l)= E~. PNOV • (CT~RO(I).Tl/3600 •• ~TP~~(I)*(EXP(-X2CI)I-I.)007760 

1/(DECOH(I)+PNOV».CLI~f 

C~PP(l)=ENPoACONT(It*~HIRM 

CBSP(I)=RNS(J).CONCP(~)*C~IRF 

CBlCI)=CBCP(I).CBSP(Il+CBRPCI) 

EJTCI)=1.7E3*CONCS(I)!CON~FEJP 

TBlCI)=CRNTCJ)+RNTStJ.)*CUNCS(I)oCON 

~VOG(I)=CONCSCI)*TBD*tVN.J.91100PFRA 001830' 

IFIKrP.Ea.O) GO TO 36Q 

DLAK=ICFM*60.*CACHRE*O.01*O.7/CONVOL)+DE£UHIII 

~XX2=DLAK~"'l.JRTIM 007A60 

IFCEXX2.GT.30.) EXX2=lO. 

EXPF=EXPI-EAX2) 007ARO 

EAPC=I~-E~PF 007A90 ~LSS=CHIHFoCONCPII)oP~IVOLoCLFI*I.892E-5/ULAK*E~PC 

CdL(I)=CHPPII)oEXPF+ELsS~~NP+CBCP(Ilo(I.-~URT1~/(8760.oOpFRA/fNI)+ 

12*CHr~F*0.lb~CONCP(II~PURIIM/124.*32.5IoEAPC/(DLAK*PURTIM) 

?+(C8SP(I)-2.oCHIRF*0.l6*lUNCPII)*PURTIM/(24.*~2.511 

360 TEST=o.nOlll 

IfCCRL(lI.LT.TEST)CBLO)=O.O ()07930 

lFIEJT(I).LT.TEST)EJT'I);O.O 007940 

IFCBVOG(I) .LT.TEST)BVOGCI,=O.O 007950 

IfCTel(I) .LT.TESTITBL~I)=O.O U07 Q AO 

IFCAXBLCII .LT.TESTIAX~LCII=O.O 007 970 

IF C F HB L C I I • LT. TE S T I F HdL C I ) = 0 • U 

370 CONT!!',"E 

M~lG=l 00H020 

I~SIG=13 

CALL SIGF2ICBL,MSIG,NSG) 008040 

CALL SIGF2CASHIMS,MSl~,N~lGI OnAOSO 

CALL SYGF2(ASHIMC,MSI~,~SlG) OoR060 

CALL SYGF2CEJT,MSIG,NSIG) OoH070 

CALL SIGF2CbVOG,MSIG,~slG) DoRDAO 

CALL SIGF2CTAL,MSIG,N?IG) OnR090 

CALL SYGF2IAXBL,MSIG,NSIG) OoRIoO 

CALL SIGF2CFHBL,MSIG,~sIGl 

IJO 380 1=1,13 TOT(I)=C~L(I)+EJT(It+1BL(1)+AXBlCI)+FHBL(1)+eYOGCI)+ASHIHCCI)+ASHI 

IMSCl) - . 

380 CONTINUE 

CALL SIGF2CTOT,MSIG,N~I6) 008140 

wRITE 16,1440) 

wklTEC6,1~SO)NAME 

wRITE(6,14 60) 

I'IkITE(6,1560) 

wRITE C6, 1480) 

LJO 385 1=12,13 wkITECA.1495)NUCLID111,CO~CP(1) ,CONCSCI) .tHRlII), 

lCjjLCI) .I1X HL(I) ,TSLcI) 18VO,?(I) ,[JTCI), TOT(l) 

385 CONTINUE 

WHlTE(6,1~~O) 

wRIT[(6,lSlO) IH3RLG 

wRlTE(6,1440) 

WklT[C6,14S0INAME 

wRITE (,r" 146 0) 

3-22 
I¥ kIT E ( 6 , 1 4.7 0 ) 

wHITE (6,1,+80) 

C;ASTOT=O,O 

uO 390 I=ltll 

wkITE(6,1490)NuCLIDtH,CONCP(I) ,CONCS(I) ,ASHII'lS(I) ,ASHP1CIII, 

1 CdL (I) ,AXI:iL (I) ,TBL (11 , BVO~ (I) ~ EJT (I) , TOT (!) 

GASTOT=GASTUT+TOTllt 

390 CONTIt'4uE 

008360 

008370 

c 

c 

c 

c 

c 

410 

,+20 

430 

440 

450 

C 

1000 

1005 

OlY=lO~oO(INT(ALOGIO(~ASTaT»~l) 

GASTOT=AINT(GASTOT/OIX+O.$)oOIV 

WkITE(6,1500) GASTOT 

WHITE (6, 1 4 80) 

vlRITE(6,1520) 

wkITE(6,1440) 

wRITE(6,14 5 0)NAME 

wRITE (6, 1530) 

wRITE (6,15 4 0) 

wRITE(f,,14 8 0) 

~H=876n,~OPfRA/EN 

THIS PART OF PROGRAM ~S FUR THE PARTICULATES 

l)U 430 I=I,t8 PHCO~T(I)=PL~P(I)/(87~0.~UPfRA) 

If(PNOv.GT.O.O) GO TO 4llJ PCI~CP II ) =0. V 

t""C fj SP ( r I :: U I <;Pf{cor,T ( I I ~(lh ~LHPRf 

"U TO 4cU 

PLoCP (I I = (t::r~~ (QHOPRCOIH (1) -PRCOhlT (I) IPNOVo (1,,,:,,E.XP (-PNOVoQH» I) 

l*CLPRF ." - . 

PCBSP(I)=«(N*(PRCONT(1)/PNOV*(1.0-EXP(-PNVVo~rJ»1*CHPRF 

PCHL(I)=PC~CP(I).PCB~~(I) --

PAXdL(I)=PA~8P(IIOAXP~F 

PfH8L(I)=PfHHP(I)OFHP~F 

P~~L(l)=~G~S(I)*GWPRF 

If(KIP.EQ.O) GO TO 439 

PULAK=CFMobO.OCAHRE~O,01oU.7/CONVOL 

P~XX2=pDLAKoPURTIM 

IF(PEXX2.bT.30.) PExx2~30, 

PE.XPF=EX P (-PEXX2) 

PEX.PC:;: 1. -PE.)'PF 

OOR560 

008710 

On8730 

008740 

~ELSS=PRCUNT(I)/PDLA~!pEXPCoChPHF 

PCBL(I)=PCB~P(I)OPEXP~+PELSSo~N.PC8CP(I)~(1.-~URTIH/(8760.o0PFRA/E008760 

IN» 008770 

COI'lTlr'IIUE 

M~IG=2 

NSIG=la 

CALL SIGF2(PCBL,~SIG,NsIG) 

DO 440 1=1tla PTOTP(I)=PCdL(I)+PAXB~(I)~PGWL(I)+PFH~L(I) 

CONT 1 i''4UE 

CALL SIGF2(PTOTP'MSIG1NSI~) 

DO '+50 1=1,18 

wRITE(6,1550)PPAQT(I) ,pGWL(I) ,PCBL(I) ,PAXt1l(Il,PFH~L(I) ,PTOTP(!) 

CONT 1 r~lIE 

wHITE (,;,14 8 0) 

STOP 

FOfo(I'-lATS FORMAT~ FORMATS fORMATS 

fORMAT~32X,8A4,1?X,A4l 

fURMAT(16X,IIRLOWDOWN IS P~OCE~SED THRUUGh CO~LJENSATE Dt:.HINII) 

3-23 

008B50 
1001 

1010 

1020 

1030 

1040 

10S0 

1060 

1070 

1080 

1090 

1100 

1110 

1120 

1130 

1140 

FORMAT Cl6X,JIRLOWnOwN IS ",UT PROCESSED THR UlJGt1 CONDo IJOHN.II) 

FURMAT(I6X,13A4,A2,FI0. 5 ) 

FORMAT (}6)l.,"PLANT CAP~CITl' FACTOH",T7 4 ,"O.RO") 

FORMAT/16X,IIPERCENT FUEL .ITH CLADDING l)EFECT~II,T74,F7.5) 

FORMAT(3 6 )l.,f8.4,35X,I£) 

FORMAT (l6X,IIBLOWDOI<'N ~ATE (THOUSAND L£;S/Ht:<) II,t:4X,F9.5) 

FORMAT(lSX ,4A4,A2,8X,t8.0) 

FORMAT(20X.F8.0,2(5X,~8.0» 

FORMAT(27X.F6.2,14X.F~.2,18X,F6.2) 

FORMAT(/." LIQUID WA~TE ilNPUTS") 

FORMAT (30X,"FRA€T.~ON FRACTION COLLECTlDN. 

1 F.LOW RATE OF peA DiSCHARGED TIME 

?INATION FACTORS"/20X;,,(G~L/DAY)"'23X,"(D~YS) 

DECAY"/8X;IISTREAM 

TIME",SX," DECONTAM 

(DAYS)",7X, 

311III,8X,"CS",8X,"OTHER~") 

FORMAT(2X,4A4,A2'lPE9'2,l~,4(OPF8.4'2X),3(lPE~.2,lX» 

FORMAT~ISX,4A4,A2,8X,E8.0~7X,F6.4) 

FORMAT~70X,FIO.5) 

FORMAT(2 X,"tiLOWDOWNII,10X,lPE9.2,11X,F8.3,'X,2(F8.3,2X), 

]3(lPE9.2dX) ) 

ll50 FORMAT (2X,"UNTREATED I;ILOwI'lOWN",lPE 9.2,llX.1I 

10.000 1.00E+00 1.OOE+OO 1.00E+00 1I ) 

1.000 0.000 

111'.0 FORMAT C2X,"r<EGENERANT SOLe; ",lPE9.2t14X'OPF5.3,2X,2(F8.J,2X), 

13(lPE9.2.l X) ) 

1170 FURMATCI,II GASEnUS w~STE ItIPlJTS"1 

1180 

1190 

1200 

ldlO 

FORMAT (791\ tll ) 

FU~MAT(}6X,IITHERE IS ~J~TlNU(lUS STRIPPING OF tULL LETDOwN FLOWII) 

FURMAT(lh>""TH[HE IS ~O"'TINUnuS LOw VOL PURGE OF VOL. CUNTROL TK") 

FORMATC16A,"THERE IS r"oT ~UNTlNlJvUS ST~IPr'IN(, of FULL LETUwN FL(l"'" 

II 

1220 FORMAT(16X,"FLOW RATE THRUUGH GAS STRlpPER(6~M)"tl9}(,F9.~) 

1230 fORMATC1e>}(,IIPRI~14RY CUOLAIIjT LEAK TO AUXILIARY HLDG CLI'I/UAYI".T72, 

1250 

12bO 

1270 

1280 

1"160.00nOO") " ' 

F0RMATC16}(,4A4,6X,F3.V) 

fUrlMAT (16A.,'tA4,4X,IIPAKTlCULATt:: RELEASE FR~cTI~rJ",6X,FI0.S) 

F U f( MAT ( 1 6 X , 5 A 4 , lOX, F 3" , 6 X , F 3 • 0 ) 

FUHMATCl6X'!::lA4,"IOLJII'I~ RELEASE FRACTIONII,11X,':lO.5/36X,"PARTICuLAT 

IE RELEASE. fRACTlON",6",FIO.5) 

1290 FORMAT'C16X,IIFR(QLJENCY OF PRIMARY COOLANT IJEGA?SING (TIMtS/yR)".T74 

1 ,"2.0000011/16X,"PHP1A~Y To SECO'WARY LEA~ RAT~ (Lt3/DAY) ", T72, 

2" 75.00(\00 " ) 1300 FORMAT(l6X,IITHERE IS ~ KIIJNEY FILTER"/20X."CO~TAINMENT ATMOSPHER[ 

1CLEANUp RATE. (THOUSAN~ CFM)",T71,FI0.~120X,"P~RGE TIME OF CONTAl"JM 

2ENT (HOURS)",T71,FIO.~) 

1310 

1320 

1330 

FuR MAT (l 6)( , II THE REI S l'oj 0 T e K I () N E Y F I L T E R" ) 

FORMAT (l6X,"THERE IS NOT ~ CONDENSATE DEMINEk~LIZER") 

fORMAT (16X, "FRACTION .~OD INE ByPASSING CONjJEN~~TE DEMINERAL.I ZERII, 

17X,TU,F9.5) 

1340 FORMAT(l6X,IIIODINE PAt:1TlTION FACTOR (GAS/LIQU.l.D) IN STEAM GENERA TO 

lR ",F7.5) 

1350 FORMAT(16X,5A4,10X,F3,0,oX,F3.0,1 9 X'F3.0) 

1360 FORMAT(l6X'''FREQLJE~CY OF ~NTMT BLDG HIGH YOL ,:,uRGE(TlHES/YR)II, 

1 T7 3 • F i '. 5 ) 1370 FORMAT(Ie>)(,5A4,10X,F3.0,oX,F3.0,14x,F8.2) 

13AO FORMAT(16X,5A4,"RATE(~FM)",24X,Fll.5/16X.SA4,IIIOOINE RELEASE FRACT 

1390 

1400 

1430 

1440 

1450 

1460 

1470 

1 ION",ljx,FIO.5/36X."P~RTIEULATE RELEASE FKACT.i.ON",6X,FI0. 5 ) 

FUrlMAT (16X,IITHERE IS ~"OT A CNTHT BLOG LOW vOLUr.4E PURGE") 

f()RMAT(16A.,"STEA"1 LEAI\ TO TURI::HNE ALDG (Ltjs/H~)",19X,fl0.~) 

FURMATc"O",15X,"THERE. IS NOT ~N UN_SITE LAUNDt:<yll) 

fOR"1AT(lHl) 

FUHMAT(16)1"bA4) 

F 0 I"( ~I A T C1 H 0 , b 7 X , II Ii A S E LJ U S f·(f:, LEA ~ ERA T E - C lJ I"< IE 5 PER YEA R II ) 

FUR MAT ( 1 H IJ , 11 X, IIPR I MA ':'Y ",4 X, "SECOND A RY" , 7 X, II GI~ SST R I Pf> I NG" • 11 x • 

1"t3UILPING VI:.NTILATIONII/12 ... ,IICU()LANT",~X,II<"OOL~NT",5X,21 ("-II), 

3-24 
1480 

1490 

1'+9$ 

1500 

1510 

1520 

1530 

1540 

1550 

1560 

~DECK 

c 

C 

?4X,30("-").,5X,IIBLOwDO,N AIR EJt.CTOR TUTALII/10X,II(~ICROCI/GM) (M 

31 CROCI/GM) SHUTDOWN CU~TINUOUS REACTUR AUXILIARY TURBI~E 

4 VEl'oT OFFbAS EXHAUST") 

FORMATIIHO,1301"-II)) 

FORMATI"O II,A8,212X,lt'EI0.3),8(3X'I PE8.1,lx)) 

FORMAT("O "'Ae,212x,lt~ElO.3),12X,7(3XtlPEIjI.1'lx») 

FORHAT(1HO," TOTAL N~BLE GASESII,lOlX'lPEf;t.l) 

FORHAT (lHO,30X,"TOTAL H .. 3 RELEASED VIA GASEOU~ PATHWAY = ",14," CI 

l/YR"1I31X,"C-14 RELEA$ED ~IA GASEOUS PATHWAY IjI 7.3 CI/YR"1'131X, 

2"AR-41 REL~ASED VIA C~NTA~.NHE~T VENT = 3~ CI/YRII) 

fORMAT (l HO, "0.0 APPEARING IN THE TABLE IN~ICA!ES RELEASE IS LESS 

llHAN 1'.0 CI/YR FOR NO~L~ \PAS, 0.0001 CI/V~ FOt.:< I") 

FORMAT (lHO,54X'''AIRBO~NE PARTICULATE RELEASE ~ATE-CURIES PER YEA~" 

II) 

FORMAT 11 HO, 36X, "wASTE GAS", 16X, 118U ILD I NG VENfl.LA TI ON" 12X, IINUCL I DE" 

1,28X,"SySTEM"'14X,IIRE~CTO~ AUXILIARY FUEL HANDLG",7X,"TOTAL'" 

FORMAT(lHO,AB,2BX,lPE~.1.11X,lPE8.1,4X,lPE8.1'4X,lpE8.1.10X.lPE8.1 

11 

FORMAT ( 1 H 0,11 X ,"PfH MAl:(y" , 1+ X, "SECONDARY", 2!)X, IIljUI LD I NG VE.NT I LAT I O'~" 

1/12X'''COOLANTII,5X,''CO~LA~T".15X,44(''-II),6~,''BLOWDO\llN AIH EJfCTuR 

'2 TOTAL"/I0X,"(MICRO~I/GH) nHCROCI/GM) "tl2X,"FUEL HANDLG REAC10 

3R Aux 1 L IARY TUR8,J..NE VENT OFFGA? tXHA':'STII) 

ENO . 

SlGF2 

SUbROUTINE SIGF2(RLPT,MSIb,NSIG) 

UlMENSION HLPT(NSIAt 

IF (MSIG.f.Y.?) GO TO 30 

1-10 20 1=1, '.,;" I G 

IF (R~PT(I).E(,j.O.O) Gy' TO 20 

IF(I.GT.ll) GO TO 10 

THIS PART Of SUBROUTINE l~ fOR NOBLE GASEb 

I>lV=l O. ~~ (II~T (ALOGI 0 (tjLPT (Il)) -1) 

If (RLpT(I).LT.I0.) D.l.V=l.OO . 

RLPT(I)=AINT(RLPT(It/~IV+0.5)~DIV 

GlJ TO ?n 

nils PAPT lJF SUBROIJTl!~E I~ fOR IODINE 

10 CONTINuE 

C 

ISUB=2 

IF(RLPT(I).GT.l.O)ISU~=1 

D I V= 10.~'" (1111 (ALOG 10 (HLPT ( I ) ) ) - I SUS) 

HLPT(II=AINT(RLPT(It/~IV+O.5)ODIV 

20 CONTINUE 

30 CONTI NilE 

THIS PART UF SUBROUTI~E IS FoR PARTICULAT~S 

[JO 50 I=I,NSIG 

IF (RLPT(I).EQ.O.) GO TO ~O 

uIV=10.~o(INT(ALOG10(~LPT(I)))-2) 

HLPT(I)=AINl (RLPT(I'/~IV+II.5)~DIV 

50 CUNTINUE 

RlTURN 

END 

3-25 

On A930 

008940 

OoA950 

01'139('0 

001; 9 70 

On89RO 

008990 

00 9 010 

009020 

009030 

009040 

ll(}9050 

00901;0 

On9070 

009080 

009090 

00 9 100 

0091}0 

009120 

009130 

009140 

009150 

009160 

009170 

009180 

009190 

009200 

009210 
oOECK 

C 

C 

C 

C 

C 

C 

C 

C 

C 

C 

C 

c 

C 

C 

C 

20 

C 

C 

C 

C 

C 

C 

C 

30 

FIGURE 3-4 

PROGRAM LISTING FOR LIQUID DETER~lINATION 

PGALELQ OOO?~O 

bALE COUE FOR CALCUL~TINb LIYUID EFFLUENTS F~OM pwRS. MOllIFIED 

AUG. lq79 TO IMPLEMENT APPENDIX I TO 10 efR P~RT 50. R~ALTOR 

WATER CONCENTRATIONS ~ALCULATED USING ~ET~ODS-OF DRAFT STANDARD 000300 

ANS 237 IIRAuIOACTIVE ~ATE~IAL? IN PRINCIP~L FLUID ST~EAMS OF 000310 

LIGHT WATER COOLED NU~LEA~ PowER PLANTS" DRA~T DATED MAY 20. 1974 000320 

MODIFIED EDITION OF O~IGEN PROGRAM TO COMPUTE EFFLUENTS FROM 8wR 000330 

AN~ PWR RADWASTE SYST~MS 000340 

0003S0 

STATE~ENTS *PROGRAM P~ALE~Qo AND oLE VEL 20 ARS FOR CDC USERS. 

FOR IBM USERS DELETE !H~S~ STATEMENTS. 

PROGRAM PGALELQ (INPU!,OUTPUT,TAPE5=INPUT'TAP~6=OUTPUT, 

1 TAPf.~) 

REALLETDwN,NOGEN 

REALLETDWA 

COM~O~/MATRIX/A(2500)'LOC(2500) ,NONO(800)'KD(~OO) 

LEVEL ?A,LuC,NONO.KD 

COMMON/CONST/MMN,ERR,MZERU 

COMMON/EY/XTEMP(AOOt,ANEw(10.800)'B(800)t~(800) 

COMMON/FLUXN/REGENT,O~S(800) .1LITE.IA~T,I~OT 

COMMON/OUT/NUCL(~OOt 

CUMMO~/CU~C/PCONC(AOOl,SCUN(AOO) ,Rr NV(800) COMMON/COOL/REACTR,PO~I,S~LDR'BLWDwN,FpEF'HEF'EUFLR,DFIED,nFcs[D, 

1 UFED'DFIuw,DFCSDw,DF~w,EuA,D~A,CWA,DFc~,VFICM,DFCSCM, 

2 DFCW'DFICw,DFCSCW,BDIFR,tDFD,DwFD,C~FD,CMFD,!S,TE,TQ,TC,TCH, 

3 TSTOijc,TSTURD,TSTOR~'DWfL2,DW2,DWF2,'2,T~TO~~,OFID2,DFCSU2,DFD2, 

4 PFLAUN,uwFLR 

COMMO~/APCOUL/RGWFR,OtIRG'DFC5RG,DFRG,TRG'TST~RR,RGFU 

COMMON/ROTE5/HFNRT 

COMMON/CO~P/PWCONC(80U) ,S~UTV(800) ,SCOT(800) pIMEN~ION WURD15(4).WURDl~(5)tWORD56(14),wORD~(2),REACTR(7) 

READ NUCLEAR DATA AND CON~TRUCT TRANSITION MA!RIX 

CALL NtJDATA (NLIBE) 

DO 20 1=2,1TOT NONO(I)=NUNO(I).~ONO(l-l) 

KO(I)=KD(I).NONOCI-l) 

~UILT-rN PAkAM[TERS 

pF=o.ei) 

TBLK=1700. 

MMN=O 

MZERO=?l 

QO 30 J=l,tlOO 

PCUNC(J)=O.o 

SCON(J)=O.O 

RINV(JI=O.O 

CUNTINuE 

HEAD DESCKIPTION OF R~ACTVR AND fJADWASTE TREArMENT PLANT 

PHINT 9n2b 

KEAU 9010,H£ACTR.TY~~ 

PRINT QOIO,KEACTR,TYP& 

R£AU 9~11,~0RD56,POwl 

PRINT 9011,~nRD56.po~1 3-26 

0003AO 

000390 

0004?0 

Oo03AO 

000"'",0 

000670 

000A70 

OOOR90 

000910 

D009?0 

000 9 30 

000940 

OD1170 

U 0 11 90 

Dnl?OO 

IJ01?C:;0 

00 PhO 

OolC70 

lln1290 

001300 

001310 

0013?0 

Uo1330 
C 

C HEAD PAT A fUR LIOUID lNFU~MATlO~ 

C 

~EAD 9022,WURDS6,PCVO~ 

PRlNT q022,wORDS6,PCV~L 

~EAD 9012,wURDS6,LETDWN 

PRINT 9012,wORD56,LET~WN 

READ 9012,wURD56,CBFL~ 

PRINT q012'WORD56,CBF~R 

READ 9011,WORDS6,NOGE~ 

~RINT 9011'WORD56,N06~N 

READ 9022,WURDS6,STMF~ 

PRINT 9022,WORD56,ST

MtR 

HEAU 9~22,WURDS6,WLI 

PRINT q022,wORD56,wLI SCVOL=NOGEI~~WL I 

PRINT 9029,~CVOL 

HEAU 9nS5,BLWDWN,KFNRT 

PRINT 9051,~LwDWN 

RFNRT=i.o 

IF (KFNRT.tU.2) RFNRTjO.O 

PRINT 9041 

HEAD 9~12,wURD56,REGENT 

PRINT 9012,wORDS6,RE~~NT 

If(BLWDWN.EQ.O.O) GO !O 40 

FPEF=o.o05 

tjEF=O.Ol 

PRINT 9030,FPEF,HEF 

GO TO ~(l 

40 FPEF=l.O 

IiEF=I.0 

PRINT Qo30,FPEF,HEF ~o HEAD 9n20,WURDS6,FFCDM 

PRINT q020,wORD56,FFC~M 

If(FFCDM.LT.O.OOl) GO TO bO 

DFCB=10.0 

DFCBCS=2.0 

GO TO 70 

60 uFCA=l.o 

OFCBCS=l.O 

10 READ 9056,wORD18,S~LD~ 

C .. 'A=l.O 

~EAD 9014,UFICW,DFCSCw,DF~w 

READ 9015,TC,TSTORC.CWFD 

PRINT 9045 

pRINT 9016 

PRINT 9011,WORDIR,SBL~R,CWA,CWFO,TC,TSTORC,DFLcw,DfCSCW,DfCW 

HEAU 9013,WURDI8.EDFL~,~O~D8,tDA 

READ 9014,UfIED,DFCSE~,Of~D 

HEAD 9015,Tt,TS,EDFD 

PRINT 9017'WORDIA,EDF~R,E~A,EUFD,TE,TS,OF1ED,~FcSEn,OFEO 

READ 9013,WURDI8.DwFL~,WOHDB,DWA 

READ 9014,DfIDW,DFCSOw,Df~w 

READ 9015,TD,TsTORD.UWfD 

PRINT q~17'WORDIR,DWf~R'D~A,DWfp,TD'TSTOR~,DF!DW,DfCSOW,D~DW 

HEAU 9nI3,~UROI8.DwFL~,wO~D8,~W2 

HEAD 9014,D~ID2,OFCSU~,UFV2 

READ 9015,T2,TSTOR2.D~F2 

PRINT 9011,wORDI8,Owf~2,O~2,DWF2,T2,TSTOR~,DF'D2,DFCS02,DFn2 

REAU 9n31,tiUTfR 

HEAD 9~14,UfICM,DFCSCM,DftM 

READ 9n15,TCM,TSTOR~,CMfD 

3-27 

001350 

0013AO 

001390 

001420 

001430 

001440 

0014C;0 

001460 

001410 

0014AO 

001490 

001520 

001530 

Ool~40 

0015~0 

Oo15AO 

001590 

OOlf.20 

Onlf.30 

001640 

001710 • 

OnI7;?O 

001150 

001170 

On17AO 

OnlRIO 

001f\60 

Oo1R70 

OolARO 

001890 

OolQoo 

Oolq]O 

001920 

0(19)0 

001

940 

OolQso 

001%0 

001970 

OolQAO 

001990 

002000 

002010 

0020?0 

Oo?(l30 

On2040 

002050 

0020('0 
C 

HEAD 903],H('WFR 

READ 9014,O~IRG,DFCSR!,UFHG 

~EAD 90]5,THG,TSTORR,~GFU 

IF(~LWDWN.EQ.O.O) GO !O TS HUFR=~LwU~N*lE30RDTFR/O.347b 

PRINT 9n34'~DFR,CMFD,TC~,TSTO~B,DFICH'DFC~CM,~FCM 

~UFR=aLWDWN*1.0E3*ARS'1.-~DTFH)/O.3476 

pRINT 9035,IjDFR 

If(FFCDM.EQ.O.O) GO T~ 90 

75 IF(REYENT.EU.O.O) GO To 80 

PRINT 903ti,RGWFR,RGFDITRG,TSTORR,DFIRG,DfCSRG'DFRG 

GO TO 9n 

80 RGWfR=O.O 

PRINT 903ij,RGWFR,RGFDITRG,TSTORR,DFIRG,DfC SR6,DFRG 

~O If (KFNRT.El,/.2) GO TO 100 FNRTSO=I.O-l.O/(DfCM*~FCB) 

FNRTSI=I.O-l.0/(DfICM~DFC8) 

FNRTSC=1.0-1.0/(DFCSC~oDF~BCS) 

GO TO 110 

100 FNRTSO=1.0 

FNRTSI=l.O 

FNRTSC=].O 

C READ DATA FOR GA~ INfP'R~ATION 

C 

HO PRINT 9046 

HEAD 9~?1,KGTRWT 

If (KGTI~wT .EQ.O) PRINT 90!JI3 

If (KGTRWT.t:.Q.l) PRIN! 9052 

IF (KEPTRwT.EQ.2) PRINT 9075 

READ 9n12,wORD56,TAUlPRINT Q012,wORD56,TAU' 

READ 9012,WORD56,TAU2 

PRINT 9n12,wORD56,TAU~ 

READ 9012,wORD56,TAU3 

PRINT 9012,wORD56,TAU~ 

GwPRF=i.o 

AXIRf=i.o 

AXPRF=I.O 

CHIRf;].O 

CHPRF=1.0 

CLIRF:;i.o 

CLPRF=1.0 

FHIRF=i.o 

FHPRF=i.o 

CAIRF=i.o 

CAPf<F=1.0 

READ 9065,WORD15.GwHR~ 

If(GWHRE.bT.O.O) GWPRF=l.O-GwHRE/IOO. 

PRINT 90bb,WORD15,GWP~f 

REAU 90b7,WORD18.FHCHKE,FbHRE 

IF(FHCHPE.GT.O.O) FHIHf=l.O-FHCHRE/IOO. 

IF(FHHRE.GT.O.O) fHPRt=l.O-fHHRE/IOO. 

READ 9067,W()RD18.AxCH~E,AXHRE 

IF(AXCHRE.GT.O.O) AXI~F=l.O-AXCHRE/IOO. 

IF(AXHRE.bT.O.O) AXPRt=l.O-AXHRE/IOO. 

pRINT 9n~~.wORDIA,AXI~F,AXPHF 

R~AD 9n22,wORD56,CONVUL 

PRINT 902i,~ORD56,CON~OL 

READ 9069.WURD18,CACHKE,C~HRE,CFM 

IF(CACHPE.GT.O.O) CAI~F=l.O-CACHRE/IOO. 

IF(CAHRE.GT.O.O) CAPRt=l.U-CAHRE/IOO. 

3-28 

002070 

On20M 

002090 ' 

0021]0 

002120 

, 002130 

002140 

002IRO 

002210 

002740 

0';7250 

OOn60 

002290 

On?300 

002320 

002340 

002360 

U02370 

0023AO 

On?3QO 

002

4 00 

u02410 

002420 

002430 

002440 

0024.:;0 

002600 

002610 
C 

READ 9071,WORDIA,CHCHHE,C~HHE 

IFCCHCHPE.GT.O.O) CHIKF=I.0-CHCMRE/I00. 

IFCCHHRE.bT.O.O) CHPR~=1.0-CHHRE/I00. 

EN=2.0 

PHINT 9072,EN 

PHINT 906B'WORD18,CHIHF,C~PHF 

READ 9069,WORD18,CLCH~E,CLHRE,PNOVI 

IfCCLCHRE.GT.O.O) CLIRF=1.0-CLCHRE/IOO. 

IFCCLHRE.GT.O.O) CLPRE=l.O.CLHRE/IOO. 

IFCPNOVl.LT.l.0) GO T~ 120 

PRINT 9070,wORDI8,PNOY1,WQRDl~,CLIRf,CLPRf 

GO TO 130 

120 PHINT 9073 

130 PHINT 9064,TBLK 

READ 9020.WORD56,FVN 

PRINT 9020,wORD56,FVN ~EAD 9020,WORD56,FEJ 

( FE.J=I. O-FEJIl 00. 

l pRINT 9020,l'/ORD56,FEJ 

READ 9020,WORDS6,PFLAUN 

IFCPFLAUN.LE.O.O) PRINT 9048 

PRINT 9020 

C CONVERSION OF UNITS 

C 

c 

EDFLR=EDFLRo48.8 

UWFLR=DWFLRo 48.8 

owFL2=DWFL2°48.8 

C CALCULATE PHIMARY COOLA~T CONCENTRATIONS 

C 

C 

C 

C 

C 

c 

C 

C 

AFPTES=O.O 

DO 140 I=l,lTOT 

140 PCONCCI)=PwCONCCI) 

POW A=Powl 

PCVOA=PCVOLolE3 LETOW~=LETUWNo500.S3 

SBLDA=SBLOR*.3476 

C~FLA=C8FLkoSOO.S3 

150 

CHECK TO SEE IF PRIMA~Y PLANT PARA~ETERS ARE WITHIN SPECIFIED 

KANGES 

IF(POWA.LT.3000 •• 0R.P~WA.~T.3800.) GO TO 150 IF(PCVOA.LT.S.OES.OR.~CVOA.GT.6.0E5) GO TU 150 

If(LETDWA.LT.3.2E4.0R;LET~wA.uT.4.2E4) GO TO 150 

IFCSBLOA.LT.2S0 •• 0R.~~LDA.GT.I000.) GO TO 150 

If(CBfLA.GT.7S00.) GO TO 150 

GO TO 190 

CALCULATE PRIMARY COOLANT ADJUST~ENT FACTORS 

AfPTES=l.O 

RhAL2=(LETDwAoO.q9+0.~I~SOLOA)/PCVOA 

kCSRBc=(LETOWAoO.5+0.~o(S~LDA.C~FLAoO.9»/PCVUA 

H~FP2=(L~TUWAOO.q8+0.~2~(9BLDA+CUFLAoU.9»/PCVOA 

~K2=lbl.7b~POWA/PCVOA 

PO 180 J=ltlTUT 

IF(PCONC(J).EQ.O.O) G~ TO 180 

tll=NUCL(J)/lOOOO 

DL=UI:l(J)l)3b(l0. 

IF(NZ.ED.53.0R.N7.EQ.~S) ~O TO 160 IF(Nl.EQ.37.0R.Nz.Ea.~S) ~O TO 170 

3-29 

0026AO 

002800 

002AfO 

002820 

002830 

002 9 00 

002910 

002920 

u02930 

002940 

Oo2QSO 

002960 

002970 

0029AO 

On40AO 

004100 

004120 

004150 

004160 

004170 

004180 

004190 

0(j4?00 

00 4 210 

004360 

0(14370 
C 

PCONC(J)~PCONC(J)~RK2!(0.U66.UL)/(RCFP2+U~) 

GU TO )RO 

1&0 PCONC(J)=PCONC(J)~RK2!(0.067.UL)/(RHAL2.D~) 

60 TO 180 

170 PCONCeJ)=PCONC(JloRK2!(O.U37.DL)/(RCSRB2+ VL) 

180 CONTINUE 

190 SBLDR~SBLURo48.8 

PCVOL~PCVOL~1000.~0.7/62.& 

C CALCULATE SECONDARY C~OLANT CONCENTRATIONS 

C 

C 

C 

C 

C 

C 

SCVOA=SCVOL~lE3 

~LWDW~=BLwDWN~lE3 

STMFA=STMFR~lE6 

FFCDA=FFCDM 

CnECK TO SEE IF SECON~ARY PLANT PARAMETE~S Aa~ WITHIN SPECIFIED 

RANGES 

C PWTYPE=l.O IS FOR pWR~ .ITH U-TUBE STEAM bENE~ATORS 

C 

C 

pwTYpE=1.0 

UO 200 I=ltlTOT 

200 SCON(I)=SCUTV(I) 

IF(AFPTES.EQ.l.O) GO !O 2)0 

IF(SCYOA.LT.4.0E5.0R.SCVOA.GT.S.OES) GO TU 250 IF(STMFA.LT.l.3E7.0R.~TMF~.GT.l.7E7) GO TV 25U 

IF(BL~DWA.LT.5.0E4.0H.8LW~WA.GT.1.0ES) GO TO ~50 

IF(FFCDA.~T.O.OlI GO TO 2)0 

IFeFNRTSC.LT.0.8999. ~O TU 370 

<.;0 TO 390 

C pWTYPE=?O IS FOR pWR~ wITH ONCE-THROUGH ~TEAM GENERAToRS 

C 

C 

230 PLolTYPJ;:=2.0 

[}O 240 I=ldTOT 

240 SCON(I)=SCOT(I) 

IF(AFPTES.EQ.1.0) GO TO 250 IF(STMFA.LT.1.3E7.0R.~TMF~.GT.l.7E7) GO TV 25U 

IF(FFCDA.LT.0.S5.0R.FtCDA.6T.0.75) GO TO 250 

GO TO 390 

C CALCULATE SECONDARY COOLA~T ADJlISTMENT FA~TOR~ 

C 

250 IF(FFCDA.GT.O.Ol.AND.tFCDA.LT.l.O) FFCDA=0.2 

~HAL3~~ALWUWA*FNRTSI+V.?*HEFoSTMFA*FFCDA)/SCVUA 

IF(FFC~A.GT.O.Ol.AND.tFCD~.LT.l.O) FFCDA=O.l RCSR~3=(HL~PWA*FNRTSC·0.5~fPEFOSTMFA*tFCDA)/S~VOA 

HCFP3=~BLwD~A*FNRTSO.Y.9*fPEF*STMFA*FFCDA)/ScvOA 

IF(PWTyPE.EQ.2.0) GO !O 3 J O 

RK3=4.5E5/ S <":VOA 

DO 320 1=1,ITOT 

IF(SCUN(l).EQ.O.O) GO TO 620 

NZ=NUCL(I)/IOOOO 

PL=UISII)~3000. 

IF(NZ.EQ.5J.OR.Nl.EQ.J5) ~O TO 300 IF(NZ.EQ.37.0R.N7.EQ.~5) bO TO 310 SCON(II=SCON(I)*RK3.(U.17+DL)/(RCFp3.UL)*(PCDI~C(I)/PWCONC(I» 

GO TO 320 

3 r) 0 SCQN (I I =SCON ( I ) oRK 3* (\).17. DL) I (RHAL 3 + DU * (PC OhC ( I ) IPwCQNC ( I) ) 

3-30 

004470 

004490 

004500 

004510 

004520 

004530 

004540 

004910 

004Q20 

OQ4 Q40 

005.120 

00S130 
310 

320 

330 

340 

350 

360 

370 

380 

390 

400 

C 

C 

c 

GO TO 3?0- SCON(1)=SCON(I)~RK3*(U.15.DL)/(kCSRB3.DL).(P~UNC(I)/P~CONCel») 

CONTI('jIlE 

GO TO ~90 

RK3=1.0E5/SCVOA 

1)0 360 I=ltlTOT 

IF(SCONCIl.EQ.O.O) GO TO a60 

NZ=NUCLCI)/loooo 

1)L=DHic I) 030 00. 

IF(NZ.Ea.53.0R.NZ.EG.JS) io TO 340 IF(NZ.EQ.55.0R.NZ.EQ.~7) ,0 TO 350 SCQN(1)=SCONCIl~RK3.(14.0.DL)/(RCFp3.UL)~(PCO~C(ll/pWCONCtl» 

GO TO 360 

SCON (I) =SCON C I) oRK3* (~7. O+DL) I CRHAL3.0U w (PCQI~C (I) /PwCONC I I) ) 

<;0 TO 360 SCONIl)=SCUN(I)~RK3*(1.~.~L)/(RCSRB3+0L).'PCQf~C(I)/PWCONC(I» 

CONTltlIUE 

GO TO 390 RCSRB3=(BLWDWAoFNRTSC!0.5!FPEfoSTHFAoFfCD~)/S~VOA 

ijK3=4.SE5/SCVOA 

UU 380 1=1,1TOT IF(ScO~(I).tQ.O.O) GO TO aao 

NZ=NUCLIl)/lOOOO 

IF(NZ.NE.37.AND.NZ.NE,55) GO TO 38 0 

DL=DIS11)~3bOO. 

RSC=0.t5 

SCON(II=SCON(I)~RK3.(~SC+~L)/(RCSRA3+UL)O~PCQN.C(Il/PWCONCII» 

CONTINIIE 

~LWDWN=AL~DWN~lE3/500.53 

SCVOL=sCVOL~1000./62.! 

5TMFR=ST~FRo2000. 

DO 400 l=ltlTOT 

If (PCONC(I) .EQ.O.O) G~ TO 400 PCONC(I)=PCONC(I)/(DI~(I)~1.6283E13) 

SCON(I)=SCON(I)/(DIS(~)Ol.62a3E13) 

CONTINUE 

COMPUTE REMUVAL CONsT~NT tOR CONDENSATE D~MIN~RALIZER 

IF(FFCD~.GT.O.Ol.AND.tFCDM.LT.l.Ol FFCDM=U.l 

005260 

005270 

0053M 

005410 

005430 

0054t=,0 

no5490 

005500 

005530 

005540 

005560 

CIXRC~(O.9~BLWDWN~RfN~T/D~CM~0.9*STMFR~FP~F~FtCDM)/(SCVOLo7.4B~60. 

1) On5~20 

CIXRCS=(0.5~BLWDWNoRF!1RT/~FCSCM+0.S~STMF~oFPEt~FFCDM)/(SCVOL~7.4b0005630 

1 60.) Oo5~40 

IF(FFCDM.GT.0.Ol.AND.~FCO~.LT.1.0) FFCDM=0.2 CIXRIa=(0.90BLWDWN~RFI~RT/~FIcM.0.9~STMFRo~EF*~FCDM)/(SCVOL~7.4B*b0005650 

1.) On5~60 

DO 410 I=ltlTOT 

NZ=NUCL(I)/IOOOO 0056AO 

pR=CIMC 

IF(NZ.EQ.37.0R.NZ.EQ.~5) PR=clXRCS IF(NZ.EQ.53.0R.N7.EQ.~S) ~R=C1XRI8 

XZHJ=SCON(I)~PR~SCVOL~0.Oia32 

B(I)=~ZHJ 005730 

410 CON T II'lUE 

C 

C CALCULATE RADIOISOTOP~ INVENTORI[S ON CON~EN~~TE RESINS 

C 

CALL SOLVE 

DO 420 l=ltlTOT 

420 ~INV(I)=XT£MP(I) 

CALL EFFTAB 

STol-' 

3-31 

0057QO 
C 

C 

C 

FURMATS 

9010 FORMAT(32~,7A4,16X,A4~ 

9011 FORMAT(16X,13A4,A3,F9,4) 

9012 FOHMAT(16X,14 A4,F8.4) 

FOI1Io1ATS 

00 64 90 

006500 

0065JO 

006f>20 

006630 

006640 

9013 fORMAT(J5X,4A4,A2,AK,t8.0'lX,A4,A2,F6.4) 

9014 FORMAT(20X~F8.0,2C5K,fa.O» 

9015 fORMAT(27X,F6.2,14X,F~.Z,18X,f6.2) 

9016 FORMAT(II0",30X,"FRACnON FRA~TlON COLLE~TlQ~ 

1 f.LOw RATE OF P~A ~ISCHAHGED TIM~ 

2MINATlON FACTORSII120X,II(GAL/DAV)1I23X,"CDAYS) 

0066&0 

On6E>70 

DECAYII/8X,"STREAM 0066AO 

TlME",10X,IIDECONTA006690 

(DAYS)",7X, 006700 

3"11,8~,IICSI,6X,II0THER$") - 9017 FORMAT(2X,4A4,A2.1PE9.2,lX,4cOPF8.4t2X),J(lP~V.2,lX» 

9020 FORMAT(lbX,14A4,F8.4) 

9021 FORMAT(79X,Il) 

9022 fORMAT~16X,14A4,F8.4) 

9026 FOfH1ATC]rlll 

0067}0 

0067:)0 

006140 

006750 

006870. 

9027 fURMAT!16X,IIPLANT CAP~CITl' fACTORII,T75,IIO.800 0,,) 

9029 FORMAT/16X,IIMASS OF W~TER IN ~TEAM GENERATORS (THOUSAND 

]F8 ... ) 

L8S)",T73,006900 

9030 FURMAT Cl6X'''FISSIO~~ PKODU<"T CARRY-OVER F~~CTluN",T75,F6.4/16X, 

l"HALOGEN CARRy':'OVER FHACTION",T75,F6.'+) 

9034 fORMAT(2X'''bLOWDOWIlj'',~OX,lpE9.2,14X,OPf5.3,2l\.'2(f8.3,2X), 

13 C IPE9.2, IX» 

9035 FORMATC?X,IIUNTREATED I:'LOIIII)OWNII,IPEq.2,11 X,1I 

10.000 1.00E+00 ].01,1[+00 i.OOE+OOII) 

1.000 0.000 

9037 fURMAT(72X,f8.2) 

9038 FORMAl (2X,IIREGENERIINT Olb ", IPEQ.2, 14X'OPF~.3.2x,2 (FB.3,2X), 

13CIPE9.2tlX) ) 

9039 FURMAT/3I6.E21.1 4 ) 

9040 FORMAT(I6,E21.14) 

9041 FORMAT(16X,IIPRIMARY TV SE~ONDARY LEAK RATt. (L«:1S/OAY)II,T73, 

006 910 

0069?0 

006930 

007000 

007010 

007020 

I" 75.0000") 9045 FORMAT (/,"0 LIQUID WASTE INPUTS,,) 007060 

9046 FORMAT (/,110 GASEOU~ wASTE INPUTS") 007070 

9048 FURMATC"0",15X,"THERE IS NOT AN ON_SITE L~UNO~v") 0070AO 

9051 FORMAT(l6X,"BLOWDOWN ~ATE CTHOUSAND LBS/H~)".~5X,F8.4) 0~70qO 

9052 FORMAT (l6X,"THERE IS ~ONTiNUOUS STRIPPING OF ~ULL LETDOWN FLO!.oJ") 007100 

9053 fORMAT (lbX,"THERE IS NOT CONTINOUS S!~IR"'IN~ OF FULL LUDWN FLO""Oo71l0 

1) 007120 

9055 fORMAT~36X,F8.4,35X,I~) 007140 

9056 FORMAT(15X,4A4,A2,RX,t8.0) 007150 

9064 fORMAT (16X,"STEAM LEA~ TO TURt:lINE BLDG (LjjS/~K)IItl9X,FlO.ft) 

9065 FORMAT(16X,4A4,6X,F3.0) 

9066 FURMATC16~,4A4,4X,"PA~TlCULATE RELEASE fR~CTl~N",6X,FIO.4) 

9067 FORMAT/16X,5A4,10X,F3,0,6X,F3'0) 

90&8 FORMAT(16X,5A4,"IODIN~ IiiElEASE FRACTION",11X,UO. 4/36X,"PARTIClJLAT 

IE RELEASE FRACTIONII,6l'.,flO.4) 

90&9 f ORMATClbX,5A4,10X,f3,0,6X,F3.0,1 4X,f8.2) 

9010 FORMAT(16X,5A4,IIRATEC~nO",25X,FI0.4/16X'~A4."IOOINE RELEASE FRACT 

1 ION", 11 X ,f! 0 .4;36X, .. PART I~ULA TE RELEASE F~AC T ~,ONII, 6X, FlO. ~) 

9071 fURMAT~16X,5A4.10X,F3,0,6~,F3.0,19X,f3.0) 

9072 fORMAT (16X,IIFREQUENCY OF CNTMT RLOG HIGH VOL pURGE (":"It-IES/YR)II, 0~1330 

] T74 ,F7.4) 

9 0 7 3 ~- 0 rH>1 A T (1 6 X , II THE REI S NOT A C N T I'n B L D G L 0 ~ v 0 ~ PUR G E") 

9075 FORMAT (lbX,IITHERE IS ~ONTlNUOUS LOw VOL PURGE Of VOL. CONTROL 

END 

007350 

ODECK f.YFTAB 

TKIIl007360 

007370 

Oo73AO 

SUt:lROUTI~E 0073QO [FfTAB 

DIMENSION lSOTPC3,100l ~IMENSION ~t.ACTR (7) ,N~'1E (3) ,CwCONC (800) ,DflCO~~ (800) ,CMeONC (800) 

3-32 
C 

C 

C 

C 

30 

C 

C 

C 

C 

C 

C 

C 

C 

50 

~IMENSIO~ TURBDR(800).DWCUN2(800,.EDCUNC(~OO)'IOTHEH(lOO' 

COMMO~/FLUX~/REGENT,OAS(MPO"lLITE,IACT,ITOT 

COMMO~/OUT/NUCL(ROOt 

CUMMON/COOL/REACTR,~O~l,S~LDR'BLWDWN,fPEf'HEf'EUfLR,DFIED,DFCSED, 

1 DfED'DFIUw,DFCSDW.D~~W,EDA'DWA,CWA,DfCM,UFICM,OFCSCM, 

2 OFCW'DFICW'DFCSCW,B01.F~.~DFD,DWFD,CWFD.C~FD,!S,TE,TU,TC,TCM, 

~ TSTORc,TSTORD,TSTORBjOW~~2'DW2.0WF2,T2,T~TO~c,OFID2,OFCSP2,DFD2, 

4 PfLAUN,DWFLR 

COMMO~/APCOOL/RGWfR.OtIRG;DfCSRG.OfRG.TRG'TSTURR,RGFD 

COMMO~/BDTES/RfNRT 

COMMO~/CONC/PCONC(8001,SCYN(800),RINV(800) 

COMMO~/DET/LAUNDRY(251,WL~UND(25) 

H3COP* IS THE PWR TRI!IU~ PRIMARY COOLANT CON~ENTRATION I~ 

UCI/GM 

H3PRPW=0.4 4t POWl 

H3COPW=1.0 

DO 30 J=l,ITOT 

CwCONC(J'=O.O 

E,DCONC(J)=O.O 

uwCONC(J'=O.O 

DWCON2(J)=0.0 

CMCONC(J'=O.O 

NZ=NUCL(JI/IOOOO 

IF(NZ.EQ.36.0R.NZ.EQ.~~' ~O TO 30 CWCONC(J,=PCONC(J,*CW~ 

EDCONC(J)=P~ONC(J,*EOA 

UWCONC(J,=PCONC(J,4tDW1 

()WCON2(J)=PCONC(J,*DW~ 

CMCONC(J,=SCON(J) 

LJfCVCS=50. 

If(NZ.EQ.IIUFCVCS=l.O 

IF(NZ.EQ.35.0R.N7.EQ.~J) VFCVCS=lOO. 

IF(NZ.EQ.37.0R.N7.EQ.55) ~FCVCS=2. 

CWCONC(J,=CWCONC(J)/OtC'VCS 

CONTINuE 

CALCULATE RADIOACTIVITY AfTER COLLECTION AT A CONSTANT HATE 

CALL COLLECT(TC*B6400"CWCONC,ITOT) 

CALL COLLECT(TE*A6400"EOCONC,ITOT) 

CALL COLLECT(TO*~6400"DW~ONc,ITOTI 

CALL COLLECT(T2*A640 0"OW CON2,ITOT) 

CALL COLLECT(TCM*8640~.,CMCONC,ITOT) 

IF(REGENT.LE.O.O) GO TO 50 

CALL STORAG(TRG~86400;,RINV'ITOT' 

va 100 1=I,1TOT 

NZ=NUCL(I)/looOO 

TUR80~(II=199l.*5.*SC~N(II 

If (NZ.EI'l.ll GO TO 100 IF(NZ.EQ.35.aR.Nz.EQ.~31 ~O TO 60 IF(NZ.EQ.37.0R.N7.EQ.~51 ~O TO 70 

CHEMICAL TR~ATMENT fO~ OT~ER CATIONS 

CWCONC(II=CWCONC(I)/utcw 

~DCONC(I)=EUCONC(I)/D~ED 

UWCONC(I)=UwCONC(I)/D~Dw 

DWCON'(I)=DwCON2(1)/D~D2 

CMCONC(II=CMCONC(I,*('.0-oDTFR~(1.0-CMFD/UFCMI) 

TU TREAT PW~ TURRINE olJILUING FLOOR DRAIN~ T~~nUGH DIRTY WASTE 

3-33 

007790 

007800 

007810 

007920 

007930 

007940 

007950 

007960 

007970 

On19AO 

007990 

OOAOOO 

OOA010 

008020 

OnA050 

008010 

OnAOM 

00R090 

00R100 

008110 

008120 

008130 

008210 

008220 

008230 

008240 

On R250 

008r60 

00 8 210 

OOR2AO 

OOA290 

00831)0 

008310 

008320 

008330 

0';8340 

008350 
c 

C 

C 

C 

C 

C 

c 

c 

C 

C 

c 

60 

70 

100 

C 

C 

C 

110 

130 

SYSTE~, DELETE C fOR ~OMMtNT ON CARDS BELOw, UNTIL NEXT MtSSAGE 

RINV ~I)=RINV CI,/DtRG TURBD~(I)=1~91.o5.oSC~N(I)oFpEF 

TURBDij~I)=1991.o5.oSC~N(I)ofPEF/DFDW 

eo To 100 

CHEMICAL TREATMENT fO~ ANlONS 

CWCONC(I)=CwCONC(I)/D~IC~ 

~DCONC~I)=EDCONC(I,'DEIED 

DWCONCcI)=DwCONC(I)/D~IPw 

DWCON~cI,=OWCON2(I,/DtIa2 

CMCONC(I,=CMCONC(I,.(J.O-~DTFRO(I.0-CMfD/~FICM,) 

~lNV (I,=RINV (I)/DtIRG TURBDR~I'=1~91.o5.osCHN(I)oHEF 

TURBDRcI,=1991. o5.oSCp'N(IloHEF/DFIOW 

GO TO ioo 

CHEMICAL TREATMENT FO~ RB AND CS 

CWCONCCII=CWCONCCII/DtCSCW 

~UCONCcI,=EDCONCcI,/DtCSE~ 

DWCONCcI,=DwCONCCI,/DtCSDw 

uwCON2cI,=DWCON2CI)/DtCSD2 CMCONC~I,=CMCO~CCI'O('.0-~DTFRo(I.0-CMFD/~fCS~M)' 

RINV cI,=kINV CI"otCSR~ 

TURBDRCI'=1~91.*5.*SC~N(I)*FPEF 

TU~~DRCI,=1991.*5.*SC~N(l'*fPEF/OFCSDW 

CONTINuE 

COMPUTE RADIOACTIVE D~CAY DURING PROcESSl'~G A'~D SAMPLING 

CALL STORAGCTSTORC*B6~00.,CWCUNC,lTOT' 

CALL STOKAGCTS*86400.iEDCUNC,lTUT) 

CALL STORAGCTSTORD*8 64 00.,DWCONC,ITOT) 

CALL STORAGCTSTOR2*86400.,DWCUN 2,ITOT' 

CALL STORAGCTSTORB*86~00.,CMCUNC,ITOT) 

CALL STO~AGCTSTORR*86~00.'RINV,ITOT) 

CALL STORAGC21600.,TU8BDH.ITOT) 

po 130 l=l,ITOT 

At;LOw=o.O 

IFcREGENT.LT.O.OOl) G~ TO 110 ABLOW=RINV(I)*292.4*R~fD/~EGENT 

ABLOW=A8LOW+ALWDWN*19~1.*~MCONC(I)*(1.0-R~NRT) 

01CONCC I) =ABLOW 

GWFR=SBLDR*CWFD*0.02 8J2 

~UFLR=EDFLk*EDFD*0.02~32 

UWFR=P~FLR*DWFD*0.028J2 

UWfR2=DWFL2*DWF2*0.02~32 

TPLRpW=cwFR*CWA+EDFLk~EDA+DWFk*DwA+DwFR2*~W2 

H3RLPw=TPLR~w*H3COPW 

IFcH38LPW.GT.0.9*H3PR~W' ~3RLPW=0.9*H3PRP. 

RH3RLP=H3~LPW/I0. 

INTRlt'l=RHJKLP 

lH3RLP=INTRIM*10 

TOTAL=O.O 

Il=ILITE+IACT+} 

I..lU 140 I=ltlTUT 

NZ=NuCLcI)/lnOOO 

IFCNZ.EQ.]6.0R.NZ.EO.~4' bO TU 140 

O!SI=PrsCI)*1.62R3E13 3-34 

0;)8360 

008370 

008380 

008390 

008400 

0084}0 

00B420 

008430 

008440 

008450 

008460 

008410 

Oo84AO 

008490 

008500 

Oii85}0 

008520 

008530 

008540 

00 8 5,0 

008560 

008510 

Oii8580 

008590 

008600 

008610 

Oo8~20 

0081'130 

008f>40 

008f>,0 

008660 

008610 

008750 

008170 

008780 

0(iAB60 

008B70 

00A8BO 

008890 

008900 

008910 

008 930 

008940 

008950 

009050 
140 

150 

152 

155 

160 

C~CONC(I)=DlSI~cCWCON~CI)~CWF~+EOCONC(I)O~DFL~) 

OWCONC(I)=CUWCONC(I,~~WFR.OWCUN2CI)~DWFR2)~Dl?I 

CMCONCCI)=CMCUNCCI)oD!SI 

TURBD~CI)=TURBDRCI)oDlSI 

IFCNUCLCI).EQ.I0030t ~O TO 140 

TOTAL =TUTAL .CWCONCC~).O.CONCCI).CMCONCCl).TUR~DRCII 

CONTINuE 

AOI=0.i6 

AOR=C~OI.TOTALI/TOTAL 

SCNORH=o.O 

SAPRHbO .0 

SSEC=O '.0 

SCWAST=O.O 

50wAST=O.O 

SABLOw=O.O 

5TB=000 

STOTAL =0.0 

PAPRIH=n.O 

PSEC=O'.O 

PCWAST=O.O 

PIJwAST=o.O 

PAHLOI'I=O.O 

PTB=O.O 

PTOTAL =0.0 

PNORM=O.O 

TLAUNP=O.O 

CTOTAL=O.O 

PRINT QOOI. REACTR 

PRINT Q002 

PRINT 9010 

r\OUNT~=l 

DO 180 l=l.1TOT 

IFCI.EQ.Ill PRINT 90U 

NZ=NUCLCII/I0000 

IFCNZ.EQ.3b.OR.NZ.EQ.~41 ~O TO 180 

IFCNZ.EQ.ll GO TO 1M DISI=P1SCI)~1.6283E.IJ 

APRIM=PCONCCI)oDISI -

ASEC=SCONCl)~DISI 

CwASTE=cwcu,,"CCI) 

~WASTI;=DwCONCCII 

AtiLOW=CMCONCCII 

Tb=TU~RDRCI) 

TUTAL~cwA~TE+D~ASTE.A~LOW·TB 

TOTALN=TUTAL~AOR 

NUCLI=NUCLCII 

XLAUND=O.O 

IF CI.GT.155.AND.I.LT.190) GO TO 152 

IF CI.EQ.22~1 GO TO 1~2 

DO 150 L=1.25 IFCLAU~DRYCL).EQ.NUCL~) XLAUND=WLAUNDCLloPFLAUN 

CONTINUE 

CONTINuE 

TOTALG=TOTALN+xLAUND 

IFCTOTALG.LT.0.OOOOl1 GO TO 1

6 0 

ISuB=2 

IF CTOTALG.uT.1.)ISUH~1 

DIV=IO~o.CINTCALOG10C!OTALG)I~ISUHI 

TOTALG=AINTCTOTALG/DIV+O.~)·DIV 

IFCNUCLCI).[Q.10030t TOTALN=TOTAL 

IFCNZ.EQ.11 GO TO 162 SAPRIH=SAPKI~+APRIM 

S~t::C=SSEC+A~EC 

3-35 

U09090 

009100 

009110 

009120 

009140 

009150 

009170 

009190 

On9200 

0092}0 

009220 

009230 

00924-0 

009?50 

0092AO 

On9?90 

009300 

0093)0 

009320 

On9330 

On9340 

009370 

0093RO 

009390 

009400 

009430 

009440 

009630 

009710 

009720 

009730 

009140 

0098)0 
SABLO.=SA~LOW+ABLOW 

SCWAST=ScwAST+CWASTE 

SDWAST=SDwAST+DWASTE 

5T6=ST8+T8 

5TOTAL=STOTAL+TOTAL 

SCNORM=SCNORM+TOTALN 

TLAUNP=TLAUND+XLAUND 

CTOTAL=CTOTAL+TOTALG 

162 IfCTOTALG.LT.0.00001) GO TO 1dO 

168 IFC MOPCKOUNTR.50'.NE.Q, GO TO 170 

PRINT 9000, REACTR 

RRINT 9002 

170 CALL NOAHCNUCL~I'.NAM~) 

THALf~8.0225E-6/DIS~Ii 

PRINT 9003.NAME.THALfIAPRIH.ASEC.CWASTE.U.AST~.ABLOW, 

ITB.TOTAL.TUTALN,XLAUN~.TOTALG 

KOUNT8=KOUNTR+l 

IFCNZ.EQ.l) GO TO 180 PAPRI~=PAPRIM+APRIM 

PSEC=PSEC+ASEC 

pCWAST=PCWAST+CWASTE 

pDWAST:PUwAST.OWASTE 

PABLow=PA~LOW+ABLO~ 

PTH=PTEI+T~ 

PTOTAL=PTUTAL+TOTAL 

PNORll=PNORM+ TOTALN 

180 CONTINUE 

PAPRI~=SAPRIM-PAPRIM 

PSEC=SSEC-PSEC 

PCWAST=SCWAST-PCWAST 

PUWAST=SU~A~T-pDwAST 

PAHLow=SAoLOW-PABLOW 

PTB=ST8-PTB 

PTOTAL=STUTAL-PTOTAL 

PNORM=SCNURM-PNORM 

ISUI::jC=? 

IF (CTOTAL.GT.l,'ISUB~=l 

DIV=lO.**(I~TCALOGIOCCTOTAL»~ISU8C) 

CTOTAL=AINTCCTOTAL/DI~+O.')*DIV 

IF(PNORM.LT.O.OOOOl~ ~O TU 190 

DIv=10.**CINTCALOGIVCPNURH)'-2) PNORMT=AINTCPNOR~/DIV~e.5)*DIV 

GO TO 200 

190 pNORMT=PNORM 

200 pRINT 9004, PAPRIM.~SEC'PCWAST,PDWAST.PABLO~.PTB,PTOTAL.PNOR~. 

PNORMT 

SAPRIM.~SEC,SCWAST.SDWAST,SARLow,STB.STOTAL'SCNORM, 1 

pRINT 9005. 

1 TLAUND,~TOT~L 

pRINT 9012, IH3RLP 

009A20 

0"09830 

009840 

009P50 

0098,.,0 

009890 

009900 

009910 

009940 

010030 

010050 

0;0060 

0;0070 

0100RO 

0100QO 

010100 

0;0110 

0;0]40 

U;0150 

010160 

0]0]70 

O,OlRO 

OlO]QO 

010200 

0;02]0 

Oi0220 

0'0250 

010260 

010270 

0102Ro 

010290 

fll03]O 

0;0320 

0,0380 

0103QO 

0;0400 

0]0410 

pRINT 9013 RtTUR~ 010420 

9000 FORMATClHl.20X,7A4.1I ~I(JUID EFFLUENTS CCOl'4TIN':lED)II) 0104AO 

9001 FORMATClHl,20X,7A4." qQUIO EfFLUENTS", 0;0490 

9002 fORMATC]HO.~sX."l\NNUA~ RELEASES TO DI5ChAHGE ~ANAL"/20X,IICOOLANT CU10sOO 

lONCENTRATlONSII,57 ("""1.." ADJUSTED UETE~GEN! TOTAL 11/" NllCLlOO;05io 

.?E I1ALF-LIFE PRIMARY SECONDARy BOt-<ON ~S MISC. WASTES SECON010520 

3l)ARY TlJR~ bLDG TOTAL. Lwb TOTAL ~ASH.S "/lOX, 0;OS30 

4"CDAYS, "2("(t-lICRO ~I/ML)II)tlX.4C"CCURH:.S) ")."(ClJRIES) ". 010540 

5" (CI/yR) CCI/YRl (CI/yR,II, 0105<;0 

9003 FORMAT(lX,A2,I3.Al,2X'lPt~.2,2(2X.E9.2.2K),OP'7(IX.F9.5.1X"FIO.5) 

9004 fORMAT (lX."ALL OTHERSII.~X'lPE9.2,4X.E"'.2.0P.~",6CIX.F9.s.1X) .3X. 

1 "0.00ijnO".lx,FI0.5, 3-36 
9005 FORMAT (II lOTALlllt" iEACEPT TRITIUM) ".It'·E9.~,4X,E9.2,OP·.2X, 

1 1(lX'F9.S.1X),FIO.S) 

9010 FORMAT(II CORROSION ANI:' ACTIVATION PRODUCTSII) 

9011 FORMATC"OFISSION PRODUCTS") 

9012 FORMAT ClHOtlX,IITRITIU~ RI;.LEASE"tl2X,I3," CURIES PER YEAR") 

9013 FORMAT~lHO,IX,"NOTEt ,00000 INDICATtS THAT T~~ VALUE IS LhSS THAN 

11. OE-If.") 

9014 FORMAT(3X.I0(2X,A2.13.Al)/3X,10(2X,A2.13tAl)' 

END 

-DECK BLKDAT 

~LOCK DATA BLKDAT 

c 

C PWCONC CONTAINS PRIMARy COOLANT CONCENTRATION~ FOR PWRS. SCUTV 

C AND SCOT CONTAIN SECONDARY COOLANT CONCENrRAT~ONS FOH PLANTS 

C WITH U~TU8E STEAM GEN~RATORS AND FOR ~LANTS WlTH ONCE-THRUUGH 

C ST~AM GENERATORS, RESPECTIVELY. 

C 

COMMO~/CONP/PWCONc(eOO).S~UTV(800),SCOT(aaO) 

COMMON/DET/LAUNDRY(251,WLAUND(25) 

DATA P.wCONC/3600'4.7E~2.61*0,3.1E-3.400,1.6E~~,5*0,1.2E-3,300,3.0 E 

1_4,0.O,4.6E_3'200,5.3~_~,1700,5.1E_4,102oQ,2'~E-3,64*0,2.2E-3,68*0 

2,1.6E-2,1800,1.9E_l.4fO,li4E_4,300,1.2E-S,500'9.6E-4,4.6E~4,5.2E-6 

3,900"~2E-3,1100,3.9E44.0.0'2.8E-4,15*0,6.4E-j,4.7E-3,1600,7.5E-3, 

41500,~~nE-2,0.o,2000,!.3E-3'104*0'1.9~-4.2.4E-2,12*0,1.5E-3,7.7 3 

5,4.5E_2,500'1.7E-3.2.1E-l;4*0.1.4E~I,5*0,a.4E~I,2*0,7.IE.B,2*0.2. b 

6E-l,8*o,8.7E.4,3*0,9.~E-3,13*0,1.3E-2'2.S~_2,~*0,1.5E-4,1200,2.8E73.300,3. 9E -3,92*01 

OATA SCUTVI3600,1.5E·2,Q7*0'1.3E-7,400,6.~E-e'5*0,4.9E.8,300,1.2Ela,0.0'1.9E-7,2*0,?.2E~A,17*0,2.1E-8,102*0'8.1~-8,64*0,8.4E-8,68° , 

27.5t.e,]8~0,5.3E-7,400,5.7E-9,300,4.9~-10'5*~'2.BE·B,3.2E-9,2.1[-1 

30,9*0,1.2£-7,11*0,1.6~_8,n.0,1.IE-8,15*0,2.5E-7,1.ll-7,16~0,3.1E-7 

4.15*O,3.7E-b,0.0.20*O,5,3~.8,104~O,7.8E-9,2.2~_7,12*0,5.4~-8,2.9E­

~e,I.BE_6,5*O,6.6F_8,3,IE_b,4*a,4.8E-6'500'2.4~-6,2*O,3.3E-1,200,6. 

66E_b,e*0,4.0E_B,300,4.4E.7,13*0,5.2E-7,9.3E-7,5*0,6.1E_9,12*0,I.OE 

1-7 ,3*0,I. oE-7,92*01 

DATA 5COT/3600,1.0E~7,61~O,6.9E-9'4*O,3.6~_9,~*0,2.7E-9,300,6.7E-l 

10,o.o,i.OE-~,2*0,1.2E~9.17*0,1.1E-9'102*0'5.b~-9,64*0,4.9E-9,6B*0, 

21.BE-e,IS* ,6.0E_7,4*0,3.1E-I0,300,2.1E-11,5*U,2.1E-9,9.7[-10,1.2E 

3-11,9*o,9.3E_9,1100,8.7E-I0,0.0,6.2E-I0,1~*0,1.4E-8,1.oE-e,16*o,1. 

47E.8,15*0,2.0E_7,0.0,~0.O'2.9E-9,10~.O,4.~E-IO,5.1E-8,1200,3.3E-9, 

Sl.5E_8,S.2E-B,5*0,3.8~_9,'.4E-7,400'1.6E-7,50U,3.8E-7,2*0'3.0E.8,2 

6*O,3.0E_7,~00,3.6E_9,~*O,3.9E-8'1300,2.9E-8,~.6E-8,5*0,3.3E-I0,12* 

70,6.2E.9,300,a.7E-9,9~*01 

C 

c LAUND~y ARE THE RADIOISOTUPES IN THE DETE~GEN! WASTES. 

C WLAUN~ ARE THE CORRE~PONUiNG CONcENTRATIO~S. 

C 

OATA LAUNURY/150320,240510,250540'260550,2605~0,270580,270600,280 6 

130,380A90,380900.390910,4U0950,410950.420990,441030,441069,471101, 

2511240,531310,551340,~513bO,551370.561400,581410,5814 401 

~ATA WLAUNU/l.8E_4,4.~E.3,3.8[_3,7.2E-3,2,2E-~,7.9E-3,1.4E.2,1.7E13,R.at~5,1.3E-5,B.4E-~,I.lE-3.1.9E.3,6.0E-5,2.9E-4,R,9E-3.1.2£.3. 4 

2.3E_4,i.6E-3,1.1E_2,3.7E-4,1.~E-2,9.1E~4,2.3E-4,3.9E- 31 

END 

*DECK SOLVE 

SUBROUTINE SOLVE 

COMMON IE~I X TEMP (R 0 0) , XNE \II ( 10 ,~O 0) ,f:l (800) , lJ (800) 

COMMON/FLUXN/REGENT,D~S(800),ILITE,lACT,I!OT 

L}O 10 1=1 tITOT 

P ( I) =-D 1 5 ( I ) 

10 XH.Mp(I)=/).U 

uELT=REGE~T.R6400. 

CALL DECAY(l,OELT,ITO!) 

3-37 

0;0740 

0;0750 

0;07M 

0;0170 

0;0780 

0;0790 

0;1230 

011240 

Ul1250 

0;1350 

0;1360 

011370 

0;1390 
30 

oDECK 

c 

c 

C 

C 

C 

C 

C 

C 

C 

C 

C 

C 

C 

c 

c 

c 

C 

C 

C 

C 

10 

CALL_TERM(UELT,I,ITOT) 

CALL EQUIL(l,ITOT) 

DO 30 1=1, lTOT 

XTEMP(I)=XNEW(I.I) 

RETURfl! 

END 

TERM 

SUBROUTINE TERM(T,M,I!OT) 

TERM ~DDS ONE TERM TO E~C~ ELEMENT OF THE SOL~TION VECTOR 

CSUM(J) IS THE CURRENT APPROXIMATION TO XNEW'~.J) 

CIMO(J) IS THE VECTOR CONTAINING THE LAST TERM ADDED TO EACH 

ELEME~T OF CSUM(J) 

CIMN(J) IS THE VECTOR CONTAINING I/TON TI~ES !HE NEW TERM TO BE 

ADOEO TO CSUH(J) 

CIMN(J) IS GENERATED tROM CIMO(J) BY A RE~URSION RELATIONI 

CIMN(Jl= SUM OvE~ L QF (~P(J;L)O~IMO'L» 

AP(I,J) IS THE REDucE~ TRANSITION MAT~IX fOR !HE LONG-LIVED 

NUCLIDES 

LOGIC~LLONG 

DIMENSION AP(2500) .CIMR(800) ,CIMO(BOO) ,CHIN( 80 0) ,C5UM(800) 

DIMENSION QUB(50),LOC~(25PO),NONP(800) 

011410 

011420 

Oi1430 

011440 

011450 

0;1460 

011480 

oi 1490 

011500 

011510 

011520 

0)1530 

0,1540 

011550 

011560 

0,1570 

011580 

Oi15QO 

Oi1600 

COMMON/SERIES/ XP(Ao01,XP6R(ROO),LONGCBOO) 

COMMON/CONST/MMN,ERR,MzERU 

COMMON/EW/ATEMP(AOOt,XNEW(10,~00)~B(BOO).~(800) 

COMMON/MATRIX/A(2500).LOC(2500),NONO(HOO)'KD'~00) 011730 L~VEL 2,A,LOC,NONo,KO 0~0540 COMMON/TERMU/DD(100t'~XP(lOO),QUEUE(50),N~U(50)'NQUEUE(50),NQC800)OI1 50 

NUL=O 0]1760 

NN=O 0,1770 

FIRST CONSTRUCT REDUC~D T~ANsITION MAT~IX FOR LUNG-LIVED ISOTOPES 011780 

00 220 L=l,lTOT 0117QO 

IF(.NOT.LlH~\:i(L» GO TIJ 21U 011~00 

NUM=NO~OCL) 011 8 10 

IF(M.GT.MMN.OR.M.EQ.MtERQ) NUH=KDCL) 0,1820 CIM~(L)=B(L) 0,1830 

IF(NUM~LE.NUL) GO TO ~~O Oi1840 

NS=NN+l 0;18~0 

N=NUL Oi1860 

NL=NUM_NUL 0;1870 

PO 200 Nl=I,NL OilA80 

N=N+l OIIAQO 

J=LOC(N) 0;1900 

DJ=-D(J) 0)1910 

0,1920 

THIS IS A TEST TO SE! IF ONE Of THE ASSYMPTO!IC SOLUTIONS APPLIES011930 

Oil Q40 

Oi1950 

Oil960 

OilnO 

OilQAO 

Oi1990 

0,2000 

0,2010 

0,2020 

0,2030 

012040 

Oi2050 

0,2060 

012070 

Oi20AO 

IF (,NOT .LONG (J» 

NN=NN+l 

AP(NN)=A(N) 

LOCP(NN)=J 

GO TO 200 

GO T~ 10 

GOING RACK UP THE CHA~N TU FIND A PARENT ~HIC~ IS NOT IN 

El.IUILl8RIUM 

NSAVE=il 

l.IUE=A(N)/DJ 

l>RB=I.0 

CIMB(L)=CIMH(L)+QUE*B'~) 

NQ(L)=o 3-38 
NQ(J)=L 20 NUX=NONn(J) 

IF(M.GT.MMN.OR.M.Ea.MtERO) NUX=KDCJ) 

NUF=O 

IF(J.~T.l) NUF=NONO(~-l) 

N).::;NU,l(_NUF 

If(NX.LT.l) GO TO 19Q 

I<.=NUF 

DO lAO KK=l,NX 

K=K +1 

Jl::;LOC'cK) 

DJ::;-D (J1 ) 

KP=J 

30 If(Jl.EQ.NQ(KP» GO!O l~O 

KP:;r~Q (KP) 

If(KP.NE.O) GO TO 30 AKDJQ=QUE~A(K)/DJ 

IF(.NOT.LuNGCJi» GO TO 1~0 

TRM=l.O-XtJ(.Jl) 

IF(TRM.LT.l.0E-6) GO !O liO 

NQ(Jl)=J 

1::;1 

KP=Jl 

40 UD(I)=_D(KP) 

DXP(I)=XP(KI-') 

I\P:;NQ(KP) 

IF(KP.EQ.O) GO TO 50 

1=1·1 

IF(I.LE.IOO) GO TO 40 

C IF WUEUE 0F SHORT-LIV~D NUCLIDES EXCEEDS 100 ISOTOPES, TERMINATE 

C CHAIN AND WRITE MESSA~E 

PRl~T QOOO, M.L,JIJJ,AKDJQ 

9000 FORMAT(II1T UO LONG A QUEUE HAS BEEN FORMED IN !ERMII,415,E12.5) 

GU TO 190 

50 j;iATM=O.EO 

IM=!-1 

00 110 I=2tlM 

OL=LlD(I) 

XPL=DXPCI) 

IjATE=O.EO 

Jl=I-l 

C ~ R VONDY fORM OF 8AT~MAN EQUATIONS -- ORNL-T~-36l 

PO 100 KB=ltIl 

XPJ=DXPC Ktj ) IF(XPL~xPJ.LT.ERR) GO TO 100 

I,)I<=DO(K8) 

~ROO=(DL/DK_l.O) 

DKR=PKOD 

If( A8S(PROD).GT.l.E-') GU TO 60 

C USE THIS FORM FOR TWO NEA~LY EQUAL HALf-LIVES PROD=T~DK~XPJ~(1.0_0.~~(DL-OJ)~T) 

GO TO 70 

60 RROD=(XPJ-XPL)/PROO 

PkOl=XPJ/UKH 

70 PI::;I.0 

51=2./(Dl\oT) 

DO 90 JI<=ltll 

IF(JK.EQ.Kd) GO TO 90 

S=l.O-DK/LJLl(JK) 

IF( ABSCS).GT.l.E-4t ~O TO 80 IF(A8S(DKH).GT.I.OE-4~ PkOD=PRUI 

S=SI 

80 PI=PIos 

3-39 

0}2090 

012100 

0,2110 

0;2120 

0;2130 

0;2140 

0;2150 

0;2160 

0,2170 

0121AO 

0;2190 

012?00 

0,2210 

0;2220 

012230 

0;2246 

012250 

0,2760 

0,2270 

0;27AO 

012290 

0,2300 

0,2310 

0;2320 

0;2330 

012340 

012350 

0;2360 

0;2370 

0'23~0 

0,2390 

0,2400 

0;2410 

0;24 20 

0,2430 

0;2440 

012450 

Oj2460 

0,2470 

0124AO 

0;2490 

0;2500 

0;2510 

0,2520 

0;2530 

0,2540 

012550 

012560 

0;2570 

0;2580 

0;2590 

0}2600 

0,2610 

012620 

0;2630 

0;2640 

0;2(,50 

0;2660 

0;2670 

0;2MO 

0,2690 

0;2700 

0}2710 
IF(ARSCPI1.&T.I.E25t GO TO 100 ~;2720 CONTI~UE 0;2730 

BATE=aATE.PROD/PI 0;2740 

90 

100 CONTI~UE 012750 

C IF SU~MATION IS NEGATAVE, SET EQUAL TO ZEHO A~D PRINT MESSAGE 012760 

IF(BATE.LT.0.EO)PRINT1001,L,IM,BATE,BATM 0;2770 

9001 FOHMATC"1~ArE IS NEGATIVE IN TEHM. THERE ~RE MORE THAN TWU SHOPT-LO;2780 

lIVED ~UCLIOES IN A €~AIN WITH NEARLY EQUAL DI~GONAL ELEMENTS"/ 0;2790 

2" LtIM,BATE,BATM = ";~I!j,lP2E12.5) - 0;2800 

IF(AATE.LT.O.EO)RATE=9.~0 0128JO BATM=~ATM.BATE 012820 

110 CONTI~UE 0;2830 DRA=AKDJQ*~J*(TRM-BAT~)/T~M 012840 

GO To i30 0,2850 

120 ~RA=AKDJQ*AMAXi(DRB,O,O).PJ 012860 

130 IF(NS.GT.NN) GO TO 159 012870 

DO 140 lJ=NS,NN 0,2880 

IF (LOCP (LJ) .NE.Jl) G~ TO 140 0,2890 

AP (LJ) =AP (LJ) .DRA 012900 

GO TO iAO 012910 

140 CONTINlIE 012920 

150 NN=NN.] 0;2910 

AP(NN)=DRA 012940 

LOCP(~N)=Jl 0;2950 

6U TO lAO Oj2960 

16G IF(AKDJa.LE.l.0E-06t GO To ieo Oi2970 IFCNS~VE.GE.50) GO TO leo 0;2980 

170 NSAVE=NSAV[.1 012990 N~UEuECNSAVE)=Jl 0,3000 

UUEUE(NSAvt)=AKDJQ 013010 NQU(N~AVE)=~ 0;3020 

QU8(N~AVE)=ORB-l./CDJ'T) 0;3030 

lAO CONTINUE 0;3040 

190 IF(NSAvE.LE.O) GO TO ~OO 0;3050 

J=NQUEUE(NSAVE) 0;3060 

QUE=QUEUE(NSAVE) 013070 

NU(Jl=NQU(NSAVEl 0;3080 

DRH=QUBCNSAVEI 013090 CIM~(L)=CIMB(L).QUE.B'J).eMAX1(OR8,0.01 013100 

NSAVE~NSAVE-l 0131JO 

GO TO 20 0;3120 

200 CONTINUE 013130 

210 NUL=NONO(L) Oi3140 

NONP(L)=NN 013150 

220 

C 

CUNTINuE 013160 

FIND ~ORM OF MATRIX AND EDTIMATE ERROR AS DESCRIHED IN LAPIDUS 0;3170 

AND LUUS, OPTIMAL CON!ROL OF ENGINEERING P.ROC~SSES BLAISDELL 1967 Oi31RO 

FIND THE MINIMUM OF TME ~AXIMUM ROw SUM AND TME MAXIMUM COLUMN SU M0131QO 

C 

ASUM =0. 0 - - °132 00 

ASUHJ=O.O 0132}0 

NUL=l 0;3220 

DO 250 I=1,ITOT 013230 

IF(.NOT.LONu(I» GO T~ 250 Oi3240 

01=-D(I)*T 013250 

AJ=DI 013260 

NUM=NONP(I) Oi3270 

IF (NUL .GT .fW"1) GO TO ~40 0132~0 

00 230 N=NUL,NUM 013290 

230 AJ=AJ+AP(N) 0;3300 

240 AI=DI+DI 013310 

IF(AI.r,T.ASUM) ASUM =AI Oi3320 

If CAJ.GT.ASUMJ) ASUM.J=~J Oi3330 

C 

3-40 
250 NUL=NONPC~)+} Oi3340 IFCASU~J.LT.ASUM) ASUM=AbUMJ 013350 

USE ASUM TO DECIDE HOW MAI~y TERMS ARE REQUI~E~ AND ESTIMATE ERRO~ 0133~0 

NLARGE=3.5*ASUM +5. 013370 KLARG~=NLAR~E 0]3380 

ERR1=~xpeASuM )oeASUM *2.71828/XLAR~E)**N~ARb~/SQRTe6.2832*XLAPGE)013390 

IFeER~1.GT.l.E~3) PRLNT ~002. ERR1,ASUM .NLARGE 013400 

9002 FORMAToc"OMAXIMUM ERROij GT o.oOlt =IIFI0.6,II, T~ACE = IIFI0. 4 • 0;3410 

1 "NLARGE = "16) 0;3420 

NElT GENERATE MATRIX ;XPONENTIAL SOLUTION 0;3430 

DO 260 I=I.ITOT 0;3440 

CSUMeI)=XTEMPCI) 0;3450 

CIMN(J)=XTEMPeI) 01 3460 

260 CONTINUE 0,3470 

EkR3=O~OOl*ERR 0;34~0 PO 3}0 NT=l,NLARGE 013496 

UO 270 I=},ITOT 0;3500 

CIMOCI)=CIMN(I) 013510 

270 CONTI~UE 013520 

TON=T/NT 013530 

NUL=} 013540 

DO 300 1=1,ITOT 013550 

IFC.NOT.LUNGeI» GO TU 300 013560 

NUM=NONPCI) 013570 

CIMNI=o.O 013580 

IFCNT.EQ.l) CIMNI=CIM~eI) 013590 

IF CNUL.GT .I~UM) GO TO ~90 013t'00 

DO 280 N=NUL,NuM 0,3610 

J=LOCPc N) 0;3620 

2AO CIMNI=CIMNI+APCN)oCIMVCJ) 0,3630 

290 CIMNI=CIMNI+DCI)OCIMOiI) oj3640 

CIMNI=TONoCIMNI 013~50 

IFcABSeCI~NI).LT.ERR31CI~NI=0.EO 0,3660 

CIMNel)=CIMNI 0,3670 

CSUMCI)=CSUM(I)+CIMNI 0}3680 

300 NUL=NONPeI)+} 0;3690 

3}0 CONTINUE 013700 

DO 320 1=1,ITOT U;3710 

IF(CSUMCI).LT.ERR) C~U~CI)=O.O 0,3720 

IFcLONGeI» XNEW(M,I)7C5UMCI) 0,3730 

320 CONTINuE 0}3740 

~ETURN 0}3750 

C 

C 

~ND 0;3760 

*DECK DECAy 0;3770 

SU~ROUTINE ~ECAY(M,T,~TOT) Oj37AO 

c 

c ~ECAY TPEATS SHORT_LIVED ISOTOPES AT ~EGI~NIN~ OF CHAINS USING 013790 

BATEMAN E~UATIONS 0;3800 

LOGICALLONG 0]3810 CO~MON/SERIESI XPc800l,XP~R(800) ,LONG(8001 013860 

CUMMON/CONST/MMN,ERR,MZE~U 

COMMON/EUj)f,TEMPe800, ,XNEwClO,dOO) ,R(800) ,l)c800) 

COHMON/~ATRIX/A(2500)ILOC(2500),NONOe800)'KD'~00) OI3 9 00 

LEVEL ?,A,Loc.NONO.KD 000560 COMMON/TERMP/DDCIOOt'~XP(100),QUEUEC50) ,N~u(~0),NQUEUEC50) ,NQ(800)013910 

AXN=-ALOGCo.iiOl) 

IJU 10 1=1,110T 

XPAR(I)=O.O 

LONGCIl=·fALSE. 

XPI=O.O 

IJT=DCI)~T 

IF(DT.LT.-50.) GO TO 10 

IF(A8SCOT) .LE.AXN) L~~GCII=.TRUE. 

)(PI=EXP(OT) 

3-41 

0;39?0 

0)3930 

0;3940 

013950 

0,3960 

Oj3970 

0;39~0 

0"13990 
C 

C 

20 

30 

40 

10 KPe

1 l=xPL 

NUL=1 

~O 160 l=lt ITUT 

XTEM=O'.O 

DL=-D ell 

NUM=N~NoeL) 

IFeM.~T.MMN.OR.M.EQ.MtERO) NUM=KDeL) 

IFeNUM~lT.NUL) GO TO ~50 

DO 140 N=NUL,NuM 

J=LOCeN) 

DJ:o-OeJ) 

IFeLoNGcJ» GO TO j40 USE THIS FORM FOR TWO NEA~LY EQUAL HALF-LIvES 

IFeABSeDL/DJ-l.0).LE.l.0E-S) XTEM=XTE~+XTEM~'J)~AeN)~XPeJ)~T 

IF,ABScDL/DJ-1.0).GT.~.OE~5) 

1 XTEM=XT~M+X!EMP(J)~AcN)~eXpeJ)~~peL»/eDL-DU) 

uUE=AIN)/UJ 

NQeL)=o 

NQeJ)=L 

N~AVE=o 

NUX=NONoeJ) 

IfCM.GT.MMN.OR.M.Ea.MtERO) NUX=KDCJ) 

NUf=1 

IFCJ.GT.l> NUf=NONoe'!-ll+1 

If,NUF.GT.NUX) GO TO 130 

uo 120 K=NUF,NUX 

Jl=LoCCKI 

IF eLONG cJl» GO TO 12Q 

KP=J 

IFeJl.EQ.N~'KP» GO TO 120 

I\P=NQ(KP) 

IfeKP.NE.O) GO TO 30 

UJ=-DeJ1) 

AKDJQ=ACK)/DJoQUE 

IFeAKPJD.LE.l.oE-06t GO To 120 

NUeJl)=J 

1=1 

KP=Jl 

{,)LJeI)=.oe KP ) 

DXpeI)=xPCKt') 

KP=NQ'KP) 

IFeKP.EQ.O) GO TO 50 

1=1+1 

IFeI.LE.I00) GO TO 40 

PRINT 9000, M.L.Jl'J,A~DJQ 

9000 fOR MA T 'c "1" ,4 I 5, E 12.5) 

50 

60 

GO To }30 

~ATE=O'.EO 

11=1-1 

XPL=XPcL) 

o R VONDY fORM OF BAT~MAN EQUATIONS -- ORNL·TM_361 

po 100 Kt3=ltIl 

XPJ=D.l(pe KB ) IFCXPL~XPJ.LT.ERR) GO TO 100 

tJl\=DDeKB) 

PROD= COL/UK-I. 0) 

LJKH=P~OD 

IFe A8sePROD).GT.l.E-~) GU TO 60 PROD=ToDKOXPJoel.0-0.5oeD~-DJ)oT) 

GU To 70 

PROD=(xPJ-XPL)/PROD 

PROl=XPJ/LJKt< 3-42 

0}4000 

0]4010 

014020 

0;4030 

014040 

0;4050 

014060 

Oi4070 

0;40M 

Oi4090 

014100 

0;4110 

014120 

0;4130 

0;41'+0 

0}4150 

0]4160 

0;4170 

0;41AO 

0;4190 

0;4200 

0,4210 

014220 

0,4230 

0;4240 

0]4250 

014260 

0,4270 

0;42AO 

0;42QO 

U,4300 

0;4310 

0;4320 

0;4330 

0;4340 

0;4350 

0;4360 

0,4370 

0,43RO 

0143QO 

0;4400 

0;4410 

014420 

014430 

014440 

014450 

014460 

014470 

0;44AO 

014490 

0;4500 

Oi

4510 

0]4520 

014530 

0,4540 

014550 

014560 

0;4570 

0145AO 

0;4590 

0)4600 

0146}0 
70 1->1::;1.0 

Sl=2./(DK~T) 

DO 90 JK=ltll 1FCJK.EQ.K~) GO TO 90 

S=I.0-DK/ODeJK) 

IfC AaSCS).GT.l.E-4t ~O TU 80 

C USE THIS FOHM FOR TWO N~A~LY EQUAL HALF-LIVES 

If(ABSCDKR).GT.l.0E-4, PBOD=PROI 

5=SI 

80 PI:;P1*S 

1F(AB::iCPIl.GT.l.E25t GO TO 100 

90 CONTIf'lUE 

~ATE=8ATE+PROD/PI 

100 CONTINuE 

IF(BATE.LT.O.EO)PRINTYOOl'L,I'BATE,XTEM,XTEMP(Jl),AKDJQ 

9001 FORMAT'(II Ld,BATE,XTEI1,XTEMPCJll ,AKDJQ ::; ",2I!J,1I"4EI2.5) 

1f(~ATE.LT.O.EO)AATE=O.EO 

XTEM=~TEM+XTEMPeJl)oA~D~Q~HATE 

IFCNSAvE.GE.50) GO TO 120 

110 NSAVE=NSAVE+l 

NQUEUEeNSAVE)=Jl 

QUEUE(NSAVE)=AKDJQ 

NQUCN~AVE)=J 

120 CONTINUE 

130 IFCNSAvE.LE.O) GO TO 140 

J=NQUEUE(NSAVE) 

jJUE=QUEUEet-/!:>AVE) 

NQCJ)=NQU(NSAVE) 

I~SAVE=NSAVE -1 

GO TO ;:>0 

140 CONTINUE 

IFCLONGeLI) XPAReLI=XTEM/~peL) 

150 NUL=NONO(L)+1 

IFe.NQT.LUN~eL» XNEW'M,L)=XTEM+XTEMPeL)OXP(L) 

160 CONTINuE 

PO 170 1=I.1TOT 

1F(LONG(1» XTEMPeIt=XTEM~eI).XPAReI) 

1Fe.NOT.LONGeI» XTEM~(I)ao.O 

170 CONTINuE 

HE TURN 

E;ND 

~DECK E;QUIL 

SUBROUTINE EQUILeM,ITUT) 

C 

C EQUIL PUTS SHORT-LIVEV DAUGHTERS IN EQUILIBRIuM WITH PAH[NTS 

C EQUIL USES GAUSS-SEI~EL iTERATION TO GEN~RAT~ STEADY STATE 

C CONCE~TRATIONS -

C 

LOGICALLONG 

COMMON/EQ/XTEMP(ROOt,XNE~(10,800)'B(800),~(800) 

COMMON/MATRIX/A(2500)'LOC~250D) ,NOND(8001'KD'~00) 

LEVEL 2,A,LuC,NONo,KD COMMON/CONST/MMN,ERR,MZ~RU 

COMMON/SERIES/ xr(8001,Xr~R(800) ,LONG(BOD) 

<.IXN=o.ilol 

DO 10 I=1.ITOT 

XPARel)=O.O 

1F(.NOT.LuNGeI» GO TU 10 

XTEMP(II=XTEMPeIloxP(AI 

XPAR(II=A~AX1(XNEW(M,1)-XTEMP(I) ,0.0) 

10 CONTII''lIiE 

IH.R=1 

20 1'4=0 

3-43 

Oi4620 

0,4630 

014640 

0,4650 

01 46 60 

0]4670 

0,46AO 

014690 

014700 

0}4710 

0;4720 

014730 

0]4740 

0;4750 

014760 

014170· 

014780 

0]4790 

0;4800 

0,4810 

0}4820 

014A10 

014A40 

014850 

014860 

014870 

014ABO 

014890 

014 9 00 

014910 

014920 

014930 

014940 

014950 

01491)0 

0,4970 

014980 

014990 

015000 

0,5010 

015020 

0,5030 

0;5040 

015050 

015060 

015070 

0}5080 

0]5090 

015100 

0]5140 

000580 

015160 

0,5170 

0;5180 

0,5190 

0;5200 

0]5210 

0,5220 

0;5230 

015240 
tjIG=o.il _ 1)0 60 1=I,1TOT 

NUM=NONO(I)-N 

DI=-D(J) 

IF(LO~G(I» GO TO 50 

)(NW=B(I) 

IF(M.GT.MMN.OR.M.EQ.MlERO) NUM=KD(I)-N IF(NUH~EQ.O) GO TO 3, 

DO 30 K=I,NUM 

t4=N+l 

J=LOC(N) 

DJ=-D(J) 

XJ=XP~R(J) 

IF(LONG(J» XJ=XJ+XTE~P(J)/(l.O-DJ/DI) 

~1'H,=X~w+A (N) *XJ 

30 CONTIf1IUE 

40 

50 

60 

70 

80 

31 XNW=XNW/DI 

IF(XNW~LT.l.OE-50) G~ TO 40 

ARG=AaS«XNw-XPAR(I.)!XNw) 

IF(AR~~GT.BIG) AIG=A~G 

XPAR(I)=XNW 

N=NONO(J) 

CONTINuE 

If (BIG".L T .l.IXN GOlD 10 

ITER=IrER+ 1 

IF(ITER.LT.I00) GO Tv 20 

PRINT 9000 

STOP 

lJO dO I=I,ITOT 

IF(.NOT.LUNG(I» XNEW,M,l)=XNlW(M,I)+XPAH(I) 

COiHINUE 

RETURN 

90{)0 fORMAT (II GAUSS SEIDEL ITE~ATION DID NOT CUNVE':<GE IN EQUILII) 

EN!:} 

ttDECK NUl.lATA 

0,5250 

Oj52F,0 

0,5270 

U152AO 

0,5290 

0,5300 

0,5310 

0,5320 

0,5330 

0,5340 

015350 

0,5360 

0,5370 

0;5380 

0,5390 

0,5400 

015410 

0,5420 

0,5430 

015440 

0;5450 

015460 

0)5470 

0,5480 

0;54QO 

015500 

0)5510 

0,5520 

015530 

0]5540 

OjS550 

0;5560 

0,5570 

0155AO 

0,5590 

01560 0 

0,5610 

0}51>20 

015630 

015640 

0]5650 

0,5610 

c 

C 

C 

C 

C 

SUBROUTINE NUDATA(NLI~E) 

NUPATA VERSION TO HANULE THREE TyPES OF NUCL;~R DATA LIBRARIES 

HAS POINTER, NLIBE, =-1 FUR HTGR 

= 2 FOR LIGHT WAT~R R~ACTvR 

= 3 FUR LMFBR 

= 4 FUR ~SBH 

INTEGERELE(99) ,STA(2) 

DIMENSION COEfF(7,A001,NP~OD(7,800) ,CAPT(b),Y~ELO(5,500) 

UIHENSION Y(5),NSORS(~),TYLO(5) ,NUCAL(6) 

DIMENSION SKIP(20),MS~S(20)'NAME(3) 

DIMENSION TUCAP(800t,tI~S<l00),TITLE(20).~(80P),fG(800), 

lALPHAI\I(lOO) ,SPONF(l00J,A8UND(500) ,KAP(800) ,MM~X(800) 

COMMON/LA~EL/ELE,STA 

COMMON/CONST/MMN,ERR,~ZERU 

COMMO~/EQ/XTEMP(AOO' ,ANEW.I10,BOO) ,B(800) ,1J(800) 

COMMON/FLUXN/REGENT,UAS(dPO),ILITE,IA~T,IrOT 

COMMO~/OUT/NUCL(AOOt 

COMMON/MATRIX/A(2500)ILOC(2500) ,NONO(800)'KD'~00) 

LEVEL 2,A,LOC,NONO.KD 

COHMON/CCOEFF/COEFF 

LEVEL ?CUEFF 

0]5750 

015850 

000600 

El.IUIVALEI>JCE (XNEW (1,4111) ,~PROP (1.1) 

El.IUIVALENCt (Al,DLAM) 01 5880 

DATA NUCAL/-20030.-10UOO,lO'11,-10.~91 0,58QO 

DATA MSRS/922330,922350,902320,922380,9423QO'~22330,922350.942410,0,5900 

1 9223AO.9423~0,942410,9223~0,942;00'~22380.942390,922330.0,5910 

? 922350.902320,922380,9423901 015920 

3-44 
C 

C 

C 

C 

C 

C 

C 

C 

C 

C 

C 

C 

9000 

10 

20 

30 

40 

SO 

60 

70 

80 

90 

100 

110 

120 

0}59)0 PROGR~M TO COMPUTE ~ ~ATRIX CTRANSITION MATRI~) FROM NUCLEAR DATA 015940 

0,5 Q 50 

015960 

0;5970 

015980 

0,59QO 

U16000 

0)6010 

016020 

0,6030 

0)6040 

0,6050 

0,6060 

0,6070 

01 6080 

016000 

0,6100 

0,6110 

0)6120 

0;6130 

0,6140 

0;6150 

0,6160 

0,6170 

0161An 

01f.lQO 

016200 

0,6210 

"'fAD 9011, (UTlE(I) ,1=1,18) ,''fLI8~ 

IF CNLlBE.LT .0) PROGRAM ~I~L READ TAP~ IN 'ASD~R FOHMAT 

IGWC=O 

IFCNLlRE.GT.O) 

IGWC=l 

NLIBE=':'NLlBE 

P.RINT 9000 

fORMATCIHO,"WILL 

Nl=4-t-lLl8E 

~EAD ~iiOl, 

PRINT 9005, 

PRINT 9006 

PRINT 9001 

PRINT 9008 

PRINT 9009 

pRINT 9010 

pkINT 9013 

PRINT Q014 

GO TO 10 

READ TAP6 GENERATED By CASD&~") 

THERM,R&S.FAST,ERR,NMO,NDAY,NYR,MPCTAB,INPT'IR 

~IMO,NI:?Ay,NVR 

THERM = RATIO OF THERMAL ~LUX TO TOTAL FLUX R~S = RATlu OF RESON~NCE FlUX TO TOT~L F~UX 

fAST = RATIU OF FAST ~LUX TO ToTAL FLUX 

ERR = TRU~CATION ERRO~ LI~IT 

READ DATA fOR LIGHT E~E~ENTS 

K=5~CNLI~l-l) 0162?0 

pO 30 Kl=l,~ 016210 

K2=K+Ki 0;6240 

N50RSCKl)=HSRSCK2) 01 6 250 

PRINT 901~' THERM.H~S,FAST,CNSORSCK),K=l,5)'NLIBE 016260 

1=0 0;6270 

NUTAPE=O 0162QO 

1=1+1 0'~290 

MEADCe,9034)NUCLCI),DhAM,lU,F81,FP,FP1,FT.FA,tsF,QCI),FGCl),ABUNO C016300 

11) ,DUMYl,UUMY2 

IFCEOf(8).NE.0)GOT02 b Q 

IFCIG~C.GT.O' GO TO lo 

DO 60 N=l,NLIBE 

READC6.9035) SIGTH,FN~I,FNA.FNP,RITH,FI~A.FIN~.SIGMEV,FN2Nl,FFNA. 

1 FFNP,IT 

GO To QO 

DO 80 N=l,NlIBE READC8,90 40) SIGTH,FN~I,FNA,FNP,RITH,FINA'fINr,SIGMEV,FN2N1,FFNA, 

1 FFNP,IT 

IFCN1.EQ.0) GO TO 110 

pO 100 N=l,Nl 

kEAD (6,9036) SKIP 

IFC1T.EQ.0) GO TO 50 

M=O 

CALL HALF C Al,lU) 

NUCL I =NIICl C I) 

IfCNuCLI.EW.O) GO TO 260 

CALL NOAHCNUCLI,NA~EI 

IfCMoDCI-l,501 .EY. 01 PHINT 9012, 

IFCMOUCI-l.50) .fU. O~ PRINT 9016 

SIGTH=TH[KM*SIGTH 

HITH=~ESIIRITH 

SlG~EV=FA~TIISIGMEV 

SIGNA=~IGTHIIFNA+RITHlltl~A'SIGMEVIIFpNA 

3-45 

,TIllE CN),N=1,181 

0)63;;>0 

0,6330 

016340 

0)6150 

0)6360 

016370 01 6380 

016390 

0)6400 

0,6410 

016420 

016430 

016440 

016450 

016460 

016470 

016480 

O,64QO 

016!100 

0111510 

0;6570 

0;6!130 

0,6540 

016550 
c 

C 

C 

C 

C 

C 

C 

C 

C 

C 

C 

c 

C 

C 

c 

C 

C 

c 

C 

C 

130 

140 

150 

160 

170 

SIGNP=SJGtrl~FNP+RITH~tI~P+SIGMEV~FFNP 

fNG=i.o-FNA-FNP -

If(FNY.LT.l.OE-41FNG=Y. 

F,ING=l~O-FINA-FINP 

IF(FING.LT.l.OE-4'FIN~=0. 

FN2N=1~0-FFNA-FFNP 

IF(FNZN.LT.l.0E.4)FN2N=O. 

SIGNG=SIGTH*FNG+RITH*~ING 

SIGN2N=SIGMEV*FN2N 

PRINT 9033, NAME. DLAM,F81,Fe'FPl,FT,~A.SlGNG, 

1 FNGl.SAGt-i2N,FN2Nl,SIGNA ,SII.1NP,~ (I) ,FG (I) ,ABtJND (I) 

TEST RADIOACTIVITy 

IF(Al.LE.ERR) GO TO lRO 

ABETA=i.o 

TEST POSITRON EMISSION 

IF(Fp ~LT. ERR) GO TO 150 

M=M+l 

COEFF(M,I)=FP~Al 

NPROD(M,I)=NUCLI-IOOOO 

~BETA=AAETA_FP 

TEST POSI1RUN EMISSIOI~ TO EXCITEO STATE Of PRUDUCT NUCLIDE 

IF(FPl .LT. ERR) GO T~ 150 

r1 =~1+1 

COEFF(M.I)=fPl*COEFF(~-I,~) 

NPROD(~.II=NPROD(M-l,l)+l 

COEFF(M_l. I I=COEFF(M-l,II-COEFF(M,I) 

TEST ISOMERIC TRANSIT~ON 

IF(FT ~LT.EHR) GO TO '60 

M=M+l 

COEFF(M,I)=FT*Al 

NPROD(M,I)=NUCLI 

AIjE TA= ABU A-FT 

TEST ~LPHA EMISSION 

IF(FA ~LT. ERR) GO TO 110 

M=M+l 

COEFF(~,I)=FA~Al 

NPROD(M,I)=NUCLI-200 40 

H=M+l 

COEFF(M,I)=COEFF(M-l'~) 

NPROD(~,I)=20040 

AbETA=A~ETA-FA 

TEST NEGATRON EMISSIO~ 

IF(ABETA.LT.l.E-4) GO TO lao 

M=H+l 

COEFF(~.II=ABETA*Al 

NPHOO(M,II=NUCLI+IOOOU 

T~ST ~EGATRJN EMISSIO~ TO ExcITED sTATE Of P~uOUCT NUCLIDE 

If (FBI .L T. ERR) GO TO 180 

11=M+l 3-46 

016560 

0]6570 

016~AO 

0165QO 

016600 

016610 

016620 

01 6630 

01'6640 

016650 

O;6MO 

016670 

Oj6680 

016MO 

01'6700 

01671() 

01'6720 

016730 

016740 

016750 

0;6760 

016170 

Ojb7AO 

o1'67QO 

016800 

016810 

016820 

0;6 A30 

016A40 

0)6850 

016 R60 

016870 

0,6880 

016890 

016 QOO 

0169}0 

016920 

l)16Q30 

0)6940 

016QSO 

0)6960 

016 Q70 

016980 

0169QO 

011000 

0)1010 

011020 

011030 

01'7040 

017050 

0]7060 

0,7070 

0;7080 

017090 

017100 

0;7110 

017120 

017130 

0;7140 

01'1150 

0]7160 

017170 
C 

C 

C 

180 

190 

200 

C 

C 

C 

C 

210 

220 

230 

240 

250 

260 

270 

280 

290 

300 

310 

320 

CO~FFCM,II;fBl~COEFfCM-l,l) 

NPRODCM,I,=NPRODCM-l,!)+1 

COEFfCM-l,I)=COEFF(M-i,II~COEFFCM.I) 

COMPUTE NEUTRON CAPTU~E C~OSS SECTIONS IN TH~~E REGIONS 

KAPCI)=M 

DO 190 KI=1.6 

CAPTCKI) =0.0 GAPT C1 ,=SIGNA 

CAPTCZ,=SIGNP 

CAPTC~)=SIbNG~fNGl 

CAPTe3,=SIGNG-CAPT(4) 

CAPTCQ)=Sl GN 2NoFN2Nl 

CAPTCS)=SIGN2N-CAPT16l 

TOCAp(I)=O.O 

TOTAL NEUTRON CROSS S~CTIUN FOR NUCLIDECI) 

uO 220 K=1,6 

CAPKBCAPT CK) 

IF(CAPKI.LT.ERR) GO!O 2~0 

1'1=1'1+1 

NPROD(M,l'=NUCLI+NUCA~(K) 

COEFF(M,I)=CAPKI 

TOCAPCI,=rO~APCII+CAP~I 

IfCK.I~E.ll GO TO 210 

1'1=1'1+1 

CUEFF(~,I,=COEFFCM_I'~) 

~PRODC~,I'=20040 

IF(K.NE.21 bO TO 220 

M=M+l 

CUEFF(M,I,=COEFF(M-l,AI 

NPKOD(~,I'=10010 

CONTJI'IIJE 

IFCMOPCNUCLI, IOI.EY,O) ~O TO 250 

1.)0 240 K=l,M 

NPRODCK,I)=NPRODCK,I)·1 

MMAX(I, =1'1 

IFcM.Gr.7) PRINT 903Y, ~ 

DIS(I)=Al 

~O TO 40 

ILITE = 1-1 

IACT=O 

READ DATA ON ACTINIDE~ 

OI71AO 

U,7]QO 

0;7200 

017210 

0;7220 

0;7230 

0;7240 

0;7250 

0;7260 

017270 

o 172M 

0;7290 

0;7300 

0;7310 

0;7320 

017330· 

0;7340 

017350 

0;B60 

0;7370 

0;7380 

0;7390 

0;7400 

Oj7410 

0;7420 

0;7430 

017440 

0;7450 

0]7460 

0] 74 70 

0,74HO 

0,7490 

017500 

0;7510 

0;75('0 

0;7530 

0]7540 

0,7550 

0;75':'0 

0'17570 

017580 

Oj7590 

0;7600 

0]7610 

0]7620 

READCi,9034)NUCLCI) .D~AM,IU,f81,fP,fPl,FT.FA'~SF.QCI),FGcJ) ,DUM~Y,0;7630 

IDUMy! ,DUMY2 IfCEO~CB).NE.0)GOT0450 Oi 7650 

DO 280 N=l,NUSE 0,7660 ~EADce.9037) SIGNG,RING.F~Gl,SIGf.RIF,SIGFF.S~GN2N,FN2NI,SIGN3~,ITO;7670 

CONTII'IUE Oi7680 

IFCNl.EQ.O) GO TO 300 017690 

00 290 N=I,Nl 0;7700 

I:(EADcS,9036) SKIP 0;7710 

IFeIT .Ew. 0) GO TO 2rO 0;7720 

M=O 017730 

NUCLI=NUCLCI) 0;7740 

IFCNUCLI.ty.O) GO TO 450 0;7750 

00 320 K=1,5 0;7760 

IFCNUCLI.EY.NSORSCKt) ~~URSC~I=I 0]7770 

CONTINUE 0117AO 

CALL ~ALFcAl,IU) 017790 

CALL NOAHCNUCLI,NA~E) 0;7~OO 

3-47 
330 

340 

C 

C 

C 

C 

C 

C 

350 

C 

360 

C 

370 

C 

C 

C 

380 

390 

400 

SIGNG=THE8M~SIG~G+~ES!RINi -0;7~10 

SIGf ~TH~HM~SIGF +~ES'RIF +FAST~SIGFF 0,7820 

SIGN2N=SIliN2NoFAST 0;7830 

SlGN3N=SIGN3N0FAST 0;7840 

IFCMOOCIACT,50).FQ.0) PRilNT 9012, cTIllE CN),N=ltlB) 017850 

IFCMOPCIACT,50).EQ.0) pRI~T 9024 017860 

PRINT 902&, NAME, DlAM,FB1,FP'FP1,FT,FA,FSF,SIGNb,0;7B70 

1 FNG1.S!GF,BIGN2N,SIGN3~,Q~~),F~CI) 0;7880 

IACT=IACT+l 0]7890 

0;7900 

017910 

oj 7920 

0;7930 

017940 

0;7950 

0,7960 

0;7970 

o 179M 

0;7990 

O;BOOO 

O;AO)O 

0;8020 

0;8030 

0;8040 

018050 

018060 

018070 

0;8080 

0;8090 

Oj8100 

0181)0 

018120 

0)8130 

018140 

0;8150 

0,8160 

0,8170 

018180 

0,8190 

0;8200 

Oj8210 

0]8220 

0;8230 

018240 

0,8250 

0,8260 

0;8270 

Oi8280 

0;8290 

0,8300 

0,8310 

0;8320 

0,8330 

0,8340 

018350 

0;8360 

0;8370 

o 183M 

018390 

0;8400 

0;8410 

0;8420 

TEST RADIOACTIVITY 

IFC Al.LT.ERR, GO TO i~o 

ABETA=l.O 

TEST POSITRON EMISSION 

IfCFp ~LT. tRR, GO TO""350 

ABETA=AAETA-FP 

1'1=1'1+1 

COEFFCM,I)=FPoAI 

NPRODCM,I,=NUCLI-IOOOO 

POSITRON EMISSION TO ~XCITED STATE 

IFCFPl .LT. ERR)GO TO 350 

M=M+l 

COEFFCM,I'=fPloCOEFFC~-l,l) 

NPROOCM,I,=NPROOCM-l,A,+1 

CUEFFCM-l,I'=COEFFCM-l,I'-COEFFCM,I' 

ISOMEHyC TRANSITION 

If CFT .LT.E~R'GO TO 3~0 

M=M+l 

COEFfCM,I)=FT*Al 

NPRODCM,I)=NUCLI 

AtiETA=ABETA-FT 

ALPHA EMISSION 

IFCFA .LT.ERR'GO TO 3~0 

M=~I+l 

COEFFC~,l)=FAoAI 

NPROOCM,I)=NUCLI-200·0 

ti=M +1 

COEFFCM,I)=COEFFCM-l'~' 

NPROOCM,I)=20040 

AtiETA=ARETA-FA 

t:;ETA DECAY 

IFCABETA.LT.l.E-4, GO TO 380 

M=M+l 

CO~fFCM,I'=ABETA~Al 

NPROOCM,I)=NUCLI+I000Q 

IFCFBI .LT. ERR,GO TO 3~0 

M=M+l 

eOEFFc~,I)=COEFF(M_lt~,oF~1 

CUEFFCM-l,l)=CoEFFcM-l,l)iCOEFF(M,I) 

NPRODCM,l,=NPRODCM-l,A,+1 

NEUTRON CAPTURE CROSS SECTIONS 

KAPCI)=M 

00 390 K=l,& 

CAPTcK ,=0.0 CAPTc,,=SlG~GoFNGI 

CAPT(1)=SIGNG-CAPTc2' 

CAPT(4)=SI GN 2NoF N2Nl CAPTc3,=SiG~2N~CAPTt4~ 

FIsscll\cn=SIGF 3-48 
c 

410 

420 

430 

440 

450 

460 

C 

C 

C 

G 

C 

C 

c 

470 

480 

490 

500 

510 

520 

531) 

54() 

550 

TO(;Ap (I) =(}. 0 

IJO I+IU K=lt4 

CAPKI:;CAPT(K) 

IF(CAPKI.LT.ERR) GO Io 410 

M=M+1 

TO'AP(I)=TOCAP(I)+CAP~I 

COEFF(M,I)=CAPKI 

NPROD(~,I)=NUCLI+NUCA~(K.i) 

CONTI~UE 

TOCAP(I)=TOCAP(I)+FIS~(lAtT) 

N-3N CROSS SECTION 

A17=SIGN3N 

IF(All.LT.ERR) GO TO ~·O 

M=M+l 

COEFF(M ,1)= A17 

NPROO(M ,1)= NUCLI~2~ 

TUCAP(I)=TOCAP(I)+A17 

IF(MOP(NUCLI,lo).EQ,Ol GO TO 440 

DO 430 K=l,M 

NPROD(K,I)=NPROD(K,I)-l 

I'1MAX(II=M 

IF(M.GT,7) PRINT 903~. M SPONF(IACT)=FSF~A1~6.U23Ec3 

~LPHA~(IACT)=FA~A1~6oQ2~E13~Q(I)~~3065 

DIS(I)=Al 

1=1+1 

<;0 TO ?70 

IL=O 

DO 460 K=1.5 

TYLD(KI=O.O 

~EAD DATA FOR FISSION pROOUCTS 

01 84 30 

0,8440 

0]8450 

0;8460 

0;8470 

0; 84M 

018490 

018500 

0;8510 

018520 

0;8530 

018540 

0;8550 

0185(,0 

018570 

0;85Ra 

0185QO 

018600 

018610 

OiS620 

0]8610 

1l;A640 

0,8f>50 

018660 

0;8670 

018ABO 

0]8690 

0]8700 

0;8710 

0187?0 

0,8730 

0;8740 

0187<;0 

READ (e ,9034) NUCL (1 ) ,D~A M , 1 U, F t; 1 ,F P ,F PI, F T ,F A,!:. SF, Q ( I ) ,FG ( I ) ,DUMMY, 0; 8 760 

llHJMY1,OU M Y2 

IF(EOF(8).NE.0)GOT0690 

DO 480 N=l,NLIBE ~EAD(e,9038) SIGNG,RI~G,FNGl,y,IT 

IF(N1.EQ.0) GO TO 500 

pO 490 N=l,Nl 

READ(8,9036) SKIP 

IF(IT ~EU. 0) GO TO 4~0 

M=O 

CALL HALF(A1,IU) 

l'lUCL I =NUCL (I) 

IF(NUCLI.EQ.O) GO TO ~90 

CALL NOAH(NUCLI,NAME) 

IF(MOP(IL,50).EQ. ot PRINT 9012, SIGNG~THERM~SIGNG+RES~RINw 

IF(NLIRE.EQ.3) GO TO ~ 0 

(TITLE (N),N=1,18) 

!F(MOP( IL ,50t.EQ.0) PRINT 9019 

PRINT 9021, NAME, DLAM,FB1,FP,FPl,FT,SIGNG, 

1 FNG1,Y,Q(Il ,FGfU 

~O TO 550 

IF (MOP( IL,50) .EQ.O) P~INT 9020 

PRINT 9~22, NAME 1 DLAM,F~1,FP'FPl,FT,SIGNG.FNG1, 

1 Y(2),Y'4),Y(5),Q(I),fG(I) 

T~ST ~ADIuACTIVITY 

If(AloLT.ERR) GO TO ~OO 

ABETA=l.O 

POSIT~ON l~lSSION 

3-49 

01.8780 

0;8790 

018~OO 

0;~f\10 

O,Sfl20 

018A30 

0;8840 

0}8f\50 

0;A860 

018870 

018A~0 

0;88QO 

0]8900 

0]8910 

0;8 Q20 

Oj8930 

0]8940 

0;8950 

0]8960 

0;8 Q 70 

0;8980 

0189QO 

01 9 000 

0;9010 

0,9020 

019030 

0;9040 

0;9050 
A3=Fp _ IF(A3.LT.ERR) GO TO ~70 

AI:iETA'lABETA-A3 

~Pl=AaoFPl 

AP=A3-API 

IF(AP.LT.ERR, GO TO ~60 

H=M+l 

COEFF(M,I,=APoAl 

NPROD(~,I'=NUCLI-IOOOO 

566 IF(APl~LT.EHR' GO TO 510 

I'1=M+l 

COEFF(M,I)=APloAl 

NPROD(~,I'=NUCLI-9999 

C ISOME~IC TRANSITION 

570 If(FT ~LT. tRR, GO TO seo 

M=M+l 

COEFF(M,I,=fToAl 

NPROO(M,I,=NUCLI 

ABET A=ABET A-FT 

C NEGATRON EMISSION 

c 

580 IF(AB~TA.LT.l.oE-41 GU TC 600 

A2=FBl 

AS 1 =Af::lETAoA2 AB=AB~TA-At:ll 

If(AB.LT.l.E-4) GO TO 590 

M=M+l 

COEFF(~,I)=ABoAl 

NPROD(M,I)=NUCLI+IOOOU 

590 IF(AHl.LT.l.E-6) GO TU 600 

H=M+l 

COEFF(M,I)=AAloAl 

NPRUD(M,I)=NUCLI+IOOO~ 

C NEUTRON CAPTURE CROSS SECTIONS FOR fISSION P~~DUCTS USING THREE 

C REGIO~ APPROXIMATION 

C 

600 KAP II) =M 

DO 6io K=1,6 

610 CAPT(K,=O.O 

CAPT(i)=SIGNGoFNGl 

CAPT(1,=SIGNG.CAPT(2) 

TOCAP(I'=O.O 

DO 620 K=1,2 

CAPKI=CAPT(K, 

IF(CAPKI.LT.ERR) GO To 6

i O 

M=M+l 

TUCAP(II=TOCAP~I'+CAP~I 

COEfF(M,I)=CAPKI 

NPROD(M,II=NUCLI+NUCA~(K+i' 

620 CONTII'IUE 

630 IF(MOP(NUCLI,lO).EQ.Ol GO TO 650 

PO 640 K=l,M 

640 NPROD(K,I)=NPROD(K,I)-l 

650 IL=IL+} 

DO 660 J=1,5 

YJ=Y(J)oO.OlO 

TYLO(J)=TYLU(J)+YJ 

660 yl[LO(J,IL)=YJ 

If(NLIBE.EU.l.OR.NLI8f.EG.4) GO TO 680 

670 If (NURE.EY.3) YIELDtl'Il.)=yJ 

ylELD(3,IL)=YJ 

680 i'1~AX(l,=M 3-50 

Oj9060 

0,9070 

O,90M 

0,9090 

0;9100 

0,9110 

0;9120 

0;9130 

0;9140 

0;9150 

0;9160 - / 0,9170 

0;91AO 

0]9190 

01920Q 

0;9210 

0;9t'20 

0;9230 

0,9240 

0;92~0 

0,9260 

O,q?10 

O;q?~O 

0;Q2QO 

01930/) 

0;9310 

019320 

0;9330 

0,9340 

0;9350 

0;93"0 

0'~310 

0193AO 

019390 

01

Q400 

0;9410 

0,9420 

0;9430 

0;9440 

0;9450 

0;9460 

0,9470 

O;94AO 

0;9490 

O;9~00 

0;9510 

0,9520 

0,9530 

0;9540 

0;9550 

O;95f1O 

0;9~10 

O,]9C;AO 

0;9590 

0;9"00 

0,96)0 

0;9620 

019630 

0;9"40 

019650 

019MO 

0;9670 
690 

C 

C 

C 

c 

C 

C 

C 

C 

C 

706 

710 

720 

730 

740 

750 

760 

IF eM.(;iT. 7>- PRINT 903't, M 

DlS(l)=Al 

1=1+1 

GO TO 470 

lFP=IL 

ALL DATA ON NUCLIDES ~AS ~EEN READ. BEGIN TO ~OHPUTE MATRIx COEFF 

lTOT=I_l 

fIND PRODUCT NUCLIDES FOR REACTIONS OF LI~HT ~LEHENTS 

NON=O 

DO 700 1<=I91TOT 

NONOeK)=O 

IFeILITE.LT.l) GO TO 760 

00 750 I=l,ILITE 

NUCLI=NUCLeI) 

UO 720 J=l,ILITE 

KMAX=KAPeJ) 

IFeKMAx.LTel) GO TO 120 

DO 710 M=I,KMAX IFeNUCLI.NE.NPRODe~,Jl) GU TO 710 

NONOel)=NONoel)+l 

NON=NON+l 

IfeNO~~GT.2500) PRIN! 90~1' NON,NUCLeI) 

AeNON)=cOEFFeM,J) 

JT=J 

LOceNON)=JT 

CUNTINuE 

CONTINUE 

KUeI)=NONOeI) 

UO 740 J=I,ILITE 

Kl=KAP eJ)+1 

I\MAX=MMAxeJ) 

IF eKMAx.LT.KlI r,o TO 740 

DO 730 M=Kl,KMAX 

IFeNuCLI.NE.NPRODeM,JA) GU TO 730 

NONoeI)= NUN OeI)+l 

NON=NON+l 

IFeNO~~GT.2500) PRIN! 90~1, NON,NUCL(I) 

AeNON)=COEFFeM,J) 

JT:;J 

LoceNON)=JT 

CONTINUE 

CONTINUE 

CONTINuE 

NON ZERO MATRIX ELEMENTS tOR THE ACTINIDE~ 

IFeIACT.LT.l) GO TO ~20 

IU=ILlTE+l 

Il=ILlTE+lACT 

DO 810 1=10,11 

NUCLI=NuCLeI) 

L>O 780 J=Io,1l 

MAx=Kt'peJ) 

IF eMAI<.LT.l) GO TO 7~0 

DO 770 M=l,MAX 

IFeNUCLI.NE.NPRODe~,Jl) ~O TO 770 

NONOeI)=NONOeI)+l 

NUN=NPN+l 

IFeNON·.GT.2~00) PRII~! 9u~1, NON,NUCLeI) 

3-51 

oIQMO 

0,9690 

019700 

0,9710 

0,9720 

019730 

0,9740 

019750 

0]9760 

o i 9170 

0;9780 

0;9790 

019800 

0;9810 

0;9820 

0]9830· 

0]9840 

0;9850 

019860 

0;9870 

o 198M 

01Q~qO 

019900 

019910 

019 97.0 

019930 

019940 

0,9950 

0199('0 

0;9970 

0,99AO 

019Q90 

O?OOOO 

0?0010 

O?OO<,O 

0?0030 

0?0040 

0?0050 

01.>0060 

020070 

O?OORO 

o;:'OOQO 

0?0100 

01.>0110 

0?0120 

020130 

070140 

0,,0150 

01.>0160 

0;:'0170 

0?0180 

O<,OIQO 

0,,0200 

020210 

070220 

0;:'0230 

020240 

0?0250 

0?0260 

0,,0270 

0?0280 

0?0290 

020300 
C 

C 

C 

710 

780 

790 

800 

810 

820 

830 

840 

850 

860 

AINON)=coEffeM,J) 

JT=J 

LOC(NON)=JT 

CONTI~UE 

CONTINuE 

KIJ(I)=NONO(l) 

pO 800 J=IOtIl 

Ml::;KAlf,J) .1 

M2:;:MMAxeJ) 

IF(MZ.LT.Ml) GO TO 890 

pO 790 M=Ml,M2 

IFeNuCLI.NE.NPROD(M,Jl) GU TO 790 

NONO(I)=NONO(I)+l 

NON=NON+l 

IFeNO~~GT.2500) PRIN! 90&1, NON,NUCL(I) 

A(NON)=COEFF(H,J) 

JT=J 

LOC(NON)=JT 

CONTINuE 

CONTINUE 

CONTINUE 

MATRIX ELtMENTS FOR ~lsSIUN PRODUCTS 

IF(IFP~LT.l) RETURN 

IM=IUTE+IACT 

IO=IM+l 

IF(ITOT.LT.IO) RF.TuRN 

DO 880 I=IO,ITOT 

NUCLI=NIJCL(I) 

I2=MAxijelO,1-10) 

I3=MINoeI TU T,I+10) 

DO tl40 J=I2,r3 

KMAX=I<AP(J) 

IF(KMAx.LT.l) GO TO ~40 

DO 830 H=l,I\MAX 

IF(NUCLI.NE.NPROD(M,Jl) G~ TO 830 

NONO(I)=NONO(I)+l 

NON=NON+l 

IF(NO~~GT.2500) PRINI 90~1, NON,NuCL(I) 

A(NON)=CU[FfeH,J) 

JT=J 

LOC(NON)=JT 

CONTINUE 

CONTINuE 

KDeI)=NONO(I) 

pO 860 J=I2t13 

K I =KAP 'e J) +1 

KMAX:/",HAXeJ) 

IF(KMAx.LT.Kl) GO TO '60 

00 850 M=Kl,KMAX 

IF(NUClI.NE.NPROD(M,Jl) GU TO 850 

MONO(I)=NONO(l)+l 

NON=NON+l 

IF(NON.GT.2500) PRINI ~O&I, NON,NUCL(l) 

AeNON)=cOEFFeM,J) 

JT=J 

LOC(NOt.l)=JT 

CONTINUE 

CONTII'IUE 

IfeIACT.lT.I) GO TO 680 

uo 870 K=l,5 3-52 

_ 02031 0 

01'0320 

020330 

020340 

020350 

0?03,,0 

0;:'0370 

0?0380 

0;'0390 

020400 

0?0410 

020420 

020430 

01'0440 

0;'0450 

01'0460 

0,.,0470 

070480 

0204QO 

01'0500 

O?05}0 

01'0520 

u?0530 

01'0540 

0?05QO 

01'0600 

01'0610 

020620 

0;,0f,30 

O?Of,40 

020650 

01'0660 

01'0670 

0?06BO 

O?06QO 

020700 

0207}0 

020720 

020730 

01'0740 

01'0750 

020760 

070170 

O?07aO 

0207QO 

0;;00800 

0?08)0 

O?0820 

0?OA30 

0?0840 

()?OB50 

0;'0860 

070870 

O?08AQ 

O?OAQO 

01'0900 

O?09}0 

070920 
870 

880 

IL=I-IM 

If(YIElD(K,IL).LT.ERRl GU TO 870 

NON=NON+l 

IF(NOb.GT.2500) PRIN! 90&1, ~ONO(I)=NONO(I)+1 

KK:;NSDRS(K) 

LOC(NON)=KK 

KF=KK"rLITE 

A(NON)=YIELD(K,IL)*PI~S'K~) 

CONTINUE 

CONTINUE 

IF(IFP~lE.O) GO TO 90Q 

If(NLIBE.NE.3) GO TO ~90 

NON,NUCL(I) 

PHINT 9027, TYLDt21,TyLD(4),TyLDIS) 

890 

C 

60 TO 900 

pRINT 9030, 

0;»0930 

020940 

0?0950 

020960 

0?0970 

0?09AO 

0;:'0990 

0;:'1000 

0;)1010 

0;)1020 

021030 

0?1040 

0;)1050 

O;»lOM 

021070 

O? 1080-

0;)1090 

0;)1100 

0?1]10 

0;»1]20 

C 

C 

C 

C 

C 

C 

ALL M~TRIA ELEMENTS A~E ~OW COMPUTED 

HEGIN TRANSIENT SOLUTAON 

TEMPORARILY wRITE OUT MAT~IX ELEMENTS 

0;»]130 

021140 

0;»1150 

900 If(IR.ElJ. 0) RETURN 0?1160 

PRINT 9029 021170 

N=O 0211AO 

UO 910 I=ldTOT O?11QO 

NUM=NONO(l) 0?1?00 If(NU~.LE.O) GO TO 910 0;»12]0 

NI=N+t'-ilfM 0?12?0 

N=N+l 0?}?30 

PRINT Q028, I,UIS U),TOCAP(I),(A(K),LUC(r(.),K=N,Nll 0;»1240 

N=Nl 0?1250 

910 CONTINUE 0?1?60 RETUR~ 0;»1210 

920 STOP O?I?AO 

C 0?12QO 

C fOHMATS FORMATS FURMATS FORMATS 0;»1300 

C 0213}0 

9001 FOHMAT(4FIO.S,612) 071320 

9005 fORMATt1Hl,43X,IINUCLE~R THANSMUTATION DATI' '!iEVISED II,I2',II/"tl2,"O?1370 1I",I2,/t"Of~UCL = NUGL,~DE T' 10000 * ATOMIC NO + 10 * MASS NO + ISO~0?13AO 

2 f:: RIC S TAT E (0 0 R 1)", 1 0 x , "D LAM = DEC Aye 0 I~ S T A ~ T (} / SEC,. II ~ "" F B, 0 ? 1 3 9 0 

3FP, F~, FT = FRACTION~L O~CAy BY BETA, PQ5ITRVN (OR ELECTRON CAPTUO?1400 

4RE), ALPI1A, INTERNAL TRANSITION. FB = 1 .. F~ .. FA .. FT"~/t" fBl,021410 

5 fPl, FNGl, FN2Nl = fijACTION UF BETA, POSITRO~, N .. GAMHA, N-2N TR4N021420 

6SITIONs TO EXCITED ST~TE OF PRODUCT NUCLI~E".I,II SIGTH, SlGNG, SIGO?1430 

7f, SIGNA, SIGNP = THE8MAL CHoSS SECTI~NS (BAR~S) FOR ABSORPTION, NO?1440 

A-GAMMA, FISSION, N-ALeH~, N-PROTON.") O?1450 

9006 fORHAT(1I SIGNG = SI~TI1 ~ (} .. fNA -FNPle Sl,GNA = SIGTH * FNA. 0?14(,Q 

ISIGNP = SIGTH * FNP. F~Ai FNP = FRACTION THE~MAL N-ALPHA, N-PROT0021410 

2N,",/t" RlTH, RING, Rl.F, IHNA, RINp = RESUNAN~E INTEGRAL FOR ABSLlRO?14~0 

3P1ION, N-GAHMA, FISSIPN. N-ALPHA, N-pHOTON.",I," RING = RITH * (021490 

41 - FINA - fINP). RI N8 = HITH * fINA. RINP = KITH * FINP. FINA, fO?1500 

5 1 r~ P = F RAe T ION RES 0 N A r~ C E N - ALP H A, N - pRO TON. " , I ," 5 I G "1 E V, S I G f f, S I 0 ? 1 5 1 0 ~G~2N, SIGNAf, SIGNPF j FAST CROSS SECTIONS (a~RNS) FOR ABSORPTIO~,0?1<;20 

7 f ISS ION, l\j - 2 N, N - ALP H A, r~ ':' P k 0 TON. " , I , " ~ I G N i N = S I G MEV .. (1 - F F 0 ? 1 S 3 0 

ANA - fFNP). SIGNAF =-SIGMEV 0 FfNA. Slij~pF; SIGMEV 0 FFNP. FF NO?1540 

9A, Ff~P = FRACTION FAST r-.7ALpHA, N_P.") 0;'1550 

9 0 0 1 FUR "1 A T (" Y 2 3, Y 2 5, Y 0 G, Y 28, Y 4 9 = f I 55 I 01, Y I ~ L 0 (P E ~ C E NT) FRO "1 23021 c; 6 0 

13-U, 23t;-U, 232-TH, 238-U, 239- pU.",/," \>I = flEAT pER OlSINTEGRATIO?1570 

2 UN. F G = FHA C TI 0 N 0 F H l A TIN G A ~I HAS U F E I ; ERG l' G R E 4 T E R T HAN o.? ,'I EO? 1 <; A 0 

3V.",/, .. O t.ftECTIvE Cr<USS ~ECTIO"'S fUR A VULUM= AVERAGED THERMAL (LO?lSQO 

3-53 
4T 0.876 E'O FLUX ARE ~S FULLOWS.",/'." r-I_Gf'tlMA SIGNG .. THEK MO?,1 6 00 

5 • RING" RES.",/," FIt:'SIoN SIGF • THEt:CM + RIF .. R[S • SIGFO?1610 

6F 0 F(lST."tlOX,"THERM = I/V cORRECTION Fall< T~~RMAL SPECTRUM AND TEO?1620 

7MPERATuRE.",/," N-4N ~IGN2N 0 FAST,II,36X,"RES = RATIU 07.1630 

aOF RESONANCE FLUX pER LETt1ARGY UNIT TO TI1~RMA~ FLUX.") O?1640 

9008 FORMAT(" N-ALPHA SiGNA 0 THERH • RINA 0 RES + SIGNAF 0 FASTO?1650 

1.",1x."FAST = 1.45 it RAHo OF FAST (GT 1.0 "'~V) TO THERMAL FLUX "0~16"0 

2/" N_PROTON _ SIG!1p. THERH + RINP 0 'f,ES • SIGNPf 0 F~ST .") 021670 

9009 FORMAT'(1HO.59X,"REFERENCEb",It" HALF l.~VES" DECAY SCHEMES, AND 0216A0 

1 THE'RM(lL POWER",/''' C " LE9ERER, J M H~LLA~DEB, AND I PERLMAN ""TABO?16QO 

2LE OF ISOTOPES - SI)(T~ EDiTION"" JoHN WIL~,Y AI~D SONS, INC (1967)",0;'1100 

31t" B s DZHElEPOV AND L" PEKER ""DECAy SC:HEM~S OF RADIOACTIVE NIICO?1110 

4LEI"" PERGAMMON PRESS (1981)"'/''' D T GOl~IMAN AND JAMES R ROSSER "0?1720 

5"CHART of THE NUCLIDE:.p"" NINTH EDITION GENE~~L ELECTRIC Co (JUlY 0?1730 

61966)",It" E 0 ARNOLD "IIP~OGRAM SPECTHA"" APPtNDIX A OF OHNL-3516 0?1740 

7(APRIL 1964)") 0?1750 

9010 fORMAT'(" CROSS SECTIONS AND FLUx SPECT~:A"'/," B E PRINCE ""NEUTO?1760 

IRON REACTION RATES IN TI1E MSRE SPECTRUM"" ORN~-4119, PP 79-83 (JlJLO?I170 

2Y 19(7)",/," B E PRIN~E ""NEuTRON ENERGY SPECTRA IN MSRE AND MSBR"O?1780 

3" ORNL_4191, PP 50-58 (DE£; 1967)",1t" M 0 GOLI:!BERG ET AL 1'''NEUTRUNO~17QO 

4 CROSS SECTIONS"" BNL-325. SECOND ED, supP NO 2 (MAY 1964 - AUG 19021AOO 

566) ALsO EARLIER EDITIONS",/,'I H T KERR, UNPUblISHED ERC COMPILATI021RI0 

~UN (FEB 19o{:I)",/," M ;\ DRt\KE ""A COMPILATION uF RESONANCE II'HEGRALO?lfl20 

7S"" NUClEUl'.IlCS, VOL 24, ~u 8, PP 108-111 (AUS 1966)"'/," (jNWL STAFO?1A30 

AF IIIII~vESTI(;ATION OF ~~_2r-. CROSS SECTIONSltll BN~C-98, PP 44-98 (JU-~E021840 

9 1965)11) O;?1850 

9011 FORMAT (l8A4d3) O?lA60 

9012 FORMAT(}Hl,20X,18A4t 021A70 

9013 FORMAT(II H ALTER AND ~ ~ WEBEH ""PRODUCTION V~ H AND HE IN METALS 0~IA80 lUUR ING REACTOR IRRADI~TlONIIII J NUCl HATLS, VOL 16, PP 68- 73 (]965)0?1 8 90 

2",/t" L L tlENNETT ,,"R~COMMENDt.D FISSI~N PI"<ODU':T CHAINS FOR USE I;~ 0?1900 

3REACTOR EVALUATION ST~DIE61111 ORNL-TM-16se (SE~T }9(6)1I) 021 Q }0 

9014 FORMAT(" FISSION PkODUCT ylELDSII,/t" M E ~~EK ANO B F FqDEH, 1I110~1920 

ISU~MARy of FISSION PR~DUCT YIELDS FOR U-235, U-238, PU-239, AND PU021930 

2-241 AT THERMAL, FIssION 5PECTRUM AND"/" 14 MEV NEUTRON ENERGIOi'}940 

3ESIIII ~PED-5398-A(REV.t,(O~T. 19(8)11/11 S KI"TCO~F 1111 FISSION PRODIICT021950 

4ylELDS FRUM NEUTRON I~DUC6D FISSION"" NUC~EONl,CS, VOL 18, NO 11, 01'1960 

S(NOV 1960)"1" N 0 DuDE-Y 1".1 REVIEW of LOW-:I1ASS ATOM PRODUCTION IN FO?lQ70 

6AST REACTOHSIIII ANl_74~4,(aPRI~ 1968) II) 0;'19~0 

9016 FORMAT(lHO,20X,"LIGHT ELEMENTS, MATERIALS Of ~ONSTRUCTION' AND ACTO?2030 

}IVATTION PRODUCTS "'/,"0 NUCL Ul~M Ft?l FP " 000640 

2"FP1 FT FA SlGNG FNGI SIbN2i11 FN2Nl" 000650 

3" SIGNA SIGNP 'Iil FG ABUNDANCE") 000660 

9018 fORMAT (lHOtlOX,IITHERMj IIFIO.5'5X,IIRES~ IIF10.S15X,IIFAST= IIflO,5, 0220QO 

11/,1 X, IINEUTHON SOURcE:; 115 (110' 5X) ,5X, "NLlpE= II I 3) 022100 

9019 FORMAT(lH0t36X,IIFISsI~N P~ODUCTSII,/t"O NU~l DLAM JJ OOOf,QO 

1 IIFBI Fp FPl FT SIGNG FNGl '1'23 II 000700 

2 IIY2!) YOl Y2~ Y49 Q FGII) 000710 

9020 fORMAT(lHO,36X,IIFISSI~N P~ODUCTS",/tIlO NUCl DLAM fBI II 000720 

} IIFP FPI FT SIGNG FNGI y?~ Y28" 000730 

2 II Y49 Q FGII) 000740 

9021 FORMATJ1H ,A2,I3,Al,l~ElO.2,OP4F7.3,IPEI0.2,OtF7.3, 060 750 

1 IP5i;.10.2,OP2F7.3) 000760 

902? fORMAT(lH ,A2,I3,Al,1~EI0.2,oP4F7.3,1~EI0.2, 000770 

1 OPF7~3,lP3E10.2,oP2t7.3) 0~0780 9024 FOI-IMAT'(lHO,32X, IIAcTlfiIUE:> ANU THEIR DAUGt1TER~",I/ 022no 

1" NUCL ' DLAM fB1 FP FPl FT" OOOAOO 

2" fA FSF E+6 -SlGNG FrJG21 SxGF" 000810 

3" SIG N2N SIGN~N Q FGII) 000820 

9026 FORMAT(lH '~2,I3,Al,1~ElO.2,oP5F7.3,6PF10.1,1~ElO.2, 0008]0 

1 OPF7~3,}P3EI0.2,OPF~.3,f6.2) 000A40 

9027 FUi'H1AT(lIoSUM OF YIELDS OF ALL FISSION PROIJUCT? ::1I,J5X,lf"3£Q.2) 0222QO 

3-54 
9028 fUHMAT(I5,2X,lPEIO.3,JX,EIO.3,5(2X,EIO.3,3x,I~)/(30X,5(2X,[10.3, 02'2300 

1 3x,15») 0223]0 

9029 FORMAT(IIl:\jO~-ZE~O MI\Tr<IX t.LEMlNTS AND THEIR LUCATIONS"I 

1" I 015(1) . CAt(l) ACI,J) J A(l,,,",) 

022320 

022330 

2J A C I , J) J A (1, J) J A ( I , J) J til 022340 

9030 FORMATc63HOSUM OF YIE~DS UF ALL FISSION P~ODU~TS 

1 ,SCE9.2,IX» 

9033 FORMAICIH 'A2,I3,Al,I~ElO.2,oPSF7.3'lPEIO.2,0~F7.3,lPElO.i, 000860 

On0870 

022420 

022430 

0;;2440 

000890 

022460 

022470 

022480-

0?2490 

022500 

022510 

1 OPFI~3,lP2EIO.2,oP2t1.3~F8.J) 

9034 fORMATCI 7 ,f9.3,I},5PS,3.1RE9.2,OP2FS.3,F7.3,2E6.0) 

9035 fO~MATC7X,f9.2,3F5.3,F.9.2~2F5.3,f9.2,3F5.J, ~~.Il) 

9036 FORMAT(20 A4 ) 

9037 fORMAT(7X,2F9.2,F5.3,'F9.~,F4.I,F9.2,Il) 

9038 FUHMAT(7X,2F9.2,F5.~,3F9.2, 4X,Il) 

9039 FORMAT(IIO \ljARNING. M~UT (,IF R~NGE IN NUDATA. =11 IS) 

9040 fORMAT( 7x,f9.2,3F8,6,F&.2,2F3.1,f9.2,afS.3'5X,Il) 

9 041 fOH~lA T( II 0 NON HAS E)(C~EDE~ 2500, EQUAL TO 112 I~n 

EI'W 

"DECK COLLECT 

10 

20 

~UBRoUTINE COLLECT(TM~,C\ljASTE,ITOT) 

C0I'1MON/El.I/XTEMP(8001 ,XNE\ljOO,BOO) ,B(800) ,~(8g0) 

UI~ENSI0N C~ASTE(800) 

IFCTHB~LT.l) RETURN 

DO 10 1=I,1TOT 

til 1) =CwASTE (I) XTE~1P(II=0.0 

CALL O~CAYC1,TMB,ITOTl 

CALL TERM(TMB,I,JTOT) 

CALL t::QUIL(l,ITOTI 

uU 20 1=1,110T CWAST~(II=XNEW(I,II/T~B 

RETURN 

ENU 

0?2550 

022c;~0 

022570 

0225RO 

022590 

022600 

"UECK STORAG 

022620 

0?2630 

0('2f>40 

022650 

022660 

022670 

10 

20 

SUBROUTINt STORAGCTM~lcwAoTE,lTOT) 

COW-10N IE'QI X TEMP P~O 0 I , XNEw ( 10, tjo 0) ,13 (800) ,j.) (800) 

IJIMENSrON CWASTE(ITOT~ 

IF(TM8~LT.1) RETURN 

DELT=TMB 

1)0 10 I=I.ITOT 

1;1(1)=0.0 

XTEMPCI)=CwASTECI) 

CALL DECAY(1,DELT,ITO!) 

CALL TER~cTMB,l,ITOT) 

0227}0 

022720 

O?2730 

O?2740 

072750 

0?2760 

022170 

CALL EQUILCl.ITOT) 0227QO 

DO 20 1=1,ITOT 022POO 

CWASTEcI)=XNEWCl,l) 022810 RETUR~ 022820 

END 022830 

*DECK ~LKDATl 022A40 

C PKOGRi\M BLOCK DATA 022850 ~LOCK DATA ~LKDATI 0?2860 INTEGtRELEC~9),STA(2) 0?2870 COMMO~/LA~ELI ELE,ST~ 022880 

IJATA ELE/" t-1II,IIHEII.IIL,l.",lIbEII,1I BII,II C"," 1'4"," 0",11 F",IINE","NAII, IIM0 22890 

}G"."AL",IISI"," P"," S","CL",,,AR"," KII,IICA"."S~II,"TI",,, VII ',IICRII, IIMN 022900 

2","FE"."COII,"NI"."CU"'ll~"'''GA",''GE'''''AS",''SE"'''HR",1IKR","R8 11 ,IISR"O?2910 3 ," Y II , "Z K II , II N B II , "1>10" , "T ~" , "R u" , II R H II, II POll, "1\ Gil', II C 0" , II IN" , II bN II , II S B" , o? 29 20 

4"TEII, ,_, I II, "XE", "eSII, "eA", "LA", "CE", "pRII, "I'iD" , "pM", "5"'", liEU", "GD". "022930 

5TH","Py"," t1 0",IIERII,IITI'1",IIYt:3 l1 ,"LU",IIHF",IIT(l.II,II W","REII,"U5","lR","P022940 

6T","AU",IIH~"'''TL","p~II,''t:!1","PO",''AT","RN".''F~",IIRAII."ACII,IITtl"."PA022950 

7"," U","NP","PU","AMII,"CM",IIR,,",IICFII,IIESII/ 0229£,0 

OATA STA/" 11,11'" "I 022970 

3-55 
END U229AO ~DECK HALF O?2990 

SUBROUTINE HALF(A,I) 073000 

C SUBROVTINE HALF CONVE~TS ~ALF-LlfE TO DECAy CUNSTANT (l/StC) 0?3010 

DIMENSION C(9) O?3020 

DATA C/6.9315E_Ol,I.1~S2E-02,l.9254E-04,8.022~E-06,2.19b5E-08,0.0,0?3030 

1 2.1~b5E~11,2.1Yb5E~14,2.1965E-171 0?3040 

IF(A.~T.O.O) GO TO lQ 023050 

IF(l.Ea.b) GO TO 20 0230~0 

A=9.99 023070 

~ETUR~ 0?30AO 

10 A=C(I)/A O?3090 

RETURN 0?3100 

20 A=O.O 0?3110 

RETUR~ 0?312Q 

END 023130 

~DECK NOAH 0?3140 

SUHROUTINE NOAH(NUCLI1NAM~) 073150 

C SU~ROUTINE NOAH CONVE~TS bIX DIGIT IDENTltIE~ TO ALPHAMERIC SYM80L073160 

INTEG~RNAME(3) O?3170 

INTEGERELE(99),STA(2) 0231AO 

COMMON/LAHELI ELE.ST~ O?31qO 

IS=MoD(NUCLI,lO)+} 023200 

NZ =~UCLI/IOO~O 073210 

MW=NUCLI/IO-NZ ~lOOO O?3220 

NAME(l)=ELE(NZ) 023230 

NAME(2)=Hw O?3240 

NAME(3)=STA(IS) 0?3250 

RtTURN 0?3260 

END 073270 

3-56 
CHAPTER 4. DATA FOR RADIOACTIVE 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 deminera1izers. 

Note: The letdown rate should include the fraction of time 

the cation demineralizers are in service. 

4. The average shim bleed flow rate (gal/min). 

4.3 SECONDARY SYSTEM 

1. The number and type of steam generators and the carryover 

factor used in the evaluation for iodine and nonvo1ati1es. 

2. The total steam flow rate (lb/hr) in the secondary system. 

3. The mass of liquid in each steam generator (lb) at full power. 

4. The primary-to-secondary system leakage rate (lb/day) used 

in the evaluation. 

4-1 
5. Description of the steam generator blowdown purification system. 

The average steam generator blowdown rate (lb/hr) used in the 

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. 

b. 

c. 

d. 

e. 

f. 

Average flow rate (lb/hr); 

Demineralizer type (deep bed or powdered resin); 

Number and size (ft3) of demineralizers; 

Regeneration frequency; 

Indication whether ultrasonic resin cleaning is used and 

the waste liquid volume associated with its use; and 

Regenerant volume (gal/event) and activity. 

4.4 LIQUID WASTE PROCESSING SYSTEMS 

1. For each liquid waste processing system, including the shim 

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 

(fraction 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 

discharged over the life of the plant. 

f. For demineralizer regeneration, provide time between rege~erations, regenerant volumes and activities, 

treatment of regenerants, and fraction of regenerant 

discharged. Include parameters used in making these 

determinations. 

g. Liquid source term by radionuclide in Ci/yr for normal 

operation, including anticipated operational occurrences. 

4-2 
2. Provide plplng 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 (ft3/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 (ft3), 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 radionuc1ide 

considered in the evaluation. List3

a1l parameters, including 

mass of charcoal (lb), 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 b1owdown 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 
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 tr.e 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 component 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-l 
TABLE A-l 

Demineralizer Service 

Primary Coolant Letdown 

Boron Recovery System 

Equipment Drain Wastes 

Floor Drain Wastes 

Steam Generator Blowdown 

Regeneration Frequency 

180 days 

180 days 

* 

* 

90 days 

* Regeneration frequency is calculated by dividing the waste quantity 

(gallons) by the waste flow rate in gallons per day. The waste 

quantity is 25000 gal/ft3 times the volume in ft3 of resin for 

equipment drain waste and 2000 gal/ft3 times the volume in ft 3 of 

resin for floor drain waste. The calculated values of 25,000 and 

2,000 gal/ft3 of resin for the waste are based on 12,000 g 

CaC03 ion exchange capacity per ft3 of resin and 5 ~mho/cm and 

50 ~mho/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 i~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 
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: 

Ca rd 12 

Spaces 18-41 enter - shim bleed demin regen 

Spaces 42-49 enter - 1440.0 

Ca rd 13 

The wastes will be processed through an evaporator which will 

provide the following DFls according to Table 1-4 of Section 

1.5.2.12.2. 

I - 102 

Cs, Rb - 103 

Others - 103 

While in operation, referring to Table 1-4 of Section 1.5.2.12.2 

demineralizer DFls are: 

I - 10 

Cs, Rb - 2 

Others - 10 

Therefore, for II I II and 1I0thers, II 90% of the act i vi ty 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 DFls entered on Card 13 

would be: 

I 100.0 

Cs, Rb - 2000.0 

Others - 1000.0 

Ca rd 14 

Spaces 29-33 IICollection Time. 1I Using the value from Table A-l 

of 180 days for the regeneration frequency 

Enter 180.0 days in spaces 29-33. 

A-3 
Use the same "Process time" and "fraction discharged" as is 

indicated for the stream in which the regeneration wastes are 

processed as indicated in Section 1.5.2.12.4 of this report. 

Note: If there is more than one stream for which non-condensate 

regenerant deminera1izer is used, follow the same procedures 

explained under item A2 for the other stream or streams. 

3. Components in Service 

a. If the waste is processed through a component other than a 

regenerab1e deminera1izer prior to processing by the regenerab1e 

deminera1izer, the activity in the steam entering the demineralizer 

will be less than the activity entered as described above. To 

compensate for this difference, the OF's for the regenerant 

waste calculation should be adjusted in a manner similar to 

that described above. The product of the OF's should be used. 

b. If two regenerable deminera1izers are used in series, follow 

the procedure in a above. Adjust the OF for nuclides removed 

from the waste stream, by using the product of the OF's for 

two demineralizers in series, i.e., consider the two demineralizers 

as one larger deminera1izer. 

4. Use of Computer Calculated Result 

Combine the values printed out in the individual liquid source term 

columns for the system in which the demineralizer is being regenerated 

(not the adjusted total value) with the normal liquid source term 

run values. 00 not use the adjusted total value from the right hand 

column since the source term run to which the regenerant waste run 

will be added has already been adjusted. 

A-4 
REFERENCES 

1. American National Standards Source Term Specification, ANSI N237-1976, 

American National Standards Institute. 

2. Regulatory Guide 1.140, "Design, Testing, and Maintenance Criteria 

for Normal Ventilation Exhaust System Air Filtration and Adsorption 

Units of Light-Water-Coo1ed 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. NUREGjCR-0140, "In-Plant Source Term Measurements at Ft. Calhoun 

Station, Unit 1," July 1978. 

5. NUREGjCR-0715, "In-Plant Source Term Measurements at Zion Station," 

May, 1979. 

6. NUREGjCR-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 lP, "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 ~1easurements at Brunswick - 2," C. Lin and H. Kenitzer 

(to be published). 

14. NUREG-0017, "Calculation of Releases of Radioactive Materials in 

Gaseous and Liquid Effluents from Pressurized Water Reactors 

(PWR-GALE Code)," Ap ril 1976. 
15. NUREG-0016, Rev. 1, "Calculation of Releases of Radioactive Materials 

in Gaseous and Liquid Effluents from Boiling Water Reactors (BWRGALE Code)," January 1979. 

16. Pilgrim 2 Preliminary Safety Analysis Report (PSAR) (Docket No. 

50-471), Appendix llE, Amendment 12, November 1974. 

17. Letter from K. Seyfrit, Technical Assistance Branch, AEC, Regulatory 

Operations, to G. Lainas, Containment Systems Branch, AEC, Licensing 

"PWR Purging and Venting Experience," September 3, 1974. 

18. NUREG-75/087, "U. S. Nuclear Regulatory Commission Standard Review 

Plan," Section 6.2.4, "Containment Isolation System," Rev. 1, 

November 1978. 

19. Rochester Gas and Electric Corporation, "Radioactivity in the 

Containment Building Atmosphere of Ginna Station," A. R. Piccot, 

1971 • 

20. Nuclear Containment Systems, Incorporated, NCS-llOl, "Dynamic 

Adsorption Coefficient and Its Application for Krypton-Xenon Delay Bed Design," J. L. Kovach, Draft, 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 St reams by Adsorption," ORNL Central Fi 1 es 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. 

24. Letter from Kerndraftwerk Lingen GMBH to Peter Lang, North American 

Carbon, "Gas Delay System at KWL," December 30, 1970. 

25. General Electric Company, NEDO-10751, "Experimental and Operational 

Confirmation of Offgas 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 Coefficients for Noble Gases on Activated Carbon," 12th Air Cleaning 

Conference, August 1972. 

28. General Electric Co., NEDO-20116, "Experimental and Operational 

Confirmation of Offgas System Design Parameters," C. W. Miller, 

October 1973. 

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29. ANSI/ANS 55.6-1979, "American National Standard Liquid Radioactive 

Waste Processing System for Light Water Reactor Plants," American 

National Standards Institute, April 1979. 

30. NWT 133-1, "OTSG Secondary Water Chemi stry Study," Nucl ear Water 

and Waste Technology, March 1978. 

31. NWT 133-2, "OTSG 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 Study 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 Effl uent From Pressuri zed 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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