Text

Rev 18 to Odcm
	ML20217Q111
Person / Time
Site:	Seabrook
Issue date:	02/25/1998
From:	Robinson D; NORTH ATLANTIC ENERGY SERVICE CORP. (NAESCO)
To:	;
Shared Package
ML20217Q094	List: ML20217Q091; ML20217Q097; ML20217Q103; ML20217Q111;
References
	PROC-980225, NUDOCS 9803110377
	Download: ML20217Q111 (300)
	v • d • e

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- co m a corr 0 70

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PROGRAM MANUAL North as)4 Atlantic Ff Offsite Dose Calculation Manual 1

I Effective Date: 02/25/98

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ODCM Manual Owen Rev.I8 D. A. Robinson I

l 9803110377 980305 '

l FDR ADOCK 05000443 P PDR

DISCIAIMER OF RESPONSIBILITY f-i

., This document was prepared by Yankee Atomic Electric Company (* Yankee"). The use of information contained in this document by anyone other than Yankee, or the Organization for which the document was prepared under contract, is not authorized and, with resnect to any unauthorized use, neither Yankee nor its officers, directors, agents, or employees assume any obligation, responsibility, or liability or make any warranty or representation as to the accuracy or completeness of the material contained in this document.

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Page 1 of 1 ODCM Rev. 16 L

ABSTRACT l The Offsite Dose Calculation Manual (ODCM) is divided into two parts: (1) the in-plant Radiological Effluent Monitoring Program requirements for liquid and gas sampling and analysis, along with the Radiological Environmental Monitoring Program l requirements (Part A); and (2) approved methods to determine effluent monitor l

setpoint values and estimates of doses and radionuclide concentrations occurring beyond the boundaries of Seabrook Station resulting from norms 1 Station operation (Part B).

The sampling and analysis programs in Part A provide the inputs for the models of Part B in order to calculate offsite doses and radionuclide concentrations necessary to determine compliance with the dose and concentration requirements of the Station Technical Specification 3/4.11. The Radiological Environmental Monitoring Program required by Technical Specification 3/4.12 and outlined within this manual provides the rnans to determine that measurable concentrations of radioactive materials released as a result of the operation of Seabrook Station are not significantly higher than expected.

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I Page 1 of 1 ODCM Rev. 16

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- Appendix X, Rev.18, was not distributed to controlled copy holders of the ODCM. Appendix X is available in RMD.

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i Appendix Y, Rev.18, was not distributed to controlled copy holders of the ODCM. Appendix Y is available in RMD.

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OFFSITE DOSE CALCUIATION MANUAL (ODCM)

TABLE OF CONTENTS CONTENT fAC3 PART A: RADIOIDGICAL EFFIEENT CONTROL AND ENVIRol0 ENTAL MONITORING PROGRAMS 10 INTRODUCTION A.1-1 2.0 ~'ESPONSIBILITIES (PART A) A.2-1 A.3-1 3.0 DEFINITIONS 3.1 GASEOUS RADWASTE TREATMENT SYSTEM A.3-1 3.2 MEMBER (S) 0F THE PUBLIC A.3-1 3.3 UNRESTRICTED AREA A.3-1 3.4 VENTIIATION EXHAUST TREATMENT SYSTEM A.3-1 4.0 CONTROL AND SURVEILIANCE REQUIREMENTS: APPLICABILITY A.4-1 5.0 RADIOACTIVE EFFLUENT MONITORING INSTRUMENTATION A.5-1 5.1 LIQUIDS A.5-1 5.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION A.5-6 6.0 RADIOACTIVE LIQUID EFFLUENTS A.f-1 6.1 CONCENTRATION A.6-1 6.2 DOSE A.6-8 6.3 LIQUID RADWASTE TREATMENT SYSTEM A.6-10 7.0 RADIOACTIVE LIQUID EFFLUENTS A.7.1 7.1 DOSE RATE A.7.1 7.2 DOSE - NOBLE GASES A.7-7 Page 1 ODCM Rev. 18

TABLE OF CONTENTS CONTENT ZAGE l

PART A: RADIOIDGICAL EFFIEENT CONTROL AND ENVIRONMENTAL MONITORING PROGRAMS 7.3 DOSE - 10 DINE-131, IODINE-133, TRITIUM, AND RADIOACTIVE MATERIAL IN PARTICULATE FORM A.7-9 7.4 GASEOUS RADVASTE TREATMENT SYSTEM A.7-ll 8.0 TOTAL DOSE A,8-1 l

9.0 RADIOLOGICAL ENVIRONMENTAL MONITORING A.9-1 1 9.1 MONITORING PROGRAM A.9-1 9.2 IAND USE CENSUS A.9 11 1 9.3 INTERLABORATORY COMPARISON PROGRAM A.9-13 10.0 REPORTS A.10-1 10.1 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT A.10-1 10.2 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT A.10-2 PAL' B: RADIOIDGICAL CALCUIATIONAL METHODS AND PARAMETERS

1.0 INTRODUCTION

B.1-1 1.1 RESPONSIBILITIES FOR PART B B.1 1 1.2

SUMMARY

OF METHODS, DOSE FACTORS, LIMITS, CONSTANTS, VARIABLES AND DEFINITIONS B.1 2 2.0 METHOD TO CALCUIATE OFF-SITE LIQUID CONCENTRATIONS B.2 1 2.1 METHOD TO DETERMINE F 3E85 AND Ct '* B.2-1 2.2 METHOD TO DETERMINE RADIONUCLIDE CONCENTRATION FOR EACH LIQUID EFFLUENT SOURCE B.2-2 2.2.1 Waste Test Tanks B.2 2 2.2.2 Turbine Building Sump B.2-3 I

Page 2 ODCM Rev. 18 l

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TABLE OF CONTENTS CONTENT FACI PART B: RADIOIAGICAL CAILUTATIONAL METHODS AND PARAMETERS 2.0 METHOD TO CALCULATE OFF-SITE LIQUID CONCENTRATIONS 2.2.3 Steam Generator Blowdown Flash Tank B.2 -

2.2.4 Primary Component Cooling Water (PCCW) System B.2-3 3.0 0FF-SITE DOSE CALCULATION METHODS B.3-1 3.1 INTRODUCTORY CONCEPTS B.3-2 3.2 METHOD TO CALCUIATE THE TOTAL BODY DOSE FROM LIQUID RELEASES B.3 4 3.2.1 Method I B.3-4 3.2.2 Method II B.3-5 3.3 METHOD TO CALCUIATE MAXIMUM ORGAN DOSE FROM LIQUID RELEASES B.3-6 3.3.1 Method I B.3-6 3.3.2 Method II B.3-7 3.4 METHOD TO CALCULATE THE TOTAL BODY DOSE RATE FROM NOBLE GASES B.3-8 3.4.1 Method I B.3 8 3.4.2 Method II B.3-10 3.5 METHOD TO CALCUIATE THE SKIN DOSE RATE FROM NOBLE CASES B.3-il 3.5.1 Method I B.3-11 3.5.2 Method II B.3-14 3.6 METHOD TO CALCUIATE THE CRITICAL ORGAN DOSE RATE FROM 10 DINES, TRITIUM AND PARIICULATES WITH Tua GREATER THAN 8 DAYS B.3-15 3.6.1 Method I B.3-15 3.6.2 Method II B.3-18 1

I Page 3 ODCM Rev. 18

TABLE OF CONTENTS CONTENI E6.GE PART B: RADIOIDGICAL CAILUIATIONAL METHODS IJiD PARAMETERS l l

3.7 METHOD TO CALCULATE THE GAMMA AIR DOSE FROM NOBLE GASES B.3-19 3.7.1 Method I B.3-19 3.7.2 Method II B.3-21 3.8 METHOD TO CALCUIATE THE BETA AIR DOSE FROM NOBLE GASES B.3-22 3.8.1 Method I B.3-22 l

3.8.2 Method II B.3-24 3.9 METHOD TO CALCUIATE THE CRITICAL ORGAN DOSE FROM IODINES, TRITIUM AND PARTICULATES B.3-25 1

3. 9.1 Method I B.3-25 l 3.9.2 Method II B.3-27 3.10 METHOD TO CALCULATE DIRECT DOSE FROM PLANT OPERATION B.3-28 3.10.1 Method B.3-28 3.11 DOSE PROJECTIONS B.3-29 3.11.1 Liquid Dose Projections B.3-29 3.11.2 Gaseous Dose Projections B.3-29 4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM B.4-1 5.0 SETPOINT DETERMINATIONS B.5-1 5.1 LIQUID EFFLUENT INSTRUMENTATION SETPOINTS B.5-1 5.1.1 Liquid Waste Test Tank Monitor (RM-6509) B.5-1 5.1.2 Turbine Building Drains Liquid Effluent Monitor (RM-6521) B.5-5 5.1.3 Steam Generator Blowdown Liquid Sample Monitor (RM 6519) B.5 6 5.1.4 PCCW Head Tank Rate of-Change Alarm Setpoint B.5-6 Page 4 ODCM Rev. 18

TABLE OF CONTENTS CONTENT A fAGE l PART B: RADIOIDGICAL CALCUIATIONAL METHODS AND PARAMETERS l

5.1 LIQUID EFFLUENT INSTRUMENTATION SETPOINTS i

1 5.1.5 PCCW Radiation Monitor B.5-7 l 5.2 GASEOUS EFFLUENT INSTRUMENTATION SETPOINTS B.5-8 5.2.1 Plant Vent Wide-Range Gas Monitors (RM-6528-1, 2 and 3) B.5-8 5.2.2 Waste Gas System Monitors (RM-6504 and RM-6503) B.5-11 5.2.3 Main Condenser Air Evacuation Monitor (RM-6505) B.5-12 6.0 LIQUID AND CASEOUS EFFLUENT STREAMS, RADIATION MONITORS AND RADWASTE TREATMENT SYSTEMS B.6-1 7.0 BASES FOR DOSE CALCULATION METHODS B 7-1 i l

7.1 LIQUID RELEASE DOSE CALCULATIONS B.7-1 1

7.1.1 Dose to the Total Body B.7-4 l 7.1.2 Dose to the Critical Organ B.7-4 l 1

7.2 GASEOUS RELEASE DOSE CALCULATIONS B.7-7 7.2.1 Total Bcdy Dose Rate From Noble Cases B.7-7 7.2.2 Skin Dose Rate From Noble Gases B 7-9 7.2.3 Critical Organ Dose Rate From Iodines, Tritium and Particulates With Half-Lives Greater Than Eight Days B.7-12 1 7.2.4 Gamma Dose to Air From Noble Cases B.7-14 7.2.5 Beta Dose to Air From Noble Cases B.7-16 7.2.6 Dose to Critical Organ From Iodines, Tritium and Particulates With Half-Lives Greater Than Eight Days B.7-18 7.2.7 Special Receptor Gaseous Release Dose Calculations B.7-20 l 7.3 RECEPTOR POINTS AND AVERAGE ATMOSPHERIC DISPERSION FACTORS FOR IMPORTANT EXPOSURE PATHWAYS B.7-34 7.3.1 Receptor Locations B.7-34 Page 5 ODCM Rev. 18

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I TABLE OF CONTENTS CONTENT fA91 1

PART B: RADIOIDGICAL CAILUIATIONAL METHODS AND PARAMETERS l

7.3 RECEPTOR POINTS AND AVERAGE ATMOSPHERIC DISPERSION FACTORS FOR IMPORTANT EXPOSURE PATHWAYS j 7.3.2 Seabrook Station Atmospheric Dispersion Model B.7-35 7.3.3 Average Atmospheric Dispersion Factors for Receptors B.7-35 8.0 BASES FOR LIQUID AND GASEOUS MONITOR SETPOINTS B.8-1 8.1 BASIS FOR THE LIQUID WASTE TEST TANK MONITOR SETPOINT B.8-1 8.2 BASIS FOR THE PLANT VENT WIDE RANGE CAS MONITOR SETPOINTS B.8-5 8.3 BASIS FOR PCCW HEAD TANK RATE-OF-CHANGE AIARM SETPOINT B.8-10 1 8.4 BASIS FOR WASTE CAS PROCESSING SYSTEM MONITORS (RM-6504 AND RM-6503) B.8-11 8.5 BASIS FOR THE MAIN CONDENSER AIR EVACUATION MONITOR SETPOINT (RM-6505) B.8-14 8.5.1 Example for the Air Evacuation Monitor Setpoint Durin5 Normal Operations B.8-14 8.5.2 Example for the Air Evacuation Monitor Setpoint During Start Up (Hogging Mode) B.8-16 REFERENCES R-1 APPENDIX A: METh0D I DOSE CONVERSION FACTORS A-1 APPENDIX B: CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND B-1 TAKEN FROM 10 CFR 20.1-20.602, APPENDIX B APPENDIX C: EMS SOFTWARE DOCUMENTATION C-1 APPENDIX X: ODCM, REV. 18 (PENDING NRC APPROVAL) X-1 l

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APPENDIX Y: 10 CFR 50.59 EVALUATION Y-1 Page 6 ODCM Rev. 18 l

TABLE OF CONTENTS

[giggg LIST OF TABLES AND FIGURES E62E l

PART A TABLES A.3.1 Frequency Notation A.3-2 A.3.2 Operational Modes A 3-2 A.S.1-1 Radioactive Liquid Effluent Monitoring Instrumentation A.5-2 A.5.1-2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements A.5-4 l

A.S.2-1 Radioactive Caseous Effluent Monitoring Instrumentation A.5-7 A.5.2-2 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements A.5-9 A.6.1-1 Radioactive Liquid Waste Sampling and Analysis Program A.6-2 A.7.1-1 Radioactive Caseous Waste Sampling and Analysis Program A.7-3 A.9.1-1 Radiological Environmental Monitoring Program A.9-3 l A.9.1-2 Detection Capabilities for Environmental Sample Analysis *'f'8 A.9-7 A.9.1-3 Reporting Levels for Radioactivity Concentrations in Environmental Samples A.9-10 I PART B TABLES B.1-1 Summary of Radio'.ogical Effluent Part A Controls and Implementing Equations B.1 3 B.1-2 Summary of Method I Equations to Calculate Unrestricted Area Liquid Concentrations B.1-6 l B.1-3 Summary of Method I Equations to Calculate Off-Site Doses from Liquid Releases B.1-7 B.1-4 Summary of Method I Equations to Calculate Dose Rates B.1-8 Page 7 ODCM Rev. 18

I&&LE OF CONTENTS NUMBER LIST OF TABLES AND FIGURES EA93 l

PART B TABLES (Continued) l l

B.1-5 Summary of Method I Equations to Calculate Doses to Air from Noble Cases B.1-11 l B.1-6 Summary of Method I Equations to Calculate Dose to an Individual from Tritium, Iodine and Particulates B.1-13 B.1-7 Summary of Methods for Setpoint Determinations B.1-14 B.1-8 Summary of Variables B.1-15 B.1-9 Definition of Terms B.1-22 B.1-10 Dose Factors Specific for Seabrook Station for Noble Gas Releases B.1-23 j

B.1-11 Dose Factors Specific for Seabrook Station for Liquid Releases B.1-24 B.1-12 Dose and Dose Rate Factors Specific for Seabrook Station for Iodines, Tritium and Particulate Releases B.1-25 B.1-13 Combined Skin Dose Factors Specific for Seabrook Station Special Receptors for Noble Gas Release B.1-26 B.1-14 Dose and Dose Rate Factors Specific for the Science and Nature Center for Iodine, Tritium, and Particulate Releases B.1-27 B.1-15 Dose and Dose Rate Factors Specific for the " Rocks" for i Iodine Tritium, and Particulate Releases B.1-28 B.4-1 Radiological Environmental Monitoring Stations B.4-2 B.7 1 Usage Factors for Various Liquid Pathways at Seabrook B.7-6 i Station B.7-2 Environmental Parameters for Gaseous Effluents at Seabrook l Station B.7-31 1 Page 8 ODCM Rev. 18 l

TABLE OF CONTENTS l

NUMBER LIST OF TABIES AND FIGURES EAfd; PART B TABI25 (Continued)

B.7-3 Usage Factors for Various Gaseous Pathways at Seabrook Station B.7-33 B.7-4 Seabrook Station Long-Tern Average Dispersion Factors Primary Vent Stack B.7-38 B.7-5 Seabrook Station Long-Term Average Dispersion Factors for

! Special'(On-Site) Receptors Primary Vent Stack B.7-39 B.7-6 Seabrook Station Long-Term Atmospheric Diffusion and I Deposition Factors Ground-Level Release Pathway B.7-40 I

PART 5 FIGURES B.4-1 Radiological Environmental Monitoring Locations Within 4 kilometers of Seabrook Station B,4-5 l B.4-2 Radiological Environmental Monitoring Locations Between l 4 kilotaters and 12 kilometers from Seabrook Station B.4-6 l

i 2 l B.4-3 Radiological Environmental Monitoring Locations Outside !

12 kilometers of Seabrook Station B.4-7 i

B.4-4 Direct Radiation Monitoring Locations Within 4 kilometers of Seabrook Station B.4 8 ,

l l B.4-5 Direct Radiation Monitoring Locations Between 4 kilometers ;

i and 12 kilometers from Seabrook Station B.4-9 B.4-6 Direct Radiation Monitoring Locations outside 12 kilometers ,

l of Seabrook Station B.4-10 l B.6-1 Liquid Effluent Streams, Radiation Monitors, and Radwaste Treatment System at Seabrook Station B.6 2 B.6-2 Gaseous Effluent Streams, Radiation Monitors, and Radwaste Treatment System at Seabrook Station B.6-3 i

Page 9 ODCM Rev. 18

LIST OF EFFECTIVE PAGES l FAGE EJUL. EMil Ki3L. IAG.E BEL.

l Cover 18 B.1-16 17 B.4-4 16 I B.1-17 16 B.4-5 17 Abstract 16 B.1-18 16 B.4-6 17 ;

B.1-19 17 B.4-7 16 I TOC 1 - 9 18 B.1-20 17 B.4-8 16 B.1-21 17 B.4-9 16 i LOEP 1 & 2 18 B.1-22 17 B.4-10 16 )

B.1-23 16 l A.1 1 16 B.1-24 16 B.5-1 17 B.1-25 16 B.5-2 17 A.2-1 16 B.1-26 16 B.5-3 '

17 B.1-27 16 B.5-4 17 A.3 1 16 B.1-28 16 B.5-5 17 A.3-2 16 B.5-6 17 A.3-3 16 B.2-1 16 B.5-7 17 A.3 4 16 B.2-2 16 B.5-8 17 !

A.3-5 16 B.2-3 16 B.5-? 17 A.3-6 16 B.5-10 17 A.3-7 16 B.3-1 16 B.5-11 17 8.3 2 16 B.5-12 17 A.4-1 16 B.3-3 16 A.4-2 16 B.3-4 16 B.6-1 17 A.4-3 16 B.3-5 16 B.6-2 17 A.4-4 16 B.3-6 16 B.6-3 16 .

A.4-5 16 B.3-7 16 I B.3-8 16 B.7-1 16 !

A.5-1 16 B.3-9 16 B.7-2 16 A.5-2 16 B.3-10 16 B.7-3 16-A.5-3 16 B.3-11 16 B.7-4 16 A.5-4 16 B.3-12 16 B.7-5 16 A.5-5 16 B.3-13 16 B.7-6 16 A.5 6 16 B.3-14 16 B.7-7 16 A.5-7 16 B.3-15 16 B.7-8 16 A.5-8 16 B.3-16 16- B.7-9 16 A.5-9 16 8.3-17 16 B.7 10 16 A.5-10 16 B.3-18 16 B.7-11 16 B.3-19 16 B.7-12 16 B.1-0 16 B.3 20 16 B.7-13 16 B.1-1 17 B.3-21 16 B.7-14 16 B.1-2 17 B.3-22 16 B.7-15 16 B.1-3 16 B.3-23 16 B.7-16 16 B.1-4 16 B.3-24 16 B.7-17 16 B.1-5 16 B.3-25 16 B.7-18 16 B.1-6 16 8.3-26 16 B.7 19 16 B.1-7 16 B.3-27 16 B.7 20 16 B.1-8 16 B.3-28 16 B.7-21 16 B.1-9 16 B.3-29 16 B.7-22 16 B.1-10 16 B.3-30 16 B.7-23 16 B.1-11 16 B.3-31 16 B.7-24 16 B.1-12 16 B.7-25 16 B.1 13 17 B.4-1 17 B.7-26 16 B.1-14 17 B.4-2 17 B.7-27 16 l

B.1-15 17 B.4-3 16 B.7-28 16 Page 1 ODCM Rev. 18

I i LIST OF EFFECTIVE PAGES 1

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l IAGI IUCL EAGI EIL EAGI BIL l B.7-29 16 B-1 16 X-B.2-1 thru l B.7-30 16- B-2 16 X-B.2-3 18 L B.7 16 B-3 16 B.7-32 16 B-4 16 X-B.3-1 thru B.7 33 16 B5 16 X-B.3-31 18 i B.7-34 B-6 i

16 16 B.7-35 16 B-7 16 X-B.4-1 thru B.7-36 16 B-8 16 X-B.4-10 18 B.7 37 16 B-9 16 l B.7-38 16 B-10 16 X-B.5-1 thru l B.7-39 16 B-11 16 X-B.5-12 18 B.7-40 16 C-1 16 X-8.6 1 thru B.8-1 17 C-2 16 X-B.6-3 18 B.8-2 17 C-3 16 B.8-3 17 C-4 16 X-B.7-1 thru B.8-4 17 C-5 16 X-B.7-40 18 B.8-5 16 C-6 16 B.8-6 16 X-B.8-1 thru B.8-7. 16 TOC X-1 thru X-B.8-19 18 I B.8-8 16 X-9 18 B.8-9 16 X-R-1 18 B.8-10 16 Abstract X-1 18 B.8-11 16 Y-1 thru B.8-12 16 X-A.1-1 18 Y-11 18 B.8-13 16 B.8-14 16 X A.2-1 18 B.8-15 16 B.8-16 16 X-A.3-1 and B.8-17 16 X-A.3-2 18 B.8-18 16 B.8-19 16 X-A.4 1 and X A.4-2 18 R-1 16 X-A.5-1 thru A-1 16 X-A.5-10 18 ,

I A-2 16 A-3 16 X-A.6-1 thru A-4 16 X-A.6-11 18 A-5 16 A-6 16 X-A.7-1 thru l A-7 16 X-A.7-12 18 i

A8 16 A-9 16 X-A.8-1 and A-10 16 X-A.8-2 18 A-1L 16 A-12 16 X A.9-1 thru A-13 16 X-A.9-13 18 A 14 16 A-15 16 X A.10-1 thru A-16 16 X-A.10-3 18 A-17 16 A-18 16 X-B.1-0 thru

! 'A-19 16 X-B.1-28 18 Page 2 ODCM Rev. 18

OFFSITE DOSE CALCUIATION MANUAL (ODCM)

TABLE OF CONTENTS CONTENT PAGE X-PART A: RADIOIDGICAL EFFLUENT CONTROL AND ENVIROl0GNTAL MONITORING PROGRAMS

1.0 INTRODUCTION

A.1-1 2.0 RESPONSIBILITIES (PART A) A.2-1 3.0 DEFINITIONS A.3-1 3.1 VENTI 1ATION EXHAUST TREATMENT SYSTEM A.3-1 4.0 CONTROL AND SURVEILIANCE REQUIREMENTS: APPLICABILITY A.4-1 5.0 RADIOACTIVE EFFLUENT MONITORING INSTRUMENTATION A.5-1 5.1 LIQUIDS A.5-1 5.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION A.5-6 6.0 RADIOACTIVE LIQUID EFFLUENTS A.6-1 6.1 CONCENTRATION A.6-1 6.2 DO.*E A.6-8 6.3 LIQUID RADWASTE TREATMENT SYSTEM A.6-10 7.0 RADI0 ACTIVE GASEOUS EFFLUENTS A.7.1 7.1 DOSE RATE A.7.1 7.2 DOSE - NOBLE GASES A.7-7 7.3 DOSE - 10 DINE-131, 10 DINE-133, TRITIUM, AND RADIOACTIVE MATERIAL IN PARTICUIATE FORM A.7 9 7.4 GASEOUS RADWASTE TREATMENT SYSTEM A.7-ll 8.0 TOTAL DOSE A.8 1 Page X-1 ODCM Rev. 18

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TABLE OF CONTENTS l CONTENT PACE X-I PART A: RADIOI4GICAL EFFINENT COtrIROL AND ENVIRONMENTAL MONITORING PROGRAMS 9.0 RADIOIDGICAL ENVIRONMENTAL MONITORING A.9-1 9.1 MONITORING PROGRAM A.9-1 l 9.2 LAND USE CENSUS A 9-11 l

9.3 INTERIABORATORY COMPARISON PROGRAM A.9-13 10.0 REPORTS A.10-1 10.1 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT A.10-1 10.2 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT A.10-2 PART B: RADIOIAGICAL CAIDUIATIONAL METHODS AND PARAMETERS

1.0 INTRODUCTION

B.1-1 1.1 RESPONSIBILITIES FOR PART B B.1-1 1.2

SUMMARY

OF METHODS, DOSE FACTORS, LIMITS, CONSTANTS, VARIABLES AND DEFINITIONS B.1-2 2.0 METHOD TO CALCULATE OFF-SITE LIQUID CONCENTRATIONS B.2-1 l

2.1 METHOD TO DETERMINE F t D* AND Ct " B.2-1 2.2 METHOD TO DETERMINE RADIONUCLIDE CONCENTRATION FOR EACH LIQUID EFFLUENT SOURCE B.2-2 2.2.1 Waste Test Tanks B.2-2 2.2.2 Turbine Building Sump B.2-3

! 2.2.3 Steam Generator Blowdown Flash Tank B.2-3 2.2.4 Primary Component Cooling Water (PCCW) System B.2-3 3.0 0FF-SITE DOSE CALCUIATION METHODS B.3-1 3.1 INTRODUCTORY CONCEPTS B.3-2 Page X-2 ODCM Rev. 18

r TABLE OF CONTENTS CONTENT PAGE X-PART B: RADIOLOGICAL CAIEUIATIONAL METHODS AND PARAMETERS l 3.0 0FF-$1TE DOSE CALCULATION METHODS l

l 3.2 METHOD TO CALCULATE THE TOTAL BODY DOSE FROM LIQUID RELEASES B.3-4 3.2.1 Method I B.3 4 3.2.2 Method II B.3-5 l l 3.3 METHOD TO CALCUIATE MAXIMUM ORGAN DOSE FROM LIQUID l

! RELEASES B.3-6 I 3.3.1 Method I B.3-6 3.3.2 Method II B.3-7 3.4 METHOD TO CALCULATE THE TOTAL BODY DOSE RATE FROM NOBLE GASES B.3-8 l 3.4.1 Method I B.3-8 3.4.2 Method II B.3-10

! 3.5 METHOD TO CALCULATE THE SKIN DOSE RATE FROM NOBLE GASES B.3-11 1

3.5.1 Method I B.3-11 l 3.5.2 Method II B.3-14 3.6 METHOD TO CALCUIATE THE CRITICAL ORGAN DOSE RATE FROM

! 10 DINES, TRITIUM AND PARTICUIATES WITH Tua GREATER l THAN 8 DAYS B.3-15 3.6.1 Method I B.3-15 3.6.2 Method II B.3-18 3.7 METHOD TO CALCUIATE THE GAMMA AIR DOSE FROM NOBLE CASES B.3-19 3.7.1 Method I B.3-19 3.7.2 Method II B.3-21 I

l Page X-3 ODCM Rev. 18

I TABLE OF CONTENTS l

l CONTENT PAGE X-PART B: RADIOIDGICAL CAIEUIATIONAL METHODS AND PARAMETERS l

3.0 0FF-SITE DOSE CALCUIATION METHODS I

3.8 METHOD TO CALCUIATE THE BETA AIR DOSE FROM NOBLE

! CASES B.3-22 l l

I

3.8.1 Method I B.3-22 3.8.2 Method II B.3-24 3.9 METHOD TO CALCUIATE THE CRITIQAL ORGAN DOSE FROM 10 DINES, TRITIUM AND PARTICULATES B.3-25 i

3.9.1 Method I B.3-25 3.9.2 Method II B.3-27

! 3.10 METHOD TO CALCULATE DIRECT DOSE FROM PLANT OPERATION B.3-28 3.10.1 Method B.3-28 3.11 DOSE PROJECTIONS B.3-29 l

3.11.1 Liquid Dose Projections B.3-29 3.11.2 Gaseous Dose Projections B.3-29 l

4.0 RADIOIAGICAL ENVIRONMENTAL MONITORING PROGRAM B.4-1 l

5.0 SETPOINT DETERMINATIONS B.5-1 5.1 LIQUID EFFLUENT INSTRUMENTATION SETPOINTS B.5-1 5.1.1 Liquid Waste Test Tank Monitor (RM 6509) B.5-1 5.1.2 Turbine Building Drains Liquid Effluent Monitor j (RM-6521) B.5-5 l 5.1.3 Steam Generator Blowdown Liquid Sample Monitor (RM-6519) B.5 6 5.1.4 PCCW Head Tank Rate-of Change Alara Setpoint B.5-6 5.1.5 PCCW Radiation Monitor B.5-7

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Page X-4 ODCM Rev. 18

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TABLE OF CONTENTS CONTENT PAGE X-l PART B: RADIOIDGICAL CAI4UIATIONAL METHODS AND PARAMETERS l

5.0 SETPOINT DETEftMINATIONS 5.2 GASEOUS EFFLUENT INSTRUMENTATION SETPOINTS B.5-8 l

5.2.1 Plant Vent Wide-Range Gas Monitors (RM-6528-1, 2 and 3) B.5-8 ;

5.2.2 Waste Gas System Monitors (RM-6504 and RM-6503) B.5-11 !

5.2.3 Main Condenser Air Evacuation Monitor (RM-6505) B.5-12 I 6.0 LIQUID AND CASEOUS EFFLUENT STREAMS, RADIATION MONITORS AND RADWASTE TREATMENT SYSTEMS B.6-1 7.0 BASES FOR DOSE CALCULATION METHODS B.7-1 7.1 LIQUID RELEASE DOSE CALCULATIONS B.7-1 7.1.1 Dose to the Total Body B.7-4 7.1.2 Dose to the Critical Organ B.7-4 7.2 GASEOUS RELEASE DOSE CALCULATIONS B.7-7 7.2.1 Total Body Dose Rate From Noble Cases B.7-7 7.2.2 Skin Dose Rate From Noble Cases B.7-9 7.2.3 Critical Organ Dose Rate From Iodines, Tritium and Particulates With Half-Lives Greater Than Eight Days B.7-12 7.2.4 Gamma Dose to Air From Noble Cases B.7-14 7.2.5 Beta Dose to Air From Noble Cases B 7-16 7.2.6 Dose to Critical Organ From Iodines, Tritium and Particulates With Half-Lives Greater Than Eight Days B.7-18 7.2.7 Special Receptor Gaseous Release Dose Calculations B.7-20 7.3 RECEPTOR POINTS AND AVERAGE ATMOSPHERIC DISPERSION FACTORS I FOR IMPORTANT EXPOSURE PATHWAYS B.7-34 I

7.3.1 Receptor Locations B.7-34 Page X-5 ODCM Rev. 18 l

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TABLE OF CONTENTS

! CONTENT PAGE X.

PART B: RADIOIDGICAL CAIEUIATIONAL METHODS AND PARAMETERS i

! 7.0 . BASES FOR DOSE CALCULATION METHODS i

l l 7.3 RECEPTOR POINTS AND AVERAGE ATMOSPHERIC DISPERSION FACTORS l FOR IMPORTANT EXPOSURE PATHWAYS l

7.3.2 Seabrook Station Atmospheric Dispersion Model B.7-35 l 7.3.3 Average Atmospheric Dispersion Factors for Receptors B.7-35 I

l 8.0 BASES FOR LIQUID AND GASEOUS MONITOR SETPOINTS B.8-1 8.1 BASIS FOR THE LIQUID WASTE TEST TANK MONITOR SETPOINT B.8-1 l

l 8.2 BASIS FOR THE PLANT VENT WIDE RANGE GAS MONITOR SETPOINTS B.8-5 I

I l

! 8.3 BASIS FOR PCCW HEAD TANK RATE-OF-CHANGE AIARM SETPOINT B.8-10 l

8.4 BASIS FOR WASTE GAS PROCESSING SYSTEM MONITORS (RM-6504 AND RM-6503) B.8-11 l 8.5 BASIS FOR THE MAIN CONDENSER AIR EVACUATION MONITOR SETPOINT (RM-6505) B.8-14

8.5.1 Example for the Air Evacuation Monitor Setpoint During Normal Operations B.8 14 l t

8.5.2 Example for the Air Evacuation Monitor Setpoint During i Start Up (Hogging Mode) B.8 16 REFERENCES R-1 I

, APPENDIX A: METHOD I DOSE CONVERSION FACTORS A-1 )

APPENDIX B: CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND B-1 TAKEN FROM 10 CFR 20.1-20,602, APPENDIX B APPENDIX C: EMS SOFTWARE DOCUMENTATION C-1 i

I

! Page X-6 ODCM Rev. 18

< TABLE OF CONTENTS NUMBER LIST OF TABLES AND FIGURES PAGE X-PART A TABLES A.3.1 Frequency Notation A.3-2 A.3.2 Operational Modes A.3 2 A.S.1 1 Radioactive Liquid Effluent Monitoring Instrumentation A.5-2 A.5.1-2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements A.5-4 A.5.2-1 Radioactive Caseous Effluent Monitoring Instrumentation A.5-7 A.S.2-2 Radioactive Caseous Effluent Monitoring Instrumentation Surveillance Requirements A.5-9 A.6.1-1 Radioactive Liquid Waste Sampling and Analysis Program A.6-2 A.7.1-1 Radioactive Cassous Waste Sampling and Analysis Program A.7-3 A.9.1-1 Radiological Environmental Monitoring Program A.9-3 A.9.1-2 Detection Capabilities for Environmental Sample Analysis *J 8 A.9-7 A.9.1-3 Reporting Levels for Radioactivity Concentrations in Environmental Samples A.9-10 PART B TABLES B.1-1 Summary of Radiological Effluent Part A Controls and Implementing Equations B.1-3 B.1-2 Summary of Method I Equations to Calculate Unrestricted Area Liquid Concentrations B.1-6 B.1-3 Summary of Method I Equat'ons to Calculate Off-Site Doses from Liquid Releases B.1-7 B.1-4 Summary of Method I Equttions to Calculate Dose Rates B.1-8 I

Page X-7 ODCM Rev. 18

TABLE OF CONTENTS I

NUMBER LIST OF TABIES AND FIGURES PAGE X-l l PART d TABLES (Continued) l B.1-5 Summary of Method I Equations to Calculate Doses to Air from Noble Cases B.1-11 B.1-6 Summary of Method I Equations to Calculate Dose to an l Individual from Tritium, Iodine and Particulates B.1 13 B.1-7 Summary of Methods for Setpoint Determinations B.1-14 B.1-8 Summary of Variables B.1-15 B.1-9 Definition of Terms B.1-22 i

B.1-10 Dose Factors Specific for Seabrook Station for Noble Gas l Releases B.1-23 1

B.1-11 Dose Factors Specific for Seabrook Station for Liquid i Releases B.1-24 l

l B.1-12 Dose and Dose Rate Factors Specific for Seabrook Station for Iodines, Tritium and Particulate Releases B.1-25 l

B.1-13 Combined Skin Dose Factors Specific for Seabrook Station Special Receptors for Noble Gas Release B.1-26 i

B.1-14 Dose and Dose Rate Factors Specific fer the Science and Nature l Center for Iodine, Tritium, and Particulate Releases B.1-27 B.1-15 Dose and Dose Rate Factors Specific for the " Rocks" for Iodine, Tritium, and Particulate Releases B.1-28 B.4-1 Radiological Environmental Monitoring Stations B.4-2 i

B.7-1 Usage Factors for Various Liquid Pathways at Seabrook B.7 6 Station B.7-2 Environmental Parameters for Gaseous Effluents at Seabrook Station B.7-31 Page X-8 ODCM Rev. 18

TABLE OF CONTENTS ,

NUMBER LIST OF TABIES AND FIGURES PAGE X-PART 5 TABLES (Continued)

B.7-3 Usage Factors for Vaitous Gaseous Pathways at Seabrook Station B.7-33 B.7-4 Seabrook Station Long-Tern Average Dispersion Factors Primary Vent Stack B.7-38 B.7-5 Seabrook Station Long-Tern Average Dispersion Factors for Special (On-Site) Receptors Primary Vent Stack B.7-39 B.7-6 Seabrook Station Long-Term Atmospheric Diffusion and Deposition Factors Ground Level Release Pathway B . 7 -t-PART 5 FIGURES B.4-1 Radiological Environmental Monitoring Locations Within 4 kilometers of Seabrook Station B.4-5 8.4-2 Radiological Environmental Monitoring Locations Between 4 kilometers and 12 kilometers from Seabrook Station B.4-6 B.4-1 Radiological Environmental Monitoring Locations Outside 12 kilometers of Seabrook Station B.4-7 I B.4-4 Direct Radiation Monitoring Locations Within 4 kilometers ;

of Seabrook Station B.4-8 '

B.4-5 Direct Radiation Monitoring Locations Between 4 kilometers and 12 kilometers from Seabrook Station B.4-9 B.4 6 Direct Radiation Monitoring Locations Outside 12 kilometers of Seabrook Station B.4-10 B.6-1 Liquid Effluent Streams, Radiation Monitors, and Radwaste Treatment System at Seabrook Station B.6 2

( B.6-2 Gaseous Effluent Streams, Radiation Monitors, and Radwaste Treatment System at Seabrook Station B.6-3 l Pabe X-9 ODCM Rev. 18 l

ABSTRACT The Offsite Dose Calculation Manual (ODCM) is divided into two parts: (1) the Radioactive Effluent Controls Program for both in-plant radiological effluent

. monitoring of liquids and gases, along with the Radiological Environmental l Monitoring Program (REMP) (Part A); and (2) approved methods to determine effluent

! monitor setpoint values and estimates of doses and radionuclide concentrations occurring beyond the boundaries of Seabrook Station resulting from normal Station l operation (Part B).

I The sampling and analysis requirements of the Radioactive Effluent Controls Program, specified in Part A, provide the input.s for the models of Part B in order to calculate offsite doses and radionuclide concentrations necessary to determine compliance with the dose and concentration requirements of the Station Technical Specification 6.7.6 g. The REMP required by Technical Specification 6.7.6.h. and as specified within this manual, provides the means to determine that measurable con-centrations of radioactive materials released as a result of the operation of Seabrook Station are not significantly higher than expected.

l Page X-1 of X-1 ODCM Rev. 18

PART A:

.l RADIOIDCICAL EFFLUENT CONTROL AND ENVIRONMENTAL MONITORING PROGRAMS

1.0 INTRODUCTION

The purpose of Part A of the Offsite Dose Calculation Manual (ODCM) is to define the Radiological Effluent Controls Program-(RECP) conducted by the Station, and provide input to the models in Part B for calculating liquid and gaseous effluent concentrations, monitor setpoints, and offsite doses. .The results of Part B calculations are used to determine compliance with the concentration and dose requirements of the controls contained in Part A of the ODCM, pursuant to Technical Specification 6.7.6.g. I The Radiological Environmental Monitoring Program (REMP) required as'a minimum to be conducted (per Technical ~ Specification 6.7.6.h) is defined in Part A, with the identification of current locations of sampling stations being utilized to meet the The information obtained from the conduct of l program the REMP requirements provides data listed in Part B. levels of radiation and radioactive materials on measurable

'in the environment necessary to. evaluate the relationship between quantities of radioactive materials-released in effluents and resultant radiation doses to individuals from principal pathways of exposure. The data developed in the surveillance and monitoring programs described in Part A to the ODCM provide a means to confirm that measurable concentrations of radioactive materials released as a result of Seabrook Station operations are not significantly higher than expected based on the dose models in Part B.

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, i X-A.1-1 ODCM Rev. 18

l 2.0 RESPONSIBfLITTES (PART A)

All changes to the ODCM shall be reviewed by the Station Operation Review l

Committee (SORC), approved by the Station Director, and documented per l Administrative Concrol 6.13 of the T -hnical Specifications. Changes made to Part A

shall be submitted to the NRC for its information in the Annual Radioactive Effluent l Release Report for the period in which the change (s) was made effective, pursuant to T.S. 6.13.

It shall be the responsibility of the Station Director to ensure that the ODCM is used in the performance of the Radioactive Effluent Control and Environmental l Monitorins Program implementation requirements, as identified under Administrative l Controls 6.7.6.g and 6.7.6.h of the Technical Specifications.

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I X-A.2-1 ODCM Rev. 18

l 3.0 DEFfNITIONS J

The defined terms of this section appear in capitalized type and are applicable throughout these Controls. Terms used in these Controls and not defined herein have the same definition as listed in the Technical Specifications. If a conflict in definition exists, the definition in the Technical Specifications takes precedence.

l 3.1 VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing. ventilation or vent exhaust geses through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. En61 neered Safety Features Atmospheric Cleanup Systems are not considered to be VENTILATION EXHAUST i TREATMENT SYSTFM components.

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r X-A.3-1 ODCM Rev. 18

EARI_A ,

IAALE A. 3.1 l

< FREQUENCY NOTATION NOTATION F1tEOUENCY S At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

W At least once per 7 days.

M et least once per 31 days.

j Q fd ieast cnce per 92 days.

l SA !.. least once per 184 days.

R At least once per 18 months.

j S/U Prior to each reactor startup.

l N.A. Not ap',11 cable.

P Completed prior to each release.

TABLE A.3.2 OPERATIONAL MODES MODE REACTIVITY Z RATED AVERAGE COOLANT CONDITION,k.re THERMAL POWER

TEMPERATURE l 1. POWER OPERATION 2 0.99 > 51 2350*F l 2 STARTUP 2 0.99 s 51 2350*F l

l I

l HOT STANDBY < 0.99 0 2350*F l l 4. HOT SHUTDOWN < 0.99 0 350*F > T,y, > 200'F ,

1

! 5. COLD SHUTDOWN < 0.99 0 $200*F l

6. REFUELING ** $ 0.95 0 $140*F l l

1

Excluding decay heat.

- Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

X-A.3-2 ODCM Rev. 18

4 CONTROL AND SURVEILIANCE REOUIREMENTS 4.0 APPLICABILITY CONTROLS C.4.0.1 Compliance with the Controls contained in the succeeding Controls is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Control, the associated ACTION requirements shall be met.

C.4.0.2 Noncompliance with a Control shall exist when the requirements of the Control and associated ACTION requirements are not met within the specified time intervals. If the Control is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

C.4.0.3 When a Control is not met, except as provided in the associated ACTION requirements, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in l a. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , l b. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and l c. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Where corrective measures are completed that permit operation under the ACTION requirements, the action may be taken in accordance with the specified time limits as measured from the time of failure to want the Control. Exceptions to these requirements are stated in the individual CONTROLS.

l This Control is not applicable in MODE 5 or 6.

C.4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made when the conditions for the Control are not met and the associated ACTION requires a shutdown if they are not met within a specified time interval. Entry into an OPERATIONAL MODE or specified condition may be made in accordance with ACTION requirements when conformance to them permits continued operation of the facility for an unlimited period of time. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements.

Exceptions to these requirements are stated in the individual Controls.

X-A.4-1 ODCM Rev. 18 l

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SURVEILLANCE REOUIREMENTS I

S.4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or I J

other conditions specified for individual Controls unless otherwise stated in an

{

individual Surveillance Requirement. )

S.4.0.2 Each SURVEILLANCE REQUIREMENT shall be performed within the specified l time interval with

a. A maximum allowable extension not to exceed 25% of the Surveillance Interval, but

b. The combined time interval for any three consecutive Surveillance Intervals shall not exceed 3.25 times the specified Surveillance Interval.

l S.4.0.3 Failure to perform a Surveillance Requirement within the allowed l Surveillance Interval, defined by Control 4.0.2, shall constitute noncompliance with the OPERABILITY requirements for a Control. The time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed. The ACTION requirements may be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit the completion of the Surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Surveillance

]

Requirements do not have to be performed on inoperc u equipment. '

S.4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement (s) associated with the Control has been performed within the stated Surveillance Interval or as otherwise specified. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements.

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MONITORING INSTRUMENTATION

, 5.0 RADI0 ACTIVE EFFLUENT MONITORING INSTRUMENTATION 5.1 LIQUIDS I

CONTROLS' ,

C.S.1 The radioactive liquid effluent monitoring instrumentation channels shown in Table'A.5.1 1 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Control C.6.1.1 are not exceeded. The Alarm / Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters _in the OFFSITE DOSE CALCULATION MANUAL (ODCM), Part B.

l APPLICABILITY: At all times.

l EIl0E:

a. With a radioactive liquid effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable,

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table A.5.1-1.

Restore the inoperable instrumentation to OPERABLE status withia 30 days and, if unsuccessful, explain in the next Annual Radioactive Effluent Release Report pursuant to' Technical Specification 6.8.1.4 and Part A, Section 10.2, of the ODCM, why this inoperability was not corrected in a timely manner.

l c. The provisions of Control C.4.0.3 are not applicable, j SURVEI N CE REOUTREMENTS l

S.S.1 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, l CHANNEL CALIBRATION, and CHANNEL OPERATIONAL TEST at the frequencies shown in Table l A,5.1-2.

BASES The radioactive liquid effluent instrumentation is provided to monitor and control, ,

as applicable, the releases of radioactive materials in liquid effluents during !

actual or potential releases of liquid effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

X-A.5-1 ODCM Rev. 18

8 N

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r 16BLE A.S.1-1 l (Continued)

ACTION STATEMENTS l ACTION 29 - With the number of channels OPERABLE less than the Minimum Channels OPERABLE requirement, effluent releases via this pathway may l continue provided that prior to initiating a release {

a. At least two independent samples are analyzed in accordance with Surveillance S.6.1.1, and i

I

b. At least two technically qualified members of the station staff independently verify the release rate calculations and discharge line valving. ,

Otherwise, suspend release of radioactive effluents via this pathway. )

ACTION 30 - With the number of channels OPERABLE less than the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are analyzed for radioaccivity at a l lower limit of detection of no more than 10-7 microcurie /ml ls

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the specific activity of the secondary coolant is greater than 0.01 microcurie / gram DOSE EQUIVALENT I-131, or i

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of the secondary coolant is less than or equal to 0.01 microcurie / gram l DOSE EQUIVALENT I-131. ,

ACTION 31 - With the number of channels OPERABLE less than the Minimum Channels :

OPERABLE requirement, effluent releases via this pathway may l continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump performance curves generated in place may be used to estimate flow.

ACTION 32 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the radioactivity levels in the Primary Component Cooling Water System and the Service Water System are determined at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 33 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the radioactivity level is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during actual releases.

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X-A.5-3 ODCM Rev. 18 l

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f TABLI A.S.1-2 i (Ccntinusd)

TABLE NQTATIONS t

l 1

(1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and Control Room alarm annunciation occurs if the instrument indicates measured levels above the normal or Surveillance test Alarm / Trip Setpoir.t.

(2) The initial channel calibration for radioactivity measurement instrumentation shall include the use of a known (traceable to National Institute for Standards and Technology) liquid radioactive source positioned in a reproducible geometry with respect to the sensor. These standards shall permit calibrating the system over its normal operating range of energy and rate. For subsequent channel calibrations, sources that have been related to the initial calibration shall be used.

(3) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic, or batch releases are made.

l (4) CHANNEL CHECK shall consist of verifying indication of tank level during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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I 5.3 RADIOACTIVE CASEOUS EFFLUENT MONTTORING INSTRUMENTATION l

l

,. CONTROLS I c.5.2 The radioactive gaseous effluent monitoring instrumentation channels shown in Table'A.5.2-1 shall be' OPERABLE with their Alare/ Trip Setpoints set to casure that the limits of Control C.7.1.1 are not exceeded. The Alarm / Trip Setpoints of these channels meeting Control C.7.1.1 shall be determined and adjusted in accordance with the methodology and parameters in the ODCM (Part B).

l APPLICABILITY: As shown in Table A.S.2-1.

ACTION:

l a. With a radioactive gaseous effluent monitoring instrumentation channel-Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable.

b. With the number of OPERABLE radioactive gaseous effluent monitoring

! instrumentation channels less than the Minimus Channels OPERABLE, take the l ACTION shown in Table A.5.2-1. Restore the inoperable instrumentation to l OPERABLE status within 30 days or, if ansuccessful, explain in the next Annual j Radioactive Effluent Release Report pursuant to Technical Specification j 6.8.1.4 and Part A, Section 10.2, of the ODCM, why this inoperability was not corrected in a timely manner.

c. The provisions of Control C.4.0.3 are not applicable.

l SURVEILIANCE R5nUIREMENTS l

i S.S.2 Each radioactive gaseous affluent monitoring instrumentation channel

shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, l CHANNEL CALIBRATION and CHANNEL OPERATIONAL TEST at the frequencies shown in Table A.S.2-2.

RASES i

The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM (Part B) to ensure that the alarm / trip will

-occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. The sensitivity of any noble gas activity monitors used to show compliance with the gaseous effluent release requirements of' Control C.7.2.1 shall be such that concentrations as low as 1 X 10-8 Ci/cc are measurable.

l l

X A.5-6 ODCM Rev. 18

8 1

v e

R N

O M I 3 5 5 2 2 3 5 5 2 C T 3 3 3 3 3 3 3 3 3 C D A O Y

T I

_ L I

_ B * *

_ A * * * * * * * *

N G * *

I O L I P T P A A T

_ N E

M U

R T

S N S

- I L E

_ G N N NE n

_ I AL i R

HB a O CA 1 1 1 1 1 1 1 1 1 m 1 T R

- I ME e 2 N UP h O MO t 7 5 M I -

- N n 5

- A T I o N M A-E E d L U e X

_ B L s A F .a T F nb E o S i s U n ti O o ) a E i s rt

_ S t s en A a e pi G n c f oop

_ i o T

_ E m r S n V r P U i e I R e ( A s .

T O T H r sar C T r M r X o i eo A I o r Ec o E t l t O N t o Ti t a rei I O i t S t i R c ern D M n i Ya n E i t o A o n o

S m o

o S d n

ssm R S M M N ui A r M Gt E r I ahn C y e Nu y D e ht o t l r e I A t N l e x i E i p o t S ) i O p t ert G v r m

a t a Sd4 v C r

m a

a oa N i R E n0 i R i lf u e n e a

_ . A t S C a5 t L S c R c l o w O 6 c A l w era A p e M o Rm- A E p e o sov E m t l PrM S m t l t e

_ D s a a e F aR s a a F di I a S l t El a D S l . nnr W- G u a r TA - G N u r s aoi e c R e S A e c e e)lMa

) T e n i l Age e I n i l m .g d N l i t w p Wns l G i t p id sr e E b d r o m ia b d r m t eeae s V o o a l a Sd e o E o a a shGs U N I P F S Uil N N I P S lUt n T Ove I l ee t N E oR B at nld o R A S r oebn .

T N l . . . . .

APf . U . . .

tNhoo _

N ( P a b c d e G( o a T a b c A(WNc -

E -

M U

R .

T

_ S *

- N . . . . *

I 1 2 3 4 ***# _

l l l l l l l l l l l l

I TABLI A.S.2-1 (Continu2d)

ACTION STATEMENTS l

l l With the number of channels OPERABLE less than the Minimum Channels

! ACTION 32 -

OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 33 -

With the number of channels OPERABLE less than the Minimum Channels OPERABLE requirement, affluent releases via this pathway may {

l continue provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i

! and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l For RM-6504, RM-6503 may be used as an alternate.

l l ACTION 34 -

(Not Used.)

l ACTION 35 -

With the number of channels OPERABLE less than the Minimum Channels j l OPERABLE requirement, effluent releases via the affected pathway l may continue provided samples are continuously collected with )

auxiliary sampling equipment as required in this document. '

l l

I I

1 l

X-A.5-8 ODCM Rev. 18 i

f

_ 8

_ 1 HS CI .

I v HE e WC D E

R RAINR * *

OI U * * * * * * * *

M FL * *

C I Q D E

SE O EV R DR OU MS S L T A N LN E *

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)

2

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E NA 3 5 V NR ( A. A. R R ( A. A. A.

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N C O

I T

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N E E K 2 M C C . . . . . . .

- U R A.

2 R U E M N

A. A. A. A. A. A. A.

S T

S OH S C N N N N N N N 9 N 5 A I A-E G L N L X B I EK A R NC T O T

NE AH D W W D D D W W D I HC N C O

M T

N E

U L

F f F T -

E S U

S R A U O H _

O T M X r E I r E c E S N Ti r o A O o r St o R t _

C M t o Ya t E a i t S m i S c _

E S n i o n N i _

V A o n Gt o E d I G M o N u) M D n r

T y M I Ae N r I C E e S s y O e A G t l r e Sd a t C l e O N i p o t E ne i p t I A v m t a C al v L m a D R i r a i R O e i A r a R A t e S n RmR t E e S R E c l o w Pr c S l w

_ D A p e M o af A p e o I m t l El o D m t l

_ W- s a a e F TA s) N a a F a S l t n as S l S A

) T G u a r Ago Gs I u r e R e e e e

_ d N c Wni G c e E e n i l it ec n i l s V l i t w p Sd a l o E i t p

_ U b d r o m Uin br N d r m T T o o a l a Ovi oP I o a a N t N N I P F S E om N( B I P S E o A $ rr R M N L . . . . .

AP e .

U . . .

U ( P a b c d e G( T a T a b c R

T S

N . . . .

I 1 2 3 4 l l l l l l l l l l l

L TABLE A.S.2-2

! (Centinusd) l

! TABLE NOTATIONS

At all times. I

- (Not Used.)

- - When the gland seal exhauster is in operation.

- - - The CHANNEL OPERATIONAL TEST for the flow rate monitor shall consist of a verification that the Radiation Data Management System (RDMS) indicated flow is consistent with the operational status of the plant.

Noble Gas Monitor for this release point is based on the main condenser air evacuation monitor.

(1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic l isolation of this pathway and Control Room alarm annunciation occurs if the instrument indicates measured levels above the normal or Surveillance test l Alarm / Trip Setpoint.

l b (2) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that Control Room ]

l alarm annunciation occurs if the instrument indicates measured levels above 1 l the normal or Surveillance test Alarm Setpoint. J

! l (3) The initial channel calibration for radioactivity measurement instrumentation shall include the use of a known (traceable to National Institute for Standards and Technology) radioactive source positioned in a reproducible

' geometry with respect to the sensor. These standards should permit calibrating the system over its normal operating range of rate capabilities.

For subsequent channel caMbrations, sources that have been related to the 4 initial calibration shall be used, i

(4) (Not Used).

(5) The CHANNEL CALIBRATION shall be performed using sources of various activities covering the measurement range of the monitor to verify that the response is linear. Sources shall be used to verify the monitor response only for the l intended energy range.

1 !

I I

i X-A.5 10 ODCM Rev. 18

~

6.0 RADIOACTIVE 1.IOUTD EFFLUENTS 6.1 CONCENTRATION i

CONTROLS C.6.1.1 The concentration of radioactive material released in liquid effluents at the point of discharge from the multiport diffuser (see Technical Specifications Figure 5.1-3) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be l limited to 2 X 10-' microcurie /mi total activity.

l APPLICABILITY: At all times.

l ACTION:

With the concentration ~of radioactive material released in liquid affluents at the point of discharge from the multiport diffuser exceeding the above limits, restore l

the concentration to within the above limits within 15 minutes.

l SURVEILIANCE REOUIREMENTS S.6.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program specified in Table A.6.1-1. 4 S.6.1.2 The results of the radioactivity analyses shall be used in accordance ,

with the methodology and parameters in Part B of the ODCM to assure that the concentrations at the point of release are maintained within the limits of Control C.6.1.1. ,

BASES This Control is provided to ensure that the concentration of radioactive materials released in liquid waste effluents at the point of discharge from the multipart diffuser will be less than the concentration levels specified in 10 CFR Part 20, Appendix B to 20. Table II, Column 2 (most restrictive). This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within (1) the Section II.A design objectives of Appendix I,10 Cf1t Part 50, to a MEMBER OF THE PUBLIC, and (2) the limi m of Appendix I, 10 CFR Part 20.106(e) to the population. The concentration limit. for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent' concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

X-A.6-1 ODCM Rev. 18 1 i

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PWMQ PWMQ _

TABLE A.6.1-1 RADIOACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM (Continued) l Notations (D The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above systom background, that will be detected with 95 percent probability with only 5 percent probability of falsely concluding that a blank observation represei,ca a "real" signal.

For a particular measurement system, which may include radiochemical l separation:

l , 4.66 s3 l E x V x 2.22 x 10' x Y x exp (-Aat)

Where:

LLD -

the "a priori" lower limit of detection (microcurie per unit mass or volume),

s3 -

the scandard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute), j l

E - the counting efficiency (counts per disintegration),

V - the sample size (units of mass or volume),

2.22 x 10s -

the number of disintegrations per minute per microcurie, Y -

the fractional radiochemical yield, when applicable, A - the radioactive decay constant for the particular radionuclide (s-1) , and at - the elapsed time between the midpoint of sample collection and the time of counting (s).

Typical values of E, V, Y, and at should be used in the calculation.

It should be recognized that the LLD is defined as an A orlori (before the fact) limit representing the capability of a measurement system and not as an A nosteriori (after the fact) limit for a particular measurement.

(U A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed to assure representative sampling.

l I

l X-A.6-5 ODCM Rev. 18

TABLE A.6.1-1 RADIOACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROCRAM )

(Continued) l Notations l (Continued)

(3) The principal gamma emitters for which the LLD specification applies include the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99 Cs-134, Cs 137, Ce-141, and Ce-144. This list does not mean that only these l

nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radioactive Effluent Release Report in accordance with Technical Specification 6.8.1.4, and Part A. Section 10.2, of the ODCM. Isotopes which I are not detected should be reported as "not detected." Values determined to be below detectable levels are not used in dose calculations.

(*) A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released.

m A continuous release is the discharge of liquid wastes of a nondiscrete volume, e.g., from a volume of a system that has an input flow during the continuous release.

(6) Sampling and analysis is only required when Steam Generator Blowdown is directed to the discharge transition structure, m Principal gamma emitters shall be analyzed weekly in Service Water. Sample and analysis requirements for dissolved and entrained gases, tritium, gross l alpha, strontium 89 and 90, and Iron 55 shall only be required when analysis for principal gamma emitters exceeds the LLD.

The following are additional sampling and analysis requirements:

a. PCCW sampled and analyzed weekly for principal gamma emitters.

b. Sample Service Water System (SWS) daily for principal gamma emitters whenever primary component cooling water (PCCW) activity exceeds lx10'3 pC/cc.

c. With the PCCW System radiation monitor inoperable, sample PCCW and SWS daily for principal gamma emitters.

l i

d. With a confirmed PCCW/SWS leak and PCCW activity in excess of lx10

pC/ce, sample SWS every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for principal gamma emitters.

l e. The setpoint on the PCCW head tank liquid rate-of-change alarm will be set to ensure that its sensitivity to detect a PCCW/SWS leak is equal to or greater than that of an SWS radiation monitor, located in the unit's I combined SWS discharge, with an LLD of lx10-8 pC/cc. If this sensitivity cannot be achieved, the SWS will be sampled once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

X A.6-6 ODCM Rev. 18

TABLE A 6.1-1 RADIOACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM (Continued)

Notations (Continued) m If the Turbine Building Sump (Steam Generator Blowdown Flash Tank) isolate due to high concentration of radioactivity, that liquid stream will be sampled and analyzed for Iodine-131 and principal gamma emitters prior to release.

m Quarterly composite analysis requirements shall only be required when analysis for principal gamma emitters indicate positive radioactivity.

l l

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r l <

t I X-A.6-7 ODCM Rev. 18

. . . .l

6.2 QQSE CONTROLS C.6.2.1 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (see Technical Specification Figure 5.1-3) shall be limited j

)

a. During any calendar quarter to less than or equal to 1.5 mreas to the whole l body and to less than or equal to 5 areas to any organ, and

b. During any calendar year to less than or equal to 3 mrems to the whole body and to less than or equal to 10 mreas to any organ.

l APPLICABILITY: At all times.

l ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.8.2, a l Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and ;

the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

b. The provisions of Control C.4.0.3 are not applicable.

l SURVEILIANCE REOUIREMENTS S.6.2.1 Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in Part 8 of the ODCM at least once per 31 days.

l i

j X-A.6-8 ODCM Rev. 18

l BASES l

This Control is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix I to 10 CFR Part 50. The Control implements the guides set forth l in Section II.A of Appendix I. The ACTION statements provide the required operating l flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept as low as is reasonably achievable. The dose calculation methodology and parameters in the ODCM implement the requirements in Section III. A of Appendix I that conforarance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid

, effluents are consistent with the methodology provided in Regulatory Cu.ide 1.109, l

" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for I the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113, " Estimating Aquatic Dispersion of Effluents l from Accidental and Routine Reactor Releases for the Purpose of Implementing l

Appendix I," April 1977.

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X-A.6-9 ODCM Rev. 18

6.3 Lf 0UID RADWASTE TREATMENT JYSTEM CONTROLS I

l C.6.3.1 The Liquid Radwaste Treatment System shall be OPERABLE and appropriate portions of the system shall be used to reduce releases of radioactivity when the projected doses dut to the liquid effluent to UNRESTRICTED AREAS (see Technical l Specification Figure 5.1-3) would exceed 0.06 mrem to the whole body or 0.2 mrem to l any organ in a 31-day period.

l l l APPLICABILITY: At all times.

l ACTION:

a. With radioactive liquid waste being discharged without treatment atid in excess l of the above limits and any portion of the Liquid Radwaste Treatment System l which could reduce the radioactive liquid waste discharged not in operation,

prepare and submit to the Commission within 30 days, pursuant to Specification l 6.8.2, a Special Report that includes the following information:

l l 1. Explanation of why liquid radwaste was being discharged without I

treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability,

2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and i

l

(

3. Summary description of action (s) taken to prevent a recurrence.

t

b. The provisions of Control C.4.0.3 are not applicable.

SURVEILIANCE REQUIREMENTS i

i S.6.3.1 Doses due to liquid releases from each unit to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the methodology and parameters in Part B of the ODCM when Liquid Radwaste Trestment Systems are not being fully utilized.

S.6.3.2 The installed Liquid Radwaste Treatment System shall be considered OPERABLE by meeting Controls C 6.1.1 and C.6.2.1.

l X-A.6-10 ODCM Rev. 18 l

BASES The OPERABILITY of the Liquid Radwaste Treatment System ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept as low as is reasonably achievable. This specification implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the Liquid Radwaste Treatment System were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix A to 10 CFR Part 50 for liquid effluents.

I

(

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X-A.6-11 ODCM Rev. 18

7.0. RADI0 ACTIVE GASEOUS EFFLUENTS 7.1 DOSE RATE QQNTROLS

, C.7.1.1 The dose rate due to radioactive materials released in gaseous effluents l from the site to areas at and beyond the SITE BOUNDARY (see Technical Specification

! Figure 5.1-1) shall be limited to the following:

a. For noble gases: Less than or equal to 500 aress/yr to the whole body and less than or equal to 3000 arems/yr to the skin, and

b. For Iodine-131, for Iodine-133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mress/yr to any organ.

l APPLICABILITY: At all times.

l &GIlDE: )

)

With the dose rate (s) exceeding the above limits, decrease the release rate within 15 minutes'to within the above limit (s).

SURVEILIANCE REOUTREMENTS S.7.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in Part B of the ODCM.

S.7.1.2 The dose rate due to Iodine-131, Iodine-133, tritium, and all j radionuclides in particulate form with half-lives greater than 8 days in gaseous j effluents shall be determined to be within the above limits in accordance with the '

-methodology and parameters in the ODCM by obtaining representative samples and ,

performing analyses in accordance with'the sampling and analysis program specified j in Table A.7.1-1.

X-A.7-1 ODCM Rev. 18 l

BASES l

l This Control is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual i dose limits of 10 CFR Part 20 to UNRESTRICTED AREAS. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column I. These limits provide reasonable assurance that radioactive material l discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE l PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to i annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106[L)). For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of the MEMBER OF THE PUBLIC will l usually be sufficiently low to compensate for any increase in the atmospheric l diffusion factor above that for the SITE BOUNDARY. Examples of calculations for l such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in the ODCM. The specified release rate limits restrict, at all times,,the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mress/ year to the whole body or to less than or equal to 3000 mress/ year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500

, mrems/ year.

t l

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I X-A.7-2 ODCM Rev. 18

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I TABLE A.7.1-1 RADI0 ACTIVE GASEOUS WASTE SAMPLING ;

AND ANALYSIS PROGRAM ;

! (Continued) l l

, Notations l

l (H The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample ' chat will yield a net count, above system background, that will be detected with 95 percent probability with only 5 percent probability of falsely concluding that a blank observation represents a "real" signal.

l l For a particular measurement system, which may include radiochemical L separation:

l LLD - 4.66 s*

E x V x 2.22 x 10' x Y x exp (-Aat)

Where: 1 i

LLD - the "a priori" lower limit of detection (microcurie per unit mass or volume),

s3 - the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute),

E -

the counting efficiency (counts per disintegration),

V - the sample size (units of mass or volume),

2.22 x los - the number of disintegrations per minute psr microcurie, !

Y -

the fractional radiochemical yield, when applicable, A - the radioactive decay constant for the particular radionuclide

( s'1) , and at - the elapsed time netween the midpoint of sample collection and the time of counting (s).

Typical values of E. V, Y, and at should be used in the calculation.

It should be recognized that the LLD is defined as an A oriori (before the fact) limit representing the capability of a measurement system and not as an A nosteriori (after the fact) limit for a particular measurement.

X-A.7-5 ODCM Rev. 18 i

~

l TABLE A.7.1-1 RADIOACTI /E CASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM I (Continued) l Notations (Continued) 1 H) The principal gamma emitters for which the LLD specification applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble ge: releases and Mn-54, Fe-59, Co 58, C -602 Zn-65, Mo-99, l I-131, Cs-134, Cs-137, Co-141 and Ce-144 in iodine and particulate releases.

l This list does not mean that only these nuclides are to be considered. Other 4 gamma peaks that are identifiable, together with those of the abovs nuclides, shall also be analyzed and reported in the Annual Radioactive Effluent Release Report in accordance with Technical Specification 6.8.1.4 and Part A, Section 10.2 of the ODCM. Isotopes which are not detected should be reported as "not j detected." Values determined to be below detectable levels are not used in ;

dose calculations.

(3) Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER within a one hour peried unless; 1) analysis shows that the DOSE EQUIVALENT I-131 concentrations in the primary coolant has not increased more than a factor of 3; 2) the noble gas activity monitor for the plant vent has not increased by more than a factor of 3. For containment purge, requirements apply only when purge is in operation.

3) Tritium grab samples shall be taken at least onca per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the refueling canal is flooded.

(s) The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Controls C.7.1.1, C.7.2.1, and C.7.3.1.

(8) Samples shall be Changed at least once per seven (7) days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing, or after removal from sampler.

Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least seven (7) days following each shutdown, startup, or THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER within a one-hour period and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months !

are analyzed, the corresponding LLDs may be increased by a factor of 10. This i requirement does not apply if 1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the reactor coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.

i 03 Samples shall be taken prior to start-up of condenser air removal system when there have been indications of a primary to secondary leak.

")

Quarterly composite analysis requirements shall only be required when analysis '

for principal gamma emitters indicate positive radioactivity.

I l

l X A.7-6 ODCM Rev. 18

1 7.2 DOSE - NOBLE CASES CONTROLS C.7.2.1 The air dose due to noble gases released in gaseous effluents to areas at and beyond the SITE BOUNDARY (see Technical Specification Figure 5.1-1) sball be limited to the following:

a. During any calendar quarter: Less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radiation, and

b. During any calendar year: Less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.

l APPLICABILITY: At all times.

l ACTION:

a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.8.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

b. The provisions of Control C.4.0.3 are not applicable.

SURVEILIANCE REOUIREMENTS S.7.2.1 Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters in Part B of the ODCM at least once per 31 days.

l l

l X A.7-7 ODCM Rev. 18 1

i

l l BASES l

This Control is provided to implement the requirements of Sections II.B. III.A, and IV. A of Appendix I to 10 CFR Part 50. The Control implements the guides set forth l

in Section I.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I at the SITE BOUNDARY that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept as low as reasonably achievable. The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109,

" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1.977, and Regulatory Guide 1.111. " Methods for Estimating Atmospheric ;

Transport and Dispersion of Caseous Effluents in Routine Releases from Light-Water l Cooled Reactors," Revision 1, July 1977. The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions.

l X-A.7-8 ODCM Rev. 18

i l

l 7.3 DOSE - IODINE-331. 10 DINE-133. TRITIUM. AND RADIOACTIVE MATERIAL IN PARTICUIATE FORM CONTROLS C.7.3.1 The dose to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, l and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released to areas at and beyond the SITE BOUNDARY (see Technical Specification Figure 5.1-1) shall be limited to the following:

a. During any calendar quarter: Lass than or equal to 7.5 mrems to any organ, and

b. During any calendar year: Less than or equal to 15 areas to any organ.

l APPLICABILITY: At all times. l l ACTION: ,

a. With the calculated dose from the release of Iodine-131, Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in j gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.8.2, a !

Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

b. The provisions of Control C.4.0.3 are not applicable.

SURVEILIANCE REOUIREMENTS S.7.3.1 Cumulative dose contributions for the current calendar quarter and current calendar year for Iodine-131, Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters in Parr B of the ODCM at least once per 31 days.

l t

X-A.7-9 ODCM Rev. 18

i i

l BASES l

This control is provided to implement the requirements of Sections II.C. III.A, and IV.A of Appendix I to 10 CFR Part 50. The Controls are the guides set forth in ;

Section II.C of Appendix I. The ACTION statements provide the required operating i flexibility and at the same time implement the guides set forth in Section IV.A of l Appendix I to assure that the releases of radioactive materials in gaseous effluents at the SITE BOUNDARY will be kept as low as reasonably achievable. The ODCM calculation methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109,

" Calculation of Annual Doses to Man from Routine Releares of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977, and Regulatory Guide 1.111. " Methods for Estimating Atmospheric Transport and Dispersion of Caseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical Iodine-133, tritium, and 3 radionuclides in particulate form with half-lives greater than 8 days are dependent I upon the existing radionuclide pathways to man in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of the calculations were (1) individual inhalation of airborne radionuclides, (2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition of radionuclides onto grassy areas where milk animals and meat-producing animals graze followed by human consumption of that milk and meat, and (4) deposition of radionuclides on the ground followed by subsequent human exposure.

l 1

l I

X-A.7-10 ODCM Rev. 18 4

f 7.4 CASEOUS RADUASTE TREATMENT SYSTEM CONTROLS C.7.4.1 The VENTI 1ATION EXHAUST TREATMENT SYSTEM and the CASEOUS RADWASTE TREATMENT SYSTEM shall be OPERABLE and appropriate portions of these system shall be used to reduce releases of radioactivity when the projected doses in 31 days due to gaseous effluent releases to areas at and beyond the SITE BOUNDARY (see Technical Specification Figure 5.1-1) would exceed

a. 0.2 mrad to air from gamma radiation, or l b. 0.4 mrad to air from beta radiation, or i

c. 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.

l APPLICABILITY: At all times. j l ACTION:

a. With radioactive gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.8.2, a Special Report that includes the following information:

1. Identification of any inoperable equipment or subsystems, and the reason for the inoperability,

2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and ,

l

3. Summary description of action (s) taken to prevent a recurrence.

l b. The provisions of Control C.4.0.3 are not applicable.

SURVEILIANCE REOUIREMENTS S.7.4.1 Doses due to gaseous releases from each unit to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days in accordance with the methodology and parameters in Part B of the ODCM when Gaseous Radwaste Treatment Systems are not being fully utilized.

S.7.4.2 The installed VENTIIATION EXHAUST TREATMENT SYSTEM and GASEOUS RADWASTE TREATMENT SYSTEM shall be considered OPERABLE by meeting Controls C.7.1.1, and C.7.2.1, or C.7.3.1. i r

X-A.7-Il ODCM Rev. 18 t

i BASES The OPERABILITY of the GASEOUS RADVASTE TREATMENT SYSTEM and the VENTI 1ATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept as low as is reasonably achievable. This Control implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I to 10 CFR Part 50, for gaseous effluents.

l 1

X-A.7-12 ODCM Rev. 18

m 8.0 TOTAL DOSE CONTROL C.8.1.1 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 areas to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems.

l APPLICABILITY: At all times.

l ACTION:

a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Controls C.6.2.1.a.

C.6.2.1.b, C.7.2.1.a C.7.2.1.b, C.7.3.1.a. or C.7.3.1.b, calculations shall be made including direct radiation contributions from the units and from outside storage tanks to determine whether the above limits of Control C.8.1.1 have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.8.2, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR 20.405(c), shall include an analysis that estimates the radiation exposure (dose) to a liEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release (s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose (s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

b. The provisions of Control C.4.0.3 are not applicable.

SURVEILIANCE REOUIREMENTS S.8.1.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Surveillance Requirement S.6.2.1, S.7.2.1, and S.7.3.1, and in accordance with the methodology and parameters in Part B of the ODCM.

S.8.1.2 Cumulative dose contributions from direct radiation from the units and from radwaste storage tanks shall be determined in accordance with the methodology and parameters in Part B of the ODCM. This requirement is applicable only under conditions set forth in ACTION a. of Control C.8.1.1.

X-A.8-1 ODCM Rev. 18

BASES This Control is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46FR18525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses due to releases of radioactivity and to radiation from uranium fuel cycle sources exceed 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mress. For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the units (including outside storage tanks, etc.) are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site are within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUiLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed.

The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Controls C.6.1.1 and C.7.1.1. An individual is not considered a MEMBER OF THE PUBLIC durin5 any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle, h

)

X-A.8 2 ODCM Rev. 18

[

l 9.0 RADIOLDCICAL ENVIRONMENTAL MONITORING j 9.1 MONITORING PROGRAM ,

CONTROL l

I C.9.1.1 The Radiological Environmental Monitoring Program (REMP) shall be

( conducted as specified in Table A.9.1-1.

l APPLICAEILITY: At all times.

l ACTION:

a. With the REMP not being conducted as specified in Table A.9.1-1, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Technical Specification 6.8.1.3 and Part A, Section 10.1 of

the ODCM, a description of the reasons for not conducting the program as I

required and the plans for preventing a recurrence.

b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceedin5 the reporting j levels of Table.A.9.1-3 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days from receipt of the laboratory analyses, pursuant to Technical Specification 6.8.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce radioactive affluents so that the potential annual dose

to a MEMBER OF ZHE PUBLIC is less than the calendar year limits of Control C.6.2.1, C.7.2.1, or C.7.3.1. When more than one of the radionuclides in the REMP are detected in the sampling medium, this report shall be submitted if concentration (1) concentration (2) reporting level (1) + reporting level (2) + .. 2 1.0 When radionuclides other than those listed in the REMP are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose

to a MEMBER OF THE FUBLIC from all radionuclides is equal to or greater than the calendar year limits of Control C.6.2.1, C.7.2.1, or C.7.3.1.

This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report required by Technical Specification 6.8.1.3 and Part A, Section 10.1 of l the ODCM.

1 I

l

The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.

X A.9-1 ODCM Rev. 18

l ACTION: (Centinusd)

With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by the REMP, identify specific locations for obtaining replacement samples and add them within 30 days to the REMP given in

(' the ODCM. The specific locations from which samples were unavailable may then l be deleted from the monitoring program. Pursuant to Technical Specification

! 6.13, and Part A, Section 10.2, of the ODCM, submit in the next Annual l Radioactive Effluent Release Report documentation for a change in the ODCM l including a revised figure (s) and table for the ODCM reflecting the new location (s) with supporting information identifying the cause of the unavailability of samples and justifying the selection of the new locations (s) for obtaining samples.

c. The provisions of Control C.4.0.3 are not applicable.

SURVEILIANCE RROUTREMENTS S.9.1.1 The radiological environmental monitoring samples shall be collected pursuant to Table A.9.1-1 from the specific locations given in the table and figure (s) in Part 8 of the ODCM, and shall be analyzed pursuant to the requirements of Table A.9.1-1 and the detection capabilities requir3d by Table A.9.1-2.

RASES The REMP required by this Control provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting from the plant operation. This monitoring program implements Section

~

IV.B.2 of Appendix I to 10 CPR Part 50, and thereby supplements the REMP by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Cuidance for this monitoring -

program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1, November 1979. The initially specified monitoring program will be effective for at least the first 3 years of commercial operation. Following this period, program changes may be initiated based on operational experience.

Detailed discussion of the LLD and other detection limits can be found in Currie, L.A., " Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements," NUREG/CR-4007 (September 1984).

! 1 X A.9-2 ODCM Rev. 18 j

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e sr/

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P n A rs C oa N bC O r I

ir T Ao C

E T

E D

)

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5 5 8

t 1 a

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(

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n r - n e o n r - s s a A C H M F c Z Z I C C B

TABLE A 9.1-2 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS (Continued)

Table Notations

a. This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report.

b. The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background,_that will be detected with 951 probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system, which may include radiochemical separation:

g, 4.66 s b l

E V 2.22 Y exp (-Aat)

Where:

LLD is the "a priori" lower limit of detection as defined above, as picoeuries per unit mass or volume; 4.66 is a constant derived from the K,1g, and Kg s. values for the 95%

confidence level; sb is the standard deviation of the backg':ound counting rate or of the counting rate of a blank sample as appropriate, as counts per minute; E is the counting efficiency, as counts per disintegration:

V is the sample size in units of mass or volume:

2.22 is the number of disintegrations per minute per picoeurie; Y is the fractional radiochemical yield, when applicable; A is the radioactive decay constant for the particular radionuclide as per second; and at for environmental samples is the elapsed time between sample collection and time of counting, as seconds.

Typical values of E, V, Y, and at should be used in the calculation.

In calculating the LLD for a radionuclide determined by gamma ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples (e.g., Potassium-40 in milk samples), i l

X-A.9-8 ODCM Rev. 18 i

TABLE A.9.1-2 l

DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS (Continued) 1 i Table Notations (Continued)

It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement. This does not preclude the calculation of an a posteriori LLD for a particular measurement based upon the actual parameters for the sample in question and appropriate decay correction parameters such as decay while sampling and during analysis. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report per Part A, Secion 10.1.

c. Parent only,

d. The Ba 140 LLD and concentration can be determined by the analysis of its short-lived daughter product La 140 subsequent to an eight-day period following collection. The calculation shall be predicated on the normal ingrowth equations for a parent-daughter situation and the assumption that any unsupported La 140 in the sample would have decayed to an insignificant amount (at least 3.6% of its original value). The ingrowth equations will assume that the supported La-140 activity at the time of collection is zero.

e. Broad leaf vegetation only.

f. If the measured concentration minus the three standard deviation uncertainty is found to exceed the specified LLD, the sample does not have to be analyzed to meet the specified LLD.

g. Required detection capabilities for thermoluminescent dosimeters used for environmental measurements shall be in accordance with recommendations of Regulatory Guide 4.13. Revision 1, July 1977.

i l

l X-A.9-9 ODCM Rev. 18

ll 1 iI 1

ilJ 8

1 s) .

t t v ce e d

uw

0 0 R o ,

0 0 rg 0 0, 0, Pk 0 M 1 2 C

/ 1 D d1 O oC op F(

S M

P M )

A g S kk

l/ 3 0 0 0 '

I A

iL 6 7 0 T MCp 3 E

F

(

e ma n

n I

V s)

N et E t e daw 0 0 0 0 0 0 0 N nr 0 0 0 0 0 0 0 ab ,

I eg 0, 0, 0, 0, 0, 0, 0, S htk 0 0 0 0 0 1 2 N sr/ 3 1 3 1 2 O i ei I FvCnp T I(

3 A R

T 1ECN r o

0 1

9.

A N 9 O e)

C t t A-MY ae B l w X AT u I c ,

TV I i g tk 9 0 0 T r/ 1 2 C ai 0 A PC Q p I

D e(

n A

R rs oa bG R r D i F A R

Y F

V .

) y M rgk 0 0 0 0 0 0

0 0 0

l 0 0 0 0 0 0 0 0 3 5 0 n G e/ 0, 0, 0 0 o N t 1 4 0, 3 3 4 1 2

i a C n i Wp 0 3

1 1 o

n ( i t

K a K t R e g

e y.v lf na oel 0

s 5 4 t i 9 1 nd s , - 4 7 - ea y 4 9 8 0 5 b 1 3 3 a ro l 5 5 5 6 6 N 3 1 1 L ar a 3 - - - - - - 1 - - - PB n - n e o o n r - s s a A H M F C c Z Z I C C B

lll

9.2 IAND.USE CENSUS CONTROL C.9.2.1 A Land Use Census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sectors of.

the nearest milk animal, the nearest residence, and the nearest garden ** of greater

!~

than 50 m2 -(500 fez ) producing broad leaf vegetation.

l APPLICABILITY: At all times.

l ACTION

a. With a Land Use Census identifying a location (s) that yields a calculated dose or dose commitment greater than the values currently being calculated in Surveillance S.7.3.1 pursuant to Technical Specification 6.8.1.4 and Part A.

Section 10.2, of the ODCM, identify the new location (s) in the next Annual Radioactive Effluent Release Report,

b. With a Land Use Census identifying a location (s) that yields a calculated dose or dose' commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordance with Control C.9.1.1, add the new location (s) within 30 days to the REMP given in the ODCM, if permission from the owner to collect samples can be obtained and sufficient sample volume is available. The sampling location (s), excluding the Control station location, having the lowest calculated dose or dose commitment (s), via the same exposure pathway, may be deleted from this Lmonitoring program after October 31 of the year in which this Land Use Census was conducted. Pursuant to Technical Specification 6.13 and Part A. Section 10.2 of the ODCM, submit in the next Annual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure (s) and table (s) for the ODCM reflecting the new location (s) with information supporting the change in sampling locations.

l c. The provisions of Control C.4.0.3 are not applicable. !

SURVEILIANCE REOUTREMENTS l

S.9.2.1 The Land Use Census shall be conducted during the growing season at i least once per 12 months using a method such as by a door-to door survey, aerial ;

survey, or by consulting local agriculture authorities, as described in the ODCM. .

The results of the Land Use Census shall be included in the Annual Radiological l Environmental Operating Report pursuant to Part A, Section 10.1 of the ODCM.

- Broad leaf vegetation sampling of at least three different kinds of vegetation may be performed at the SITE BOUNDARY in each of two different direction sectors with

( the highest predicted relative deposition values (D/Qs) in lieu of the garden census. Specifications for broad leaf vegetation sampling in the REMP shall be followed, including analysis of control samples.

X A.9-11 ODCM Rev. 18

l l

l BASES 1

l This specification is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the REMP given in the DDCM are made if required by the results of this census. Information from j methods such as the door-to-door survey, from aerial survey, of from consulting with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater then 50 m 2provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored, since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were made: (1) 20% of the garden was used for growing broad-leaf vegetation (i.e. , similar to lettuce and cabbage), and (2) there was a vegetation yield of 2 kg/m 2, l

lI i

I i

l X A.9-12 ODCM Rev. 18

1 9.3 INTERIABORATORY COMPARISON PROGRAM CONTROL C.9.3.1 In accordance with Technical Specification 6.7.6.h.3, analyses shall be performed on all radioactive materials supplied as part of an Interlaboratory ,

Comparison Program, that has been approved by the Commission, that correspond to '

samples required by REMP.

l APPLICABILITY: At all times.

l ACTION:

a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to Part A, Section 10.1 of the ODCM.

l b .- The provisions of Control C.4.0.3 are not applicable.

SURVEILIANCE REOUIREMENTS S.9.3.1 The Interlaboratory Comparison Program shall be identified in Part B of the ODCM. A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to Part A, Section 10.1 of the ODCM. ;

BASES The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the Quality Assurance Program for environmental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50.

l X-A.9-13 ODCM Rev. 18

I o

'10.0 REPORTS

, 10.1 ANNUAL RADIGIhGICAL ENVIRONMENTAL OPERATING REPORT Routine Annual Radiological Environmental Operating Reports covering the

! operation of the. station during the previous calendar year'shall be submitted prior l to May.1 of each. year pursuant to Technical Specification 6.8.1.3.

The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental Surveillance activities for t.he report period, including a comparison'with preoperational studies, with operational Controls, as appropriate,-

and with previous environmental Surveillance reports, and an assessment of the observed impacts of the plant operation on the environment. 'The reports'shall also

~

include the results of the Land Use Census required by Control C.9.2.1.

TheAnnualRadiologicalEnvironmentalOperatingReportsshallincibdethe results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in Part 5 of the ODCM, as well as summarized and tabulated results .of these analyses and measurements in the~ format of

.the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for

inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.

The reports shall also include the following: a summary description of the Radiological Environmental Monitoring Program; at least two legible maps ****

covering all sampling locations keyed to a table giving distances and directions .

from the centerline of one reactor; the results of licensee participation in the Interlaboratory Comparison Program and the corrective action taken if the specified program is not being performed as required by Control C.9.3.1; reason for not conducting the Radiological Environmental Monitoring Program as required by Control C.9.1.1,' and discussion of all deviations from the sampling schedule; discussion of environmental sample measurements that exceed the reporting levels but are not the

" result of plant effluents, pursuant to ACTION b of Control C.9.1.1; and discussion of all analyses'in which the LLD required was not achievable.

i I

l l

L ****0ne map shall cover locations near the SITE BOUNDARY; the more distant locations shall be covered by one or more additional maps.

f.

X-A.10-1 ODCM Rev. 18

l 10. 2 ANNUAL RADIOACTfvE EFFLUENT RELEASE REPORT A routine Annual Radioactive Effluent Release Report covering the operation of the station during the previous calendar year of operation shall be submitted by May 1 of each year, pursuant to Technical Specification 6.8.1.4.

The Annual Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and. solid waste released from the station as outlined in Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants,"

Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof. For solid wastes, the format for Table 3 in Appendix B shall be supplemented with three additional categories: class of solid wastes (as defined by 10 CFR Part 61), type of container (e.g. , LSA, Type A, Type B, Large Quantity) and SOLIDIFICATION agent or absorbent (e.g., cement).

The Annual Radioactive Effluent Release Report shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability ***** This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY Technical Specification (Figure 5.1-3) during the report period. All assumptions used in making these assessments, i.e., specific activity, exposure time, and location, shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the methodology and parameters, in the OFFSITE DOSE CALCUIATION MANUAL (ODCM) .

The Annual Radioactive Effluent Release Report shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190, " Environmental Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculating the dose contribution from liquid and gaseous affluents are given in Regulatory Guide 1.109, Rev. 1, October 1977.

The Annual Radioactive Effluent Release Report shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

- - - - In lieu of submission with the Annual Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.

X-A.10-2 ODCM Rev. 18

s l The Annusi Radioectiva Efflusnt Reissze Report chn11 includs any changes mada during the reporting period to the PROCESS CONTROL PROGRAM and the ODCM, pursuant to l Technical Specifications 6.12 and 6.13, respectively, as well as any major change to i

Liquid, Caseous, or Solid Radwaste Treatment Systems pursuant to Control 11.0. It l shall also include a listing of new locations for dose calculations and/or

( environmental monitoring identified by the Land Use Census pursuant to Control l l C.9.2.

The Annual Radioactive Effluent Release Report shall also include the following: an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Control l C.5.1 or C.S.2, respectively; and description of the events leading to liquid holdup tanks or gas storage tanks exceeding the limits of Technical Specification 3.11.1.4 l

X-A.10 3 ODCM Rev. 18

l l

t l

I SEABROOK STATION ODCM PART 5 RADIOLOGICAL CALCUIATIONAL METHODS AND PARAMETERS l

i I

1 I

l 1

I I

l l

X-B.1-0 ODCM Rev. 18

)

1.0 VNTRODUCTION Part B of the ODCM (Off-Site Dose Calculation Manual) provides formal and approved methods for the calculation of off-site concentration, off-site doses and effluent monitor setpoints, and indicates the locations of environmental monitoring stations in order to comply with the Seabrook Station Radiological Effluent Controls Program (RECP), and Radiological Environmental Monitoring Program (REMP) detailed in Part A of the manual. The ODCM forms the basis for station procedures which document the off-site doses due to station operation which are used to show compliance with the numerical guides for design objectives of Section II of Appendix I to 10CFR Part 50. The methods contained herein follow accepted NRC guidance, j unless otherwise noted in the text.

The references to 10 CFR Part 20 in Part 8 of the ODCM refer to revisions of I 10 CFR Part 20 published prior to 1 January 1993. The decision to continue the use of the "old" version of 10 CFR Part 20 is based on an NRC letter dated June 30, 1903, from Thomas'E. Murley to Thomas E. Tipton. For the convenience of the plant staff a copy of 10 CFR Part 20 (Rev.1 January 1992) has been included in Appendix B.

1.1 RUPONSIBILITIES FOR PART B All changes to the ODCM shall be reviewed and approved by the Station Operation Review Committee (SORC), approved by the Station Director, and documented in accordance with Technical Specification 6.13. Changes made to Part B shall be subr.itted to the Commission for their information in the Annual Radioactive Effluent Release Reoort for the period in which the change (s) was made effective.

It shall be the responsibility of the Station Director to ensure that the ODCM is used in the performance of surveillance requirements and administrative controls in accordance with Technical Specifications 6.7.6.g and 6.7.6.h, and Efflu(nt Control Program and Radiolo5i cal Environmental Monitoring Program detailed in Part A of the manual.

In addition to off-site dose calculations for the demonstration of compliance with Technical Specification dose limits at and beyond the site boundary, l 10CFR20.1302 requires that compliance with the dose limits for individual members of 4 the public (100 arem/yr total effective dose equivalent) be demonstrated in I controlled areas on-site. Demonstration of compliance with the dose limits to i members of the public in controlled areas is implemented per Health Physics Department Procedures, and is outside the scope of the ODCM. However, calculations performed in accordance with the ODCM can be used as one indicator of the need to perform an assessment of exposure to members of the public within the site boundary.

Since external direct exposure pathways are already subject to routine exposure rate surveys and measurements, only the inhalation pathway need be assessed. The l accumulated critical organ dose at the site boundary, as calculated per ODCM Part B Sections 3.9 and 3.11, can be used as an indicator of when additional assessments of on-site exposure to members of the public is advisable (see Section 3.11.2).

Off-site critical organ den s from station effluents should not, however, be the only indicator of potential on-site doses.

X-B.1-1 ODCM Rev. 18

13

SUMMARY

OF METHODS, DOSE FACTORS, LIMITS, CONSTANTS, VARIABLES AND .

l DEFINITIONS This section summarizes the Method I dose equations which are used as the i l primary means of demonstrating compliance with RECP. The concentration and setpoint l methods are identified in Table B.1-2 through Table 5.1-7. Appendix C provides l documentation for an alternate computerized option, designated as Method IA in the l ODCM, for calculating doses necessary to demonstrate compliance with RECP. The Effluent Management System (EMS) software package used for this purpose is provided by Canberra Industries, Inc. Where more refined dose calculations are needed, the use of Method II dose determinations are described in Sections 3.2 through 3.9 and 3.11. The dose factors used in the equations are in Tables B.1-10 through B.1-14 and the Regulatory Limits are summarized in Table B.1-1.

The variables and special definitions used in this ODCM, Part 8, are in Tables B.1 8 and B.1 9.

i l

I X-B.1-2 ODCM Rev. 18 i

8 1

r .

o .

v t r . . n o e l

q y r r n R m

t y a a a q a . .

/ a n . r r M i

C n n a i n r y y C _

p i i n i y / / D _

t n i a / m n O i

a a i e n a e e m - e e a r e e r r i 0 r r a e m r r a m L 1 a m e r r m 0 x 5 0 r

a m 6 0 2 0 0

0 0

0 S 0 0 0 5 N 1 2 1 3 5 1 0 0 5 3 1 O s 5 $ s $ s 5 s $ s s I

T A

U O

E _

G N

I T _

N _

=

E M

I I

)

1

( 1 2 1 2 1 2 3 4 5 P I - - - - - - - - -

M 2 2 3 3 3 3 3 3 3 I

d D o . . . . . . . . .

h q q q q q q q q q N t E E E E E E E E E A e 3 M 1

O R

T 1N

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1. C 3 BA e -

e t 1 ET l a ,

B LR fb R 3 h BA oo s s 3 t X

AP N s e e ee e e1 i T n og a s s ss es t - w T G n o o oa t a aI N v i n o D D DC aG R ss E r ti ei R ,dey J o cd l t y e y e ye e e1 nt a l z au b a d s d s dl el s3 aaD I

F e rl or o o o o ob sb o1 l F t a

Fc x

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B D B D B o oo D - mu8 E N DN I uc C lEs l e l n l n l n ii>

L a tCs e ac t n t a a g t a a g am t o nm i o amt goir 2 t _

A C oP a oo o r o r or kr rrra /

1 I TMC TC T O T O Tf Sf OfTPT G

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S l f a f ati E e o E r E R nl t r t ei sa _

t d n d d mh uR -

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r a

P 1 1 1 1 1 2 3 1 6 6 6 7 C C C C l l l l l l

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r o h B LR n f r mt -

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T s e rt o n N y o s fi ss D e i E r Ds os rey s o o e De eTt a e y e o p D

I m rs s s aD s d s D t F e ia ra o ,l o o o e F t AG iC Dsu8 D B D d S E a A ec i C ae e nni> n l n o m L ml al ait a a a r r A mb tb gd r 2 g t g y a C ao eo roa /

1 r o r h l GN BN O T O T I OIPT A C

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hn h q q ) ti s N

A t E E s sy e r e

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t l e C ue n 5

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I a rom i

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TABLE B.1-2

SUMMARY

OF METHOD I EOUATIONS TO CALCUIATE UNRESTRICTED AREA LIOUID CONCENTRATIONS Equation Number Category Equation 2-1 Total Fraction of MPC in Liquids, Except Noble Cases f.{ s1 2-2 Total Activity of Dissolved and Entrained Noble Cases ( * = { Cf from all Station Sources CP 19Ci 7

i l ~ ' s 2E-04 l

l l

X-B.1-6 ODCM Rev. 18

TABLE B.1-3 l

SUMMARY

OF METHOD I EOUATIONS TO CALCUIATE 0FF-SITE DOSES F1t0M LIOUID RFf FASES Equation Nn=her Categorv Eauntion 31 Total Body Dose D tb(area) - k E Q g DE itb i

3-2 Maximum Organ Dose D,,(area) - k Qg DEg ,,

l l

l X B.1-7 ODCM Rev. 18

8 1

v e

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D B c F B F o n D F

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r H R ro"n o T F o f

E L t i

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F U ea O O E E r O DI ce co e t

R L n Y V e R ID C A

M N n e M I r U N or u S ie a b c d e f t A tb 8 8 8 8 8 8 a

_ O am uu 3 3 3 3 3 3 N T qN E

E S e O c D nd en

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y ai - _

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TABLE B.1-7

SUMMARY

OF METHODS FOR

SETPOINT DETERMINATIONS.

Equation Nu=her Catamorv Eaumtion

$-1 Liauid Effluents:

Liquid Waste Test E I Tank Monitor Rsetpoint ( "1N )" fl F X DF"I" E Cyi (RM 6509) l 5-23 PCCW Rate of- -8 l Change Alarm RC,,g(gph) - lx10 SWF -

Caseous Effluents:

Plant Vent Wide Range Gas Monitors (RM 6528-1, 2, 3) 55 Total Body Rtb (pCi/sec) - 588 DFB 56 Skin 1 R

skin (pci/sec) - 3000 DF' I

I X-B,1-14 ODCM Rev. 18

3 TABLE B.1-8

SUMMARY

OF VARfABLES l

Variable Sefinition Units )

- .; at point of discharge and pCi/mi )

CE able gas "i" in liquid pathways etion sources

- fo6a1 ac ,;y of all dissolved and pCL C{ entrainec ..oble gases in liquid pathways from all station sources 51 Cai

- Concentration of radionuclide "i" at the pCi point of liquid discharge ml Ci - Concentration of radionuclide "i" pCi/ml C pt - Concentration, exclusive of noble gases, of pCi radionuclide "i" from tank "p" at point of "1

discharge l C.,i - Concentration of radionuclide "i" in pCi/mi mixture at the monitor

- Off-site beta dose to air due to noble mrad D{trc.> gases in elevated release

- Off-site beta dose to air due to noble gas arad I D{tr<s) in ground level release )

- Beta dose to air at Science & Nature Center mrad D{trec.: due to noble gases in elevated release

- Beta dose to air at Science & Nature Center mrad Ds .rtcs) due to noble gases in ground level release l

-- Beta dose to air at " Rocks" due to noble mrad l D{ trac.> gases in elevated release l

Beta dose to air at " Rocks" due to noble trad D{tracs3 gases in ground level release

- Off-site gamma dose to air due to noble mrad D trt.) gases in elevated release

- Off-site gamma dose to air due to noble mrad D!trca) gases in ground level release

- Gamma dose to air at Science & Nature mrad Disret.) Center due to noble gases in elevated release

- Gamma dose to air at Science & Nature mrad D!trt(s) Center due to noble gases in ground level ,

release '

Gamma dose to air at " Rocks" due to noble mrad D! trac.) gases in elevated release X-B.1-15 ODCM Rev. 18

1 TABLE B.1-8

SUMMARY

OF VARIABTVM (Continued)

Variable Definition Units

- Gamma dor

to air at " Rocks" due to noble mrad Dhrats) gases in escond level release D,,g,3 - Critical organ lose from an elevated mrem release to an off-site receptor D,,c,3 - Critical organ dose from a ground level mrem j release to an off site receptor i l

D,g33 - Critical organ dose from an elevated mrem release to a receptor at the Science &

Nature Center l

D,,gg,3 - Critical organ dose from a ground level mrem release to a receptor at the Science &

Nature Center D,,ng,3 - Critical organ dose from an elevated mrem release to a receptor at the " Rocks" Deoats) - Critical organ dose from a ground level mrem release to a receptor at the " Rocks" Da

- Direct dose mrem

- Gamma dose to air, corrected for finite mrad Dhnit. cloud D., - Dose to the maximum organ mrem Ds - Dose to skin from beta and gamma arem Deb Dose to the total body mrem D F,in - Minimum required dilution factor ratio

- Composite skin dose factor for off-site arem-sec/pci-yr DF t receptor

- Composite skin dose factor for Science & mrem sec/pci-yr DFi r Nature Center

- Composite skin dose factor for the " Rocks" mrem sec/pci-yr DFin DFB i - Total body gamma dose factor for nuclide mrem 3 "i" (Table B.1 10) pCi-yr DFB, - Composite total body dose factor 3 pCL-yr X B.1-16 ODCM Rev. 18

TABLE B.1-8

$UMMARY OF VARIABLES (Continued)

Variable Definition Units DF1.its - Site-specific, total body dose factor for a liquid release of nuclide "i" (Table B.1- mrem

11) N DFLt ,. - Site specific, maximum organ dose factor for a liquid release of nuclide "i" (Table B.1 11) W DFBo.g.3 - Site-specific, critical organ dose factor arem/pci for an elevated gaseous release of nuclide "i" (Table B.1-12) 4 DFGi ..g,3 - Site-specific critical organ dose factor mrem /pCi for a ground level release of nuclide ai" ,

(Table B.1-12)

DFGi .ogg,3 - Science & Nature Centar-specific critical arem/pci organ dose factor for an elevated release of nuclide "i" (Table B. 14)

DFGi ogg,3 - Science & Nature Center-specific critical arem/pCi organ dose factor for a ground level release of nuclide "i" (Table B.1-14)

DFGugg,3 - The " Rocks"-specific critical organ dose mres/pCi factor for an elevated release of nuclide j "i" (Table B.1-15)

DFG teonte> - The " Rocks"-specific critical dose factor arem/pCi for a ground level release of nuclide "i" (Table 8.1-15)

- Site-specific critical organ dose rate DFG scot.) mrem-sec l factor for an elevated 5aseous release of pci-yr nuclide "i" (Table B.1-12)

- Site-specific critical organ dose rate i DFG uets) factor for a ground level release of mrem-see pci-yr nuclide "i" (Table 5.1-12) 1'

- Science & Nature Center-specific critical D FGt ort.) organ dose rate factor for an elevated mrem-see release of nuclide "i" (Table 5.1-14) #C1~7#

. Science & Nature Center-specific critical DFG icosts) organ dose rate factor for a ground level mrem-see release of nuclide "i" (Table B.1-14) #C1~7#

- The " Rocks"-specific critical organ dose DFG aceat.) rate factor for an elevated release of arem-sec ci-yr nuclide "i" (Table B.1-15) l X B.1-17 ODCM Rev. 18

TABLE B.1-8

SUMMARY

OF VARIABIIS (Continued)

Variable Definition lln111

- The " Rocks"-specific critical organ dose DFG aeon (s) rate factor for a ground level release of mrem-sec nuclide "1" (Table B.1-15) pGi-yr DFF t - Beta skin dose factor for nuclide "i" (Table 5.1-10) mrem-m 3 pGi-yr DF' s - Combined skin dose factor for nuclide "i" (Table B.1 10) arem-m 3 pCi-yr

- Gamma air dose factor for nuclide "i" DF (Table B.1-10) mrad-m3 pCi-yr

- Beta air dose factor for nuclide "i" DFf (Table B.1-10) mrad-m3 pGi-yr

- Critical organ dose rate to an off site D co <e) receptor due to elevated release of arem iodines, tritium, and particulates 7#

- Critical organ dose rate to an off-site D cots) "#**

receptor due to ground level release of iodines, tritium, and particulates 7#

- Critical organ dose rate to a receptor at Decate) the Science & Nature Center due to an elevated release of iodines, tritium, and Y#

particulates

- Critical organ dose rate to a receptor at D coE(s) "#*"

the Science & Nature Center due to a ground level release of iodines, tritium, and yr particulates

- Critical organ dose rate to a receptor at b eont ) "#**

the " Rocks" due to an elevated release of iodines, tritium, and particulates Y#

- Critical organ dose rate to a receptor at D cets) "#**

the " Rocks" due to a ground level release of iodines, tritium, and particulates 7#

- Skin dose rate to an off-site receptor due D sktate) "#*"

to noble gases in an elevated release yr X-B.1-18 ODCM Rev. 18 l

TABLE B.1-8

SUMMARY

OF VARIABLES (Continued)

Variable Definition Units

- Skin dose rate to an off-site receptor due D.ktn(s) to noble gases in a ground level release yr

- Skin dose rate to a receptor at the Science

& Nature Center due to noble gases in an "#**

D.htnet.)

elevated release 7#

- Skin dose rate to a receptor at the Science

& Nature Center due to noble gases in a mrem D.htnt(s) ground level release 7"

- Skin dose rate to a receptor at the " Rocks" D ektnat.) due to noble gases in an elevated release "#*"

yr

- Skin dose rate to a receptor at the " Rocks" D akinata) due to noble gases in a ground level mrem release 7#

- Total body dose rate to an off-site D ebt.) receptor due to noble gases in an elevated "#*" 4 release 7#

- Total body dose rate to an off-site l Dtb(s) receptor due to noble gases in a ground l level release 7#

- Total body dose rate to a receptor at the Dtbet.) "#**

Science & Nature Center due to noble gases in an elevated release 7#

- Total body dose rate to a receptor at the D tbt(s) "#**

Science & Nature Center due to noble gases in a ground level release 7# !

- Total body dose rate to a receptor at the l D tbRt.) " Rocks" due to noble gases in an elevated "#**

release 7#

- Total body dose rate to a receptor at the Dtbn(s) " Rocks" due to noble gases in a ground 1evel release 7#

=

D/Q Deposition factor for dry deposition of elemental radioiodines and other particulates ka F4 - actual or estimated flow rate out of gpm or ft /sec.

3 discharge tunnel X-B.1-19 ODCM Rev. 18

h

\

TABLE B.1-8

SUMMARY

OF VARIABTM (Continued)

Variable Definition Units F, -

Flow rate past liquid waste test tank gpm monitor F,,, - Maximum allowable discharge flow rate from gpm liquid test tanks F - Flow rate past plant vent monitor cc

. SeC f t ; fa; fs; -

Fraction of total MPC associated with Paths Dimensionless

f. 1, 2, 3, and 4

- Total fraction of MFC in liquid pathways Dimensionless F{" (excluding noble gases)

MPC i - Maximum permissible concentration for radionuclide "i" (100FR20, Appendix B, pCi Table 2, Column 2) cc Qi -

Release to the environment for radionuclide curies, or "i" pcuries Release rate to the environment for pCi/sec l Di radionuclide "i" l R .spoing -

Liquid monitor response for the limiting pCi/ml ;

concentration at the point of discharge l R,gio -

Response of the noble gas monitor to epm, or pCi/sec limiting total body dose rate Rtb - Response of the noble gas monitor to cpa, or pCi/sec limiting total body dose rate

Sr - Shielding factor Dimensionless S' -

Detector counting efficiency from the gas cpm monitor calibration or mR/hr pCi-cc pC1/cc S,i - Detector counting efficiency for noble gas

.t. cpm mR/hr or pCL-cc pCL/cc St -

Detector counting efficiency from the liquid monitor calibration g Sti -

Detector counting efficiency for radionuclide "i. cps pC1/ml X-B.1-20 ODCM Rev. 18

TABLE 5.1-8

SUMMARY

OF VARIABLES (Continued) ;

'j Variable Definition Units X/Q - Average long-term undepleted atmospheric dispersion factor (Tables B.7-4, 8.7-5, and sec m,

B.7-6)

[X/Q)7 - Effective long term average gamma atmospheric dispersion factor see (Tables B 7 4, B.7 5, and B.7-6) m3 SWF - Service Water System flow rate gph PCC - Primary component cooling water measured pCi/ml (decay corrected) gross radioactivity concentration t** - Unitiess factor which adjusts the value of Dimensionless atmospheric dispersion factors for elevated or ground-level releases with a total release duration of t hours 1

l 1

1 l

I l

l X-B.1-21 ODCM Rev. 18 i

TABLE B.1 9 DEFINITION OF TERMS l

l Critical Recentor - A hypothetical or real individual whose location and behavior cause him or her to rc ceive a dose greater than any other possible real individual.

Q21g As used in Figulatory Guide 1.109, the term " dose," when applied to individuals, is used instead of the more precise term " dose equivalent," as defined by the International Commission on Radiological Units and Measurements (ICRU). When applied to the evaluation of internal deposition or radioactivity, the term " dose,"

as used here, includes the prospective dose component arising from retention in the body Seyond the period of environmental exposure, i.e., the dose commitment. The i dose commitment is evaluated over a period of 50 years. The dose is measured in area to tissue or arad to air. .

Dose Rate - The rate for a specific averaging time (i.e., exposure period) of dose accumulation.

Liould Radwaste Treatment System - The components or subsystems which comprise the available treatment system as shown in Figure B.6-1, 1

l I

X-B.1 22 ODCM Rev. 18

8 3 1 r

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as { 911161111333119 C Go , D D p O e 3 s

o m- r y 343323233334323 Dr d i- 000000000000000 S r to a C r p EEEEEEEEEEEEEEE i

E 887533631859675 S Aca m g 289909081403427 A a F E

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S it p E EEEEEEEEEEEEE R kc " 9 -.6437196461623 _

O S a ( 6 4373027901821 T F -

C a s 2 1192174937114 -

A t r F e D 3 B 0 E 1 S

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ae "~ r y 383532225443333 4 t s 8 oo r " i

- 000000000000000 TDo t * ' Cp EEEEEEEEEEEEEEE 8 ayc " ( 467127665142123 851694651591848 - -

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I TABLE B.1-11 l DOSE FACTORS SPECIFIC FOR SEABROOK STATION l m LIOUID DrtrASES l

i Total Body Maximum Organ Dose Factor Dose Factor DrLm( ) DrLw ( "*g*1 )

Radionuclide H-3 3.02E-13 3.02E-13 I Na-24 1.38E-10 1.42E-10 Cr-51 1.83E-11 1.48E-09 Mn 54 5.15E-09 2.68E-08 Fe-55 1.26E-08 7.67E-08 Fe 59 8.74E-08 6.66E-07 Co-58 2.46E-09 1.40E-08 Co 60 6.15E-08 9.22E-08 Zn 65 2.73E-07 5.49E-07 Br 83 1.30E-14 1.89E-14 Rb 86 4.18E-10 6.96E-10 Sr-89 2.17E-10 7.59E-09 Sr 90 3.22E 08 1.31E-07 Nb 95 5.25E-10 1.58E-06 Mo 99 3.72E-11 2.67E-10 Tc-99m 5.22E-13 1.95E-12 Ag-110s 1.01E 08 6.40E-07 Sb 124 1.71E-09 9.89E-09 Sb 125 6.28E-09 8.31E-09 Te 127m 7.07E-08 1.81E-06 Te-127 3.53E-10 9.54E-08 Te-129m 1.54E-07 3.46E-06 Te-129 7.02E-14 1.05E-13 Te-131m 3.16E-08 2.94E-06 Te 132 9.06E 08 3.80E-06 I 130 2.75E-11 3.17E-09 I-131 2.30E-10 1.00E-07 I-132 6.28E-11 6.36E-11 1-133 3.85E 11 1.15E-08 I-134 1.19E-12 1.41E-12 I-135 5.33E-11 4.69E-10 Cs-134 3.24E-08 3.56E-08 Cs-136 2.47E 09 3.27E-09 Cs-137 3.58E-08 4.03E-08 Ba 140 1.70E-10 3.49E-09 La-140 1.07E-10 4.14E-08 Ce-141 3.85E-11 9.31E-09 Ce-144 1.96E-10 6.46E-08 l Other* 3.12E-08* 1.58E-06*

Dose factors to be used in Method I calculation for any "other" detected gamma emitting radionuclide which is not included in the above list.

{

i X-B.1-24 ODCM Rev. 18 I

J

TABLE B.1 12 DOSE AND DOSE RATE FACTORS SPECIFIC FOR SEARROOK STATION f.QR 10 DINES. TRITIUM AND PARTICUIATE RELEASES Critical Organ Critical Organ Critical Organ Dose Critical Organ j Dose Factor Dose Factor for Rate Factor for Dose Rate Factor for T.levated Ground Level Elevated Release for Ground Level Release Point Release Point Point Releas* Point Radio- pyg,,,, g mrem 3 pyg,, gmrom3 pyg' ,,,, garem-see) Drag'%, ( mrom-see 3 nuclide pei pei yr-pci yr-pci H3 3.08E-10 3.76E 09 9.71E 03 1.19E-01 Cr 51 8.28E 09 2.84E-08 2.91E-01 1.01E+00 Mn-54 1.11E-06 3.79E-06 4.38E+01 1.50E+02 Fe-59 1.06E-06 3.65E-06 3.53E+01 1.21E+02 Co 58 5.56E-07 1.91E-06 2.00E+01 6.88E+01 Co 60 1.21E 05 4.12E-05 5.42E+02 1.85E+03 Zn 65 2.33E-06 7.93E 06 7.82E+01 2.66E+02 Sr 89 1.98E 05 6.73E 05 6.24E+02 2.12E+03 Sr 90 7.21E-04 2.47E 03 2.27E+04 7.79E+04 Zr 95 1.10E 06 3.77E-06 3.63E+01 1.24E+02 Nb-95 2.01E-06 6.86E-06 6.40E+01 2.20E+02 1

Mo-99 1.63E-08 1.10E-07 5.39E-01 3.56E+00 Ru-103 3.03E-06 1.04E-05 9.62E+01 3.31E+02 i Ag 110m 5.02E-06 1.72E-05 1.80E+02 6.15E+02 Sb-124 1.83E-06 6.28E-06 6.15E+01 2.11E+02 1-131 1.47E-04 5.04E-04 4.64E+03 1.59E+04 I-133 1.45E 06 5.72E-06 4.57E+01 1.80E+02 Cs-134 5.62E 05 1.91E 04 1.81E+03 6.18E+03 Cs-137 5.47E-05 1.86E-04 1.79E+03 6.09E+03 Ba 140 1.55E 07 6.39E 07 5.01E+00 2.06E+01 i

Ce it 1 2.65E-07 9.28E-07 8.45E+00 2.96E+01 l Ce-144 6.09E-06 2.09E-05 1.93E+02 6.62E+02 Other* 4.09E-06 1.39E-05 1.29E+02 4.38E+02

Dose factors to be used in Method I calculations for any 'other" detected gamma l emitting radionuclide which is not included in the above list.

X-B.1-25 ODCM Rev. 18 l

l

I 1 l

TLJLE B.1-13 !

COKRINED SKIN DOSE RATE FACTORS SPECIFIC FOR SEABROOK STATTON i SPECIAL RECEPTORS (D FOR !

NOBLE GAS RELEASE i

Science & Nature Science & Nature Center Center The " Rocks" Combined Skin Combined Skin The " Rocks" Combined Skin Dose Rate Factor Dose Rate Factor Combined Skin Dose Rate Factor for Elevated for Dose Rate Factor for Release Ground Level for Elevated Ground Level Release Point Releaso Point Release Point Point Radio- pr,** g arem-see3 pp,** gmrem-see) py,*** (mrem-see) nuclide pei-yr pci-yr pCi-yr Dr'* ( mr em-see )

pei-yr Ar 41 1.57E-02 1.17E-01 9.73E-02 6.99E-01 Kr-83m 2.35E-05 1.13E-04 1.07E 04 5.58E-04 Kr 85m 3.84E-03 4.08E 02 3.16E-02 2.69E-01 Kr 85 2.16E 03 3.09E 02 2.29E-02 2.15E-01 j Kr-87 2.31E-02 2.60E-01 2.00E 01 1.74E+00 !

I Kr 88 2.23E-02 1.442-01 1.25E-01 8.18E-01 Kr-89 3.73E-02 3.34E 01 2.68E-01 2.12E+00 Kr 90 3.15E-02 2.64E 01 2.141-01 1.64E+00 Xe-131m 9.52E-04 1.19E-02 8.96E-03 8.07E-02 Xe 133m 1.99E-03 2.48E-02 1.87E-02 1.68E-01 Xe-133 9.20E 04 9.11E-03 7.16E-03 5.92E 02 )

l Xe-135m 5.24E 03 3.61E-02 3.07E-02 2.11E-01 Xe-135 5.32E 03 5.41E-02 4.23E-02 3.53E-01 Xe-137 2.14E-02 2.89E-01 2.16E-01 2.00E+00 Xe-138 1.78E-02 1.49E 01 1.21E 01 9.27E-01 (D See Seabrook Station Technical Specification Figure 5.1-1.

l l

i i

l 1

X-B.1-26 ODCM Rev. 18

TABLE B.1-14 DOSE AND DOSE RATE FACTORS SPECIFIC FOR THE SCIENCE & NATURE CEh~rER FOR IODINE. TRITIUM. AND PARTICUIATE RELEASES Critical Organ Critical Organ Dose Factor for Critical Organ Dose Critical Organ Dose Dose Factor for Ground Level Rate Factor for Rate Factor for Elevated Release Elevated Release Ground Level Release Release Point Point Point Point Rad ~

om,,,,,

( mrc YI D""*'

I pcYI D""'

I p -

I U"'"**

I pcI~h I H-3 6.45E-11 9.27E-10 2.03E-03 2.92E-02 Cr-51 4.98E-09 2.88E-08 2.12E 01 1.11E+00 Mn-54 1.39E 06 5.71E-06 6.24E+01 2.39E+02 Fe-59 3.09E 07 1.89E 06 1.29E+01 7.16E+01 Co 58 3.89E-07 2.10E-06 1.72E+01 8.26E+t Co 60 2.17E-05 8.03E-05 9.87E+02 3.63E+03 Zn 65 7.34E-07 3.19E-06 3.31E+01 1.33E+02 Sr 89 1.15E 07 1.61E-06 3.63E+00 5.08E+01 Sr-90 5.14E-06 7.19E 05 1.62E+02 2.27E+03 Zr-95 3.38E-07 2.57E-06 1.35E+01 9.15E+01 Nb-95 1.53E 07 9.35E 07 6.43E+00 3.53E+01 Mo-99 1.62E 08 1.92E-07 5.58E-01 6.21E+00 Ru-103 1.30E-07 8.64E 07 5.33E+00 3.19E+01 Ag-110m 3.43E-06 1.54E-05 1.55E+02 6.34E+02 Sb-124 6.96E-07 4.46E 06 2.89E+01 1.67E+02 1-131 7.79E 07 1.08E-05 2.47E+01 3.41E+02 I-133 1.84E-07 2.56E-06 5.83E+00 8.11E+01 i Cs-134 6.83E-06 2.532-05 3.08E+02 1.14E+03 Cs-137 1.03E-05 3.81E-05 4.64E+02 1.72E+03 Ba-140 1.14E-07 1.42E-06 3.85E+00 4.54E+01 Ce-141 4.09E 08 4.51E-07 1.45E+00 1.48E+01 Ce-144 6.95E-07 9.11E-06 2.27E+01 2.90E+02 Other* 2.26E-06 9.24E-06 1.02E+02 3.91E+02 Dose factors to be used in Method I calculations for any "other" detected gamma emitting radionuclide which is not included in the above list.

X-B.1-27 ODCM Rev. 18

I TABLE B.1-15 DOSE AND DOSE RATE FACTORS SPECIFIC FOR THE " ROCKS" FOR IODINE. TRITIUM. AND PARTICUIATE RELEASES Critical Organ Critical Organ Critical Organ Dose Critical Organ Dose Dose Factor Dose Factor for Rate Factor for Rate Factor for for Elevated Ground Level Elevated Release Ground Level Release Point Release Point Point Release Point Rad o Dr4. ., ( "'cy) Dr% ,,("#y) e c DFG's m ( "* *c - r ) D FG's, ,p

( " e ~,s y

H-3 6.85E-10 6.45E-09 2.16E-02 2.03E-01 Cr 51 2.68E-08 1.75E-07 1.07E+00 6.53E+00 Mn-54 5.84E-06 3.18E-05 2.55E+02 1.31E+03 Fe-59 1.74E-06 1.17E 05 6.78E+01 4.29E+02 Co-58 2.01E 06 1.25E-05 8.11E+01 4.79E+02 Co 60 8.83E-05 4.09E-04 3.97E+03 1.85E+04 Zn-65 3.23E-06 1.80E-05 1.37E+02 7.29E+02 Sr-89 1.23E-06 1.15E-05 3.88E+01 3.63E+02 Sr 90 5.48E-05 5.14E-04 1.73E+03 1.62E+04 Zr-95 2.22E-06 1.68E 05 8.14E+01 5.83E+02 Nb 95 8.59E-07 5.79E 06 3.37E+01 2.13E+02 Mo 99 1.50E-07 1.34E-06 4.92E+00 4.32E+01 Ru-103 7.74E-07 5.47E-06 2.95E+01 1.96E+02 Ag-110m 1.54E-05 8.77E-05 6.47E+02 3.53E+03 Sb-124 4.04E-06 2.80E 05 1.56E+02 1.01E+03 I-131 8.27E-06 7.73E-05 2.61E+02 2.44E+03 1-133 1.95E-06 1.83E-05 6.18E+01 5.77E+02 Cs 134 2.7BE 05 1.29E-04 1.25E+03 5.80E+03 Cs-137 4.19E-05 1.94E 04 1.89E+03 8.77E+03 Ba 140 1.10E-06 9.99E-06 3.56E+01 3.19E+02 Ce-141 3.59E 07 3.14E 06 1.20E+01 1.02E+02 Co 144 7.02E-06 6.46E-05 2.25E+02 2.05E+03 Other* 9.56E-06 5.09E 05 4.16E+02 2.12E+03

Dose factors to be used in Method I calculations for any "other" detected gamma emitting radionuclide which is not included in the above list.

I I

X-B.1-28 ODCM Rev. 18

2.0 METHOD TO CALCUIATE OFF-SITE LIOUID CONCENTRATIONS Chapter 2 contains the basis for station procedures used to demonstrate l compliance with ODCM Part A Control C.6.1.1, which limits the total fraction of MFC in liquid pathways, other than noble gases (denoted here ast F " ) at the point of discharge from the station to the environment (see Figure B.6-1). Ft N is limited to less than or equal to one, i.e.,

F $ 1. -

The total concentration of all dissolved and entrained noble gases at the point of discharge from the multiport diffuser from all station sources combined, denoted C t ", is limited to 2E-04 pCi/m1, i.e., ,

1 C " $ 2E-04 pCi/ml.

t Appendix C, Attachments 3 and 4, provide the option and bases for the use of the EMS ,

determination of liquid concentration limits for plant discharges to the environment.

2.1 METHOD TO DETERMINEt F " AND C t "

First, determine the total fraction of MFC (excluding noble gases), at the point of discharge from the station from all significant liquid sources denoted Ft N; and then separately determine the total concentration at the point of j discharge of all dissolved and entrained noble gases from all station sources, !

denoted Ct ", as follows i

F enc

- E h 5 1. (2-1) i t

p i E _MPCi uCi/ml (pCi/ml}

and:

NG NG C

1

- E Cyt 5 2E-04 (2-2) i (pCi/ml) (pCi/ml) (pCi/ml) where:

F

- Total fraction of MPC in liquids, excluding noble gn=*s, at the point of discharge from the multiport diffuser.

X B.2-1 ODCM Rev. 18

(~

2.1 METHOD TO DETERMINE3 F t m2 AND Cg") (Continu2d) l C,3 - Concentration at point of discharge from the multiport diffuser of radionuclide "i", except for dissolved and entrained noble gases, from all tanks and other significant sources, p, from which a discharge may be made (including the waste test tanks and any other significant source from which a discharge can be made).

C,i is determined by dividin ; the product of the measured radionuclide concentration in liquid waste test tanks, PCCW, steam generator blowdown, or other effluent streams times their discharge flow rate by the total available dilution water flow rate of circulating and service water at the time of release (pCi/ml).

MPC g - Maximum permissible concentration of radionuclide "i" except for dissolved and entrained noble gases from 10CFR20, Appendix B, Table II, Column 2 (pCi/al). See Appendix B for a list of MPC values.

C2 "3 - Total concentration at point of discharge of all dissolved and entrained noble gases in liquids from all station sources (pCi/ml)

Cri - Concentration at point of discharge of dissolved and entrained noble gas "1" in liquids from all station sources (pci/ml) 2.2 METHOD TO DETERMINE RADIONUCLIDE CONCENTRATION FOR EACH LIQUID EFFLUENT SOURCE 2.2.1 Waste Test Tanks C,3 is determined for each radionuclide detected from the activity in a representative grab sample of any of the waste test tanks and the predicted flow at the point of discharge.

The batch taleases are normally made from two 25,000-gallon capacity waste test tanks. These tanks normally hold liquid waste evaporator distillate. The waste test tanks can also contain other waste such as liquid taken directly from the floor drain tanks when that liquid does not require processing in the evaporator, l from the installed vendor resin skid, distillate from the boron recovery evaporator when the BRS evaporator is substituting for the waste evaporator, and distillate from the Steam Cenerator Blowdown System evaporators and flesh steam condensers when that system must discharge liquid off-site.

If testing indicates that purification of the waste test tank contents is required prior to release, the liquid can be circulated through the waste demineralizer and filter.

The contents of the waste test tank may be reused in the Nuclear System if the sample test meets the purity requirements.

Prior te discharge, each waste test tank is analyzed for principal gamma emitters in accordance with the liquid sample and analysis program outlined in l Part A to the ODCM.

X-B.2-2 ODCM Rev. 18

~

1 2.2 METHOD TO DETERMINE RADIONUCLIDE CONCENTRATION ?OR EACH LIQUID EFFLUENT SOURCE 2.2.2 Turbine Buildine Sumn l The Turbine Building sump collects leakage from the Turbine Building floor l

drains and discharges the liquid unprocessed to the circulating water system.

Sampling of this potential source is normally done once per week for l

l determining the radioactivity released to the environment (see Table A.6.1-1).

1 2.2,3 Steam Generator Blowdown Flash Tank The primary method to process radioactive secondary liquid from the steam generators is to direct steam blowdown flash tank bottoms cooler discharge to the floor drain tanks. If no secondary pressure is available, the steam blowdown and wet lay-ups pumps can be used. From the floor drain tanks, processing through the installed vendor resin skid (WL-SKD-135) to the waste test tanke is the preferred method. Other methods may be used as defined below.

The steam generator blowdown evaporators normally process the liquid from the steam generator blowdown flash tank when there is primary to secondary leakage. j Distillate from the evaporators can be sent to the waste test tanks or recycled to j the condensate system. When there is no primary to secondary leakage, flash tank liquid is processed through the steam generator blowdown demineralizers and returned to the secondary side.

Steam generator blowdown is only subject to sampling and analysis when all or I

l part of the normal blowdown recycling liquid process is being (see Table discharged A.6.1-1). to the environment instead of the 2.2.4 Primary Connonent Cooline Water (PCCW) System The PCCV System is used to cool selected primary components.

The system is normally sampled weekly to determine if there is any radwaste in-leakage. If leakage has been determined, the Service Water System is sampled to determine if any release to the environment has occurred.

X-B.2-3 ODCM Rev. 18

r ]

1 3.0 OFF-SITE DOSE CALCU1ATION METHODS Chapter 3 provides the basis for station procedures required to meet the Radiological Effluent Control Program (RECP) dosa and dose rate requirements ,

contained in ODCM Part A Controls. A simple, conservative method (called Method I) I is listed in Tal<1es B.1-2 to B.1-7 for each of the requirements of the RECP. Each I of the Method I equations is presented in Part B, Sections 3.2 through 3.9. As an alternate to Method I, the EMS computer program documented in Appendix C can be used to determine regulatory compliance for effluent dossa and dose rates. The use of the EMS software is designated as Method IA in Chapter 3. In addition, those sections include more sophisticated methods (called Method II) for use when more ;

refined results are needed. This chapter provides the methods, data, and reference material with which the operator can calculate the needed doses, dose rates and l setpoints. For the requirements to demonstrate compliance with Part A off-site dose limits, the contribution from all measured ground level releases must be added to j the calculated contribution from the vent stack to determine the Station's total '

radiological impact. The bases for the dose and dose rate equations are given in Chapter 7.0. Method IA bases and software verification documentation are contained in Appendix C.

The Annual Radioactive Effluent Release Report, to be filed after January 1 l l

l each year per Technical Specification 6.8.1.4, and Part A, Section 10.2, requires that meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, be used for determining the gaseous pathway doses. For continuous release sources (i.e., plant vent, condenser air removal exhaust, and gland steam packing exhauster), concurrent quarterly average meteorology will be used in the dose calculations along with the quarterly total radioactivity released. For batch releases or identifiable operational activities (i.e., containment purge or venting to atmosphere of the Waste Gas System), concurrent meteorology during the period of release will be used to determine dose if the total noble gas or iodine and l particulates released in the batch exceeds five percent of the total quarterly l radioactivity released from the unit; otherwise quarterly average meteorology will i be applied. Quarterly avera5e meteorology will also be applied to batch releases if the hourly met data for the period of batch release is, unavailable.

l Annual dose assessment reports prepared in accordance with the requirements of l the ODCM will include a statement indicating that the appropriate portions of Regulatory Guide 1.109 (as identified in the individual subsections of the ODCM for each class of effluent exposure) have been used to determine dose impact from station releases. Any deviation from the methodology, assumptions, or parameters l given in Regulatory Guide 1.109, and not already identified in the bases of the ODCM, will be explicitly described in the effluent report, along with the bases for the deviation.

I X B.3-1 ODCM Rev. 18

I 3.1 INTRODUCTORY CONCEPTS l In Part A Controls, the RECP limits for dose or dose rate are stated. The term " dose" for ingested or inhaled radioactivity means the dose commitment, measured in area, which results from the exposure to radioactive materials that, because of uptake and deposition in the body, will continue to expose the body to radiation for some period of time after the source of radioactivity is stopped. The time frame over which the dose commitment is evaluated is 50 years. The phrases

" annual dose" or " dose in one year" then refers to the 50-year dose commitment resulting from exposure to one year's worth of releases. " Dose in a quarter" l similarly means the 50-year dose commitment resulting from exposure to one quarter's l releases. The term " dose," with respect to external exposures, such as to noble gas clouds, refers only to the doses received during the actual time period of exposure to the radioactivity released from the plant. Once the source of the radioactivity is removed, there is no longer any additional accumulation to the dose commitment.

" Dose rate" is the total dose or dose commitment divided by exposure period.

l For example, an individual who is exposed via the ingestion of milk for one year to radioactivity from plant gaseous effluents and receives a 50-year dose commitment of 10 mrem is said to have been exposed to a dose rate of 10 mrem / year, even though the actual dose received in the year of exposure may be less than 10 mrem.

In addition to limits on dose commitment, gaseous effluents from the station are also controlled so that the maximum or peak dose rates at the site boundary at any time are limite<t to the equivalent annual dose limits of 10CFR, Part 20 to unrestricted areas (if it were assumed that the peak dose rates continued for one year). These dose rate limits provide reasonable assurance that members of the public, either inside or outside the site boundary, will not be exposed to annual averaged concentrations exceeding the limits specified in Appendix B, Table II of 10CFR, Part 20 (10CFR20.106(a)). See Appendix B for a listing of these concentration limits.

The quantities AD and 6 are introduced to provide calculable quantities, related to off-site doses or dose rates that demonstrate compliance with the RETS.

l Delta D, denoted AD, is the quantity calculated by the Part B, Chapter 3, Method I dose equations. It represents the conservative increment in dose. The AD calculated by Method I equations is not necessarily the actual dose received by a

, real individual, but usually provides an upper bound for a given release because of l the conservative margin built into the dose factors and the selection and definition of critical receptors. The radionuclide specific dose factors in each Method I dose I

equation represent the greatest dose to any organ of any age group. (Organ dose is '

a function of age because organ mass and intake are functions of age.) The critical receptor assumed by " Method I" equations is then Senerally a hypothetical individual whose behavior - in terms of location and intake - results in a dose which is higher I than any real individual is likely to receive. Method IA dose calculations using l the EMS software evaluate each age group and organ combination to determine the I maximum organ dose for each mix of radionuclides specified in a release period.

Method II also allows for a more exact dose celculation for each individual if necessary.

I l X-B.3-2 ODCM Rev. 18 l

l

3.1 INTRODUCTORY CONCEPTS (Continusd) l D dot, denoted D, is the quantity calculated in the Part B, Chapter 3 dose Igig equations. It is calculated using the station's effluent monitoring system reading and an annual or long-term average atmospheric dispersion factor. b predicts the maximum off-site annual dose if the peak observed radioactivity release rate from the plant stack continued for one entire year. Since peak release rates, or resulting dose rates, are usually of short time duration on the order of an hour or less, this approach then provides assurance that 10CFR20.106 limits will be' met.

Each of the methods to calculate dose or dose rate are presented in the following subsections. Each dose type has two levels of complexity. Method I is the simplest and contains many conservative factors. As an alternate to Method I the EMS computer program documented in Appendix C can be used to determine regulatory compliance for effluent doses and dose rates. The use of the EMS system l is designated as Method IA in Chapter 3 of Part B. ,

Method II is a more realistic analysis which makes use of the models in l Regulatory Guide 1.109 (Revision 1), as noted in each subsection of Part B, Chapter 3 for the various exposure types. A detailed description of the methodology, assumptions, and input parameters to the dose models that are applied in each Method II calculation, if not already explicitly described in the ODCM, shall be documented and provided when this option is used for NRC reporting and l ODCM, Part A RECP dose compliance.

X-B.3-3 ODCM Rev. 18

f l

3.2 METHOD TO CALCULATE THE TOTAL BODY DOSE FROM LIQUID RELEASES l

l l Part A Control C.6.2.1 limits the total body dose commitment to a member of I

the public from radioactive material in liquid effluents to 1.5 mrem per quarter and 3 mrem per year per unit. Part A Control C.6.3.1 requires liquid radwaste treatment when the total body dose estimate exceeds 0.06 area in any 31-day period. Part A Control C.8.1.1 limits the total body dose cosmitment to any real member of the public from all station sources (including liquids) to 25 mrem in a year.

l Use Method I or Method IA first to calculate the maximum total body dose from a liquid release from the station as it is simpler to execute and more conservative than Method II.

Use Method II if a more refined calculation of total body dose is needed, l 1.e. , Method I or Method IA indicates the dose might be greater than Part A Control limits.

l l

To evaluate the total body dose, use Equation 3.1 to estimate the dose from the planned release and add this to the tota?. body dese accumulated from prior l releases during the month. See Part B, Secti n 7.1.1 for basis, j 3.2.1 Method I l The total body dose from a liquid release is:

Dtb=k Qi DFLith

, , (3-1)

(mrem) = ( ) (pC1) j where DFL itb - Site-specific total body dose factor (arem/pci) for a liquid release. It is the highest of the four age groups. See Table B.1-ll. l Qt - Total activity (gCi) released for radionuclide "i". (For strontiums, use the mos,t recent measurement available.)

k- 918/F4 ; where F4 is the average (typically monthly average) dilution flow of the Circulating Water System at the point of discharge from the multiport diffuser (in f t 8/sec). For normal operations with a cooling water flow of 918 f t 8 /sec, k is equal to 1.

l Equation 3-1 can be applied under the following conditions (otherwise, justify Method I or consider Method II):

, 1. Liquid releases via the multiport diffuser to unrestricted areas (at l the edge of the initial mixing or prompt dilution zone that corresponds to a factor of 10 dilution), and l

2. Any continuous or batch release over any time period.

X-B.3-4 ODCM Rev. 18

3.2 METHOD TO CALCULATE THE TOTAL BODY DOSE FROM LIQUID RELEASES 3.2.1 Method I (Continued)

Method IA is implemented by the EMS software as described in Appendix C.

Liquid release models are detailed in sections 2.1 - 2.6 of the EMS Technical Reference Manual (Attachment 4 of Appendix C).

3.2.2 Method II Method II consists of the models, input data and assumptions (bioaccumulation l factors, shore-width factor, dose conversion factors, and transport and buildup times) in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific data or assumptions have been identified in the ODCM. The general quations (A-3 and A-7) taken from Regulatory Guide 1.109, and used in the derivation of the simplified Method I approach as described in the Bases section, are also applied to Method II assessments, except that doses calculated to the whole body from l radioactive effluents are evaluated for each of the four age groups to determine the maximum whole body dose of an age-dependant individual via all existing exposure pathways. Table B.7-1 lists the us, age factors of Method II calculationa. As noted in Section B.7.1, the mixing ratio associated with the edge of the l'F surface isotherm above the multiport diffuser may be used in Method II calculations for the shoreline exposure pathway. Aquatic food ingestion pathways shall limit credit taken for mixing zone dilution to the same value assumed in Method I (M, - 0.10) .

l l

1 X-B.3-5 ODCM Rev. 18 l

3.3 METHOD TO CALCULATE MAXIMUM ORGAN DOSE FROM LIQUID RELEASES l Part A Control C.6.2.1 limits the maximum organ dose commitment to a Member of the Public from radioactive material in liquid effluents to 5 mrem per quarter and l 10 mrem per year per unit. Part A Control C.6.3.1 requires liquid radwaste treatment whet the maximum organ dose projected exceeds 0.2 area in any 31 days (see l Part B, Subsection 3.11 for dose projections). Part A Control C.8.1.1 limits the maximum organ dose commitment to any real member of the public from all station sources (including liquids) to 25 mrea in a year except for the thyroid, which is limited to 75 mrem in a year.

Use Method I or Method IA first to calculate the maximum organ dose from a liquid release to unrestricted areas (see Figure B.6-1) as it is simpler to execute and more conservative than Method II.

Use Method II if a more refined calculation of organ dose is needed, i.e., Method I or Method IA indicates the dose may be greater than ti.e limit.

Use Equation 3-2 to estimate the maximum organ dose from individual or l combined liquid relmasos. See Part 3, Section 7.1.2 for basis.

3.3.1 Method I l The maximum organ dose from a liquid release is:

D, = k Qi DFL 1 ,

(3-2)

(arem) = ( ) (pCi) where DFLi , -

Site-specific maximum organ dose factor (area /pci) for a liquid release. It is the highest of the four age groups. See

'"able B.1-11.

Qi - Total activity (pci) released for radionuclide "i". (For composited analyses of strontiums, use the most recent measurement available.)

k- 918/Fo; where F4 is the average (typically monthly average) dilution flow of the Circulating Water System at the point of discharge from the multiport diffuser (in ft 8/sec). For normal operations with a cooling water flow of 918 ft 8/see, k is equal to 1.

Equation 3-2 can be applied under the following conditions (otherwise, justify Method I or consider Method II):

1. Liquid releases via the multiport diffuser to unrestricted areas (at the edge of the initial mixing or prompt dilution zone that corresponds to a factor of 10 dilution), and

2. Any continuous or batch release over any time period.

X-B.3-6 ODCM Rev. 18

3.3 METHOD TO CALCULATE MAXIMUM ORGAN DOSE FROM LIQUID RELEASES 3.3.1 Method I (Continued)

Method IA is implemented by the EMS sof tware as described in Appendix C.

Liquid release models are detailed in sections 2.1 - 2.6 of the EMS Technical Reference Manual (Attachment 4 of Appendix C).

3.3.2 Method II Method II consists of the models, input data and assumptions (bioaccumulation factors, shore-width factor, dose conversion factors, and transport and buildup times) in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific data or assumptions have been identified in the ODCM. The general equations (A-3 and A-7) taken from Regulatory Guide 1.109, and used in the derivation of the simplified Method I approach as described in the Bases section, are also applied to Method II assessments, except that doses calculated to critical organs from radioactive effluents are evaluated for each of the four age groups to determine the maximum critical organ of an age-dependent individual via all existing exposure pathways. Table B 7-1 lists the usage factors for Method II calculations. As noted in Section B.7.1, the mixing ratio associated with the edge of the l'F surface isotherm above the multiport diffuser may be used in Method II calculations for ti+

shoreline exposure pathway. Aquatic food ingestion pathways shall limit credit taken for mixing zone dilution to the same value assumed in Method I (M, - 0.10) .

l 1

l I

l 1

1 l

l l

X-B.3-7 ODCM Rev. 18 f

I

i 3.4 METHOD TO CALCULATE THE TOTAL BODY DOSE RATE FROM NOBLE GASES i

l Part A Control C.7.1.1 limits the dose rate at any time to the total body from

, noble gases at any location at or beyond the site boundary to 500 mrem / year. The l

l Part A Control indirectly limits peak release rates by limiting the dose rate that is predicted from continued release at the peak rate. By limiting B tb to a rate equivalent to no more than 500 arem/ year, we assure that the total body dose accrued in any one year by any member of the general public is less than 500 mrem.

l Use Method I or Method IA first to calculate the Total Body Dose Rate from the peak release rate via the station vents or ground level effluent release points.

Method I applies at all release rates.

l Use Method II if a more refined calculation of Do, is desired by the station (i.e., use of actual release point parameters with annual or actual meteorology to obtain release-specific X/Qs) or if Method I or Method IA predicts a dose rate greater than the Part A Control limit to determine if it had actually been exceeded .

during a short time interval. See Part B, Section 7.2.1 for basis. !

Compliance with the dose rate limits for noble gases are continuously I demonstrated when effluent release rates are below the plant vent noble gas activity monitor alarm setpoint by virtue of the fact that the alarm setpoint is based on a value which corresponds to the off site dose rate limit, or a value below it.

l Determinations of dose rate for compliance with Part A Controls are performed when the effluent monitor alarm setpoint is exceeded, or as required by the Action l Statement (Part A Control C.S.2, Table A.S.2-1) when the monitor is inoperable.

3.4.1 Method I The Total Body Dose Rate to an off-site receptor due to noble gases in effluents released via the plant vent can be determined as follows:

Debc.3 = 0.85 * { ('Qi + DFB ) i i

' (3-3a) arem ,'pci-sec' 'gCi 'arem-m 3 yr pCi-m3 ,

{ ,s e c, ,'FLT-K where D tb -

The off-site total body dose rate (arem/yr) due to noble gases in elevated effluent releases,

- the release rate at the station vents (pci/sec), for each noble Qi gas radionuclide, "i", shown in Table B.1-10, and DFB t - total body gamma dose factor (see Table B.1-10).

The Total Body Dose Rate (to an off site receptor) due to noble gas in ground level effluent releases can be determined as follows:

l X-B.3 8 ODCM Rev. 18 I

I l

3.4 METHOD TO CALCULATE THE TOTAL BODY DOSE RATE FROM NOBLE CASES 3.4.1 Method I (Continued) ,

I l .

8 1 l Debts)

= 3.4 g

{ (Qi

DFB )

l (3-3b) mrem , pCi-sec' 'pci 'aree-m'3 l

,g ,

yr , Ci-m 3 ,

,s e c, ,pC L-yr ,

I l

where O t3 - The total off-site body dose rate (ares /yr) due to noble gases in elevated effluent releases, and Qi and DFBi are as defined for Equation 3-3a. j For the spec 5si on-site receptor locations, the Science & Nature Center and j the " Rocks," the te il body dose rates due to noble gases in effluent discharges can be determined as follows:

For the Science & Nature Center, elevated effluent release:

O th (3-3c) e<.) - 0.0015 * { (Qi

DFB ) i For the Science & Nature Center, grounc level effluent release:

DtbE(s) - 0.0074 * {t (Qi

DFBi) (3-3d)

For the " Rocks,a elevated effluent release:

Otbat.) - 0.038 * {$ (Q i t) (3-3e)

For the " Rocks," ground level effluent release:

Otba(s) - 0.2 * { (Qi a DFBg) (3-3f) where D ebt(*). D*E(s)*

t Otbat.), andOtba - The total body dose rate (arem/yr) at the Science & Nature Center and the " Rocks," respectively, due to noble gases in gaseous discharges from elevated (e) and ground level (g) release points, and l

! Q and DFBi are as defined previously.

i

Equations 3-3a through 3-3f can be applied under the following conditic.

(otherwise, justify Method I or consider Method II):

X B.3-9 ODCM Rev. 18

3.4 METHOD TO CALCULA1E THE TOTAL BODY DOSE RATE FROM NOBLE GASES 3.4.1 Method I (Continued)

1. Normal operations (nonemergency event), and

2. Noble gas releases via any station vent to the atmosphere.

Method IA is implemented by the EMS software as described in Appendix C.

Gaseous release models are detailed in Section 6.7.3 of the EMS Software ,

Requirements Specification (Attachment 3 of Appendix C).

l 3.4.2 Method II

! Method II consists of the model and input data (whole body. dose factors) in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific data or l assumptions have been identified in the ODCM. The general equation (B-8) taken from l Regulatory Guide 1.109, and used in the derivation of the simplified Method I approach as described in the Bases section, is also applied to a Method II l assessment. No credit for a shielding factor (S F) associated with residential I structures is assumed. Concurrent meteorology with the release period may be utilized for the gamma atmospheric dispersion factor identified in ODCM Equation 7-3 l (Part B, Section 7.2.1), and determined as indicated in Part B, Section 7.3.2 for

, the release point (either ground level or vent stack) from knich recorded effluents l have been discharged.

I t

l X-B.3-10 ODCM Rev. 18 l

l t

3.5 METHOD TO CALCUIATE THE SKIN DOSE RATE FROM NOBLE CASES Part A Control C.7.1.1 limits the dose rate at any time to the skin from noble I gases at any locat. ion at or beyond the site boundary to 3,000 mres/ year. The Part A Control indirectly limits peak release rates by limiting the dose rate that is I predicted from continued release at the peak rate. By limiting 6,gt, to a rate equivalent to no more than 3,000 mrem / year, we assure that the skin dose accrued in any one year by any member of the general public is less than 3,000 mrem. Since it can be expected that the peak release rate on which 6,mi, is derived would not be exceeded without corrective action being taken to lova it, the resultant average release rate over the year is expected to be consideraoly le.ss than the peak release rate.

Use Method I or Method IA first to calculate the Skin Dose Rate from peak release rate via station vents. Method I applies at all release rates.

Use Method II if a more refined calculation of D,gi, is desired by the station ;

(i.e. , use of actual release point parameters with annual or actual meteorology to '

obtain release-specific X/Qs) or if Method I or Me Sod IA predicts a dose rate greater than the Part A Control limit to determine if it had actually been exceeded during a short time interval. See Part B, Section 7.2.2 for basis.

Compliance with the dose rate limits for noble gases are continuously demonstrated when effluent release rates are below the plant vent noble gas activity monitor alarm setpoint by virtue of the fact that the alarm setpoint is based on a value which corresponds to the off-site dose rate limit, or a value below it.

l Determinations of dose rate for compliance with Part A Controls are performed winen the effluent monitor alarm setpoint is exceeded.

3.5.1 Method I For an off-site receptor and elevated effluent release, the Skin Dose Rate due to noble gases is:

D akint.) "

[Qt

DF'ic 3)

(3-4a) mrem ,{ pCi arem-sec yr ,s e c, ,

pCL-yr ,

where 6,gioc,3 - the off-site skin dose rate (ares /yr) due to noble gases in an effluent discharge from an elevated release point, Qi - as defined previously, and l D6g,3 - the combined skin dose factor for elevated discharges (see l

Table B.1-10).

X B.3-11 ODCM Rev. 18

3.5 METHOD TO CALCULATE THE SKIN DOSE RATE FROM NOBLE GASES j l

3.5.1 Method 2 (Continued) l i

For an off-site receptor and ground level release, the skin dose rate due to noble gases is:

l D,minc,) - (Qi

DF[g,3) (3-4b)

I where l b, ming,3 - The off-site skin dose rate (area /yr) due to noble gases in an effluent discharge from a ground level release point, Qi

- as defined previously, and DF[g,3 - The combined skin dose factor for ground level discharges (see Table B.1-10).

For an on-site receptor at the Science & Nature Center and elevated release conditions, the skin dose rate due to noble gases is:

0.htnzc.) - 0.0014 * (Qi

DFisc.3) (3-4c) where 6,gingg,3 - The skin dose rat.e (arem/yr) at the Science & Nature Center due to noble gases in an elevated release, Qi - as defined previously, and DF[g33 - the combined skin dose factor for elevated discharges (see Table B.1-13).

For an on-site receptor at the Science 6 Nature Center and ground level release conditions, the skin dose rate due to noble gases is:

D.ktngc,3 - 0.0014 * (Qi

DFisc,3) (3-4d) where 6 ,gingg,3 - the skin dose rate (ares /yr) at the Science & Nature Center due to noble gases in a ground level release, Qi - as defined previously, and j DF$r(s) - The combined skin dose factor for ground level discharges (see l Table B.1-13).

t I

! X-B.3-12 ODCM Rev. 18 l

l l I

f' 3.5 METHOD TO CALCUIATE THE SKfN DOSE RATE FROM NOBLE CASES 3.5.1 Method I (Continued)

For an on-site receptor at the " Rocks" and elevated release conditions, the skin dose rate due to noble gases is:

D.kisc.) - 0.0076 . { (Qi

Diac.3) (3-4e) l where l

D skinat.) - the skin dose rate at the " Rocks" due to noble gases in an elevated release.

Qi - as defined previously, and l

l DIr<.)

t

- The combined skin dose f ctor for elevated discharges (see Table B.1-13).

l For an on-site receptor at the " Rocks" and ground level release conditions, j

~

the skin dose rate due to noble gases is:

b kina<s) - 0.0076

i{ (Qi

DIac 3) (3-4f) where D ktna(s) the skin dose rate (ares /yr) at the " Rocks" due to noble gases in a ground level release, l Qi - as defined previously, and Ding,3 i

the combined skin dose factor for ground level discharges (see Table B.1-13).

Equations 3-4a through 3-4f can be applied under the following conditions (otherwise, justify Method I or consider Method II).

1. Normal operations (nonemergency event), and f

2. Noble gas releases via any station vent to the atmosphere.

I Method IA is implemented by the EMS software as described in Appendix C.

Gaseous release models are detailed in Section 6.7.3 of the EMS Software l Requirements Specification (Attachment 3 of Appendix C).

X-B.3-13 ODCM Rev. 18

f-3.5 METHOD TO C.ALCUIATE THE SKIN DOSE RATE FROM NOBLE GASES 3.5.2 Method II Method II consists of the model and input data (skin dose factors) in Regulatory Guide 1.109, Rev. 1 (Reference ~A), except where site-specific data or assumptions have been identified in the ODCM. The general equation (B-9) taken from i Regulatory Guide 1.109, and used in the derivation of the simplified Method I l approach as described in the Bases section, is also applied to a Method II

! assessment, no credit for a shielding factor (Sy) associated with residential i

structures is assumed. Concurrent meteorology with the release period may be l utilized for the gamma atmospheric dispersion factor and undepleted atmospheric j dispersion factor identified in ODCM Equation-7-8 (Part 5,' Section 7.2.2), and determined as indicted in Part B, Sections 7.3.2 and 7.3.3 for the release point l

(either ground level or vent stack) from which recorded effluents have been discharged.

I i

i i

l l

l X-B.3-14 ODCM Rev. 18

3.6 METHOD TO CALCUIATE THE CRITICAL ORGAN DOSE RATE FROM 20 DINES, TRITIUM AND PARTICUIATES WITH Tif: CREATER THAN 8 DAYS l Part A Control C.7.1.1 limits the dose rate at any time to any organ from 1311, issy , 3H and radionuclides in particulate form with half lives greater than 8 l days to 1500 arec/ year to any organ. The Part A Control indirectly limits peak release rates by limiting the dose rate that is predicted from continued release at the peak rate. By limiting b,, to a rate equivalent to no more than 1500 mrem / year, we at,*ure that the critical organ dose accrued in any one year by any member of the general public is less than 1500 mrea.

Use Method I or Method IA first to calculate the Critical Organ Dose Rate from the peak release rate via the station vents. Method I applies at all release rates.

Use Method II if a more refined calculation of 6,, is desired by the station

(i.e., use of actual release point parameters with annual or actual meteorology to l obtain release-specific X/Qs) or if Method I or Method IA predicts a dose rate l greater than the Part A Control limit to determine if it had actually been exceeded during a short time interval. See Part 8, Section 7.2.3 for basis.

3.6.1 Method I The Critical Organ Dose Rate to an off-site receptor and elevated release conditions can be determined as follows:

6eoc.3 -

(4

DFCge,(,)) Wa)

( mrem) , g ( pCi) sec

, ( mrem-sec) yr pci-yr where 6,,c,3 - The off-site critical organ dose rate (arem/yr) due to iodine, tritium, and particulates in an elevated release, Qi

- the activity release rate at the station vents of radionuclide "1" in pCi/sec (i.e., total activity measured of radionuclide "i" averaged over the time period for which the filter / charcoal sample collector was in the effluent stream. For i - Sr89 or Sr90, use the best estimates, such as most recent measurements),

and DFC' 3,,g,3 - the site-specific critical organ dose rate factor

(* **) for an elevated gaseous release (see Table B.1-12) .

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X-B.3-15 ODCM Rev. 18

3.6 METHOD TO CALCULATE THE CRITICAL ORGAN DOSE RATE FROM IODINES, TRITIUM AND PARTICUIATES WITH T1/2 GREATER THAN 8 DAYS 3.6.1 Method I (Continued)

For an off-site receptor and ground level release, the critical organ dose rate can be determined as follows:

6,,c,3- (Qi DFGle,g,3) (3-5b) where b,,c,3 - the off-site critical organ dose rate (aren/yr) due to iodine, tritium, and particulates in a ground level release, Qi - as defined previously, and DFCl,,g,3 - the site specific critical organ dose rate factor for a ground level gaseous discharge (see Table B.1-12).

For an on-site receptor at the Science & Nature Center and elevated release conditions, the critical organ dose rate can be determined as follows:

Deost.) - 0.0014 * (Qt DFCl,.s c.3) (3-Sc) where 6,,gg,3 - The critical organ dose rate (ares /yr)_to a receptor at the Science & Nature Center due to iodine, tritium, and particulates in an elevated release, l Qi - as defined previously, and DFGleer<.)- the Science & Nature Center-specific critical organ dose rate factor for an elevated discharge (see Table B.1-14).

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X-B.3-16 ODCM Rev. 18

3. 6 METHOD TO CALCULATE THE CRITICAL ORGAN DOSE RATE FROM IODINES, TRITIUM AND PARTICUIATES WITH Tu2 GREATER THAN 8 DAYS 3.6.1 Method I (Continued)

For an on-site receptor at the Science & Nature Center and ground level release conditions, the critical organ dose rate fs:

Dcorca) - 0.0014 * (Qi eDFGl.gc,3) (3-5d) where D,,gt,3 - the critical organ dose rate (arem/yr) to a receptor at the Science & Nature Center due to iodine, tritium, and particulates in a ground level release, Qi - as defined previously, and DFGl,,gt,3 - the Science & Nature Center-specific critical organ dose rate factor for a ground level discharge (see Table B.1-14).

For an on-site receptor at the " Rocks" and elevated rolsase conditions, the critical organ dose rate is:

Deont.) - 0.0076 * (Qi

DFC$ coat.)) (3-Se) where b eca<.) - The critical organ dose rate (ares /yr) to a receptor at the

" Rocks" due to iodine, tritium, and particulates in an elevated release, Qi

- as defined previously, and DFGle.ac , - the " Rocks"-specific critical organ dose rate factor for an elevated discharge (see Table B.1-15).

For an on-site receptor at the " Rocks" and ground level release conditions, the critical organ dose rate is:

Decac,3 - 0.0076 * (Qi e DFGieom(s)) (3-5f) where bn eo and Q - i are as defined previously, and DFGlcon(s)

- the " Rocks"-specific critical organ dose rate factor for a ground level discharge (see Table B.1-15),

r X-B.3-17 ODCM Rev. 18 i

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, 3.6 METHOD TO CALCULATE THE CRITICAL ORGAN DOSE RATE FROM IODINES TRITIUM AND PARTICULATES WITH Tgf GREATER THAN 8 DAYS 3.6.1 Method I (Continued) l l Equations 3-Sa through 3-5f can be applied under the following conditions

! (otherwise, justify Method I or consider Method II):

1. Normal operations (not emergency event), and 1

2. Tritium, I-131 and particulate releases via monitored station vents to the atmosphere.

l Method IA is implemented by the EMS software as described in Appendix C. J Gaseous release models are detailed in Section 6.7.3 of the EMS Software l

Requirements Specification (Attachment 3 of Appendix C). ,

3.6.2 Method II Method II consists of the models, input data and assumptions in Appendix C of Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific data e assumptions have been identified in the ODCM (see Tables B.7-2 and B.7-3). The critical organ dose rate will be determined based on the location (site boundary, i nearest resident, or farm) of receptor pathways as identified in the most recent l annual land use census, or by conservatively assuming the existence of all pathways

(ground plane, inhalation, ingestion of stored and leafy vegetables, milk, and meat) at an off-site location of maximum potential dose. Concurrent meteorology with the release period may be utilized for determination of atmospheric dispersion factors ,

l in accordance with Part B, Sections 7.3.2 and 7.3.3 for the release point (either l ground level or vent stack) from which recorded effluents have been discharged. The maximum critical organ dose rates will consider the four age groups independently, and take no credit for a shielding factor (S F ) associated with residential structures.

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X-B.3-18 ODCM Rev. 18

3.7 METHOD TO CALCULATE THE CAMMA AIR DOSE FROM NOBLE GASES l Part A Control C.7.2.1 limits the gamma dose to air from noble gases at any location at or beyond the site boundary to 5 mrad in any quarter and 10 mrad in any year per unit. Dose evaluation is required at least once per 31 days.

Use Method I or Method IA first to calculate the gamma air dose from the station gaseous effluent releases during the period.

Use Method II if a more refined calculation is needed (i.e.,.use of actual release point parameter with annual or actual meteorology to obtain release-specific X/Qs), or if Method I or Method IA predicts a dose greater than the Part A Control limit to determine if it had actually been exceeded. See Part B Section 7.2.4 for basis.

3.7.1 Method I The general form of the gamma air dose equation is:

D't, - 3.17E-02 * [X/QlIhr

t * (Qt

DFZ)

(3-6)

(mrad) .

PCi-yr' * *( ) * { (pci) ' mrad-m3'

,Ci-sec, p m ,

,Pci-yr, where D',1, is the gamma air dose.

3.17E-02 is the number of pCi per pCi divided by the number of second per year,

[X/Q]In, is the 1-hour gamma atmospheric dispersion factor, t** is a unitiess factor which adjusts the 1-hour [X/Q]' value for a release with a total duration of t hours, Qi is the total activity in pCi of each radionuclide "i" released to the atmosphere from the station gaseous effluent release point during the period of interest, and DF7t is the gamma dose factor to air for radionuclide "i" (see Table B.1 10).

Inco 7 orating receptor location-specific atmospheric dispersion factors

([X/Q]'), adjustment factors (t**) for elevated and ground-level effluent release conditions, and occupancy factors when applicable (see Section 7.2.7), yields a series of equations by which the gamma air dose can be determined.

a. Maximum off-site receptor location, elevated release conditions:

D'i,g,3 = 3. 2E-07

t-o.zn , (q, ,

9 77)

(arad) . PCi-yr' *( ) * { (pC1) "Y

d -*8 '

p 3

, C 1 -m , ,PC1 yr, X-B.3-19 ODCM Rev. 18 1

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3.7 METHOD TO CALCUIATE THE CAMMA ATR DOSE FROM NOBLE CASES 3.7.1 Method I (Continued)

b. Maximum off-site receptor location, ground-level release conditions:

Dji,c,3 = 1. 6E-06

t-o.2e3 { (q, , pp7)

(mrad) . P Ci-yr' *( )* (pCi) mrad-af

,pCi-m3 , ,PCL-yr, 1

c. Science & Nature Center receptor; elevated release conditions: J I

Dl tert.) " 4. 9E-10

t-o.252 * (Qi

DFI) (3-6c) 3 Ci-yr (mrad) = ( PpCi-m3 ) * ( ){ (pCie arad-mPGi-yr

)

d. Science & Nature Center receptor; ground level release conditions:

Dji,gg,3 = 4.4E-09 + t-o.321 * { (Qi

DFI) (3-6d) 8 (mrad) = ( CP i-yr) * ( ){ (pCie arad-m )

pCi-m' PGi-yr.

e. Receptor at the " Rocks"; elevated release conditions:

Dji,me,3 = 5. lE-09

t-0 155 * (Qi

DFI) (3-6e) 8 Ci-yr (mrad) = ( PpCi-m3 ) * ( ) { (pci

mrad-m pGi-yr

)

l

f. Receptor at the " Rocks"; ground-level release conditions:

Dji,gg,3 = 4. lE-08

t-o.204 * [(Qi

DFl) (3-6f) 1 3

(mrad) = ( CP i-yr) * (

3

){ (pCie arad-m )

pCL-m pCi-yr Equations 3-6a through 3-6f can be applied under the following conditions (otherwise justify Method I or consider Method II):

1. Normal operations (nonemergency event), and

2. Noble gas releases via station vents to the atmosphere.

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X-B.3-20 ODCM Rev. 18

r-I 3.7 METHOD TO CALCUIATE THE GAMMA AIR DOSE FROM NOBLE CASES 3.7.1 Method I (Continued)

Method IA is implemented by the EMS software as described in Appendix C.

l Caseous release models are detailed in Section 6.7.3 of the EMS Software Requirements Specification (Attachment 3 of Appendix C).

3.7.2 Method II Method II consists of the models, input data (dose factors) and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific data or assumptions have been identified in the ODCM. The general equations (B-4 and B-5) J taken from Regulatory Guide 1.109, and used in the derivation of the simplified 1 l Method I approach as described in the Part B Bases Section 7.2.4 are also applied to j Method II assessments. Concurrent meteorology with the release period may be utilized for the gamma atmoaphoric dispersion factor identified in ODCM Equation l 7-14, and determined as indicated in Part B, Section 7.3.2 for the release point 3 (either ground level or vent stack) from which recorded effluents have been l discharged.

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I X B.3-21 ODCM Rev. 18 n

i 3.8 METHOD TO CALCULATE THE BETA AIR DOSE FROM NOBLE CASES l Part A Control C.7.2.1 limits the beta dose to air from noble gases at any location at or beyond the site boundary to 10 mrad in any quarter and 20 mrad in any year per unit. Dose evaluation is required at least once per 31 days.

Use Method I or Method IA first to calculate the beta air dose from gaseous effluent releases during the period. Method I applies at all dose levels.

Use Method II if a more refined calculation is needed (i.e., use of actual release point parameters with annual or actual meteorology to obtain release-specific X/Qs) or if Method I or Method IA predicts a dose greater than the l Part A Control limit to determine if it had actually been exceeded. See Part B, Section 7.2.5 for basis.

i 3.8.1 Method I

'Ihe general form of the beta air dose equation is:

D{i,= 3.17E-02 *

(X/Q)1hr

t' * { (Q

Dd) (3-7) i

)

PC i-yr '

(mrad) . , 's e c'

( )* pCi

mrad-m3 '

,Gi-sec, p m3 , ,

PCi yr, where D{3, is the beta air dose, 3.17E-02 is the number of pCi per pCi divided by the number of seconds per year, (X/Q) the is the 1-hour undepleted atmospheric dispersion factor, t** is a unitiess factor which adjusts the 1-hour X/Q value for a release with a total duration of t hours, Qi is the total activity (pci) of each radionuclide "i" released to the atmosphere during the period of interest, and DF{ is the beta dose factor to air for radionuclide "i" (see Table B.1-10).

Incorporating receptor location-specific atmospheric dispersion factor (X/Q), ,

adjustment factors (t**) for elevated and ground-level effluent release conditions, and occupancy factors when applicable (see Section 7.2.7) yields a series of equations by which the Beta Air Dose can be determined.

a. Maximum off-site receptor location, elevated release conditions:

D{3,c.3 = 4.1E-7

t-0 3 * (Qi

Dd) (3-7a) a (mrad) = ( pCi-yr) *( ) { (pCi

mrad-m )

pCi-m 3 PGi-yr X B.3 22 ODCM Rev. 18

3.8 METHOD TO CALCULATE THE BETA AIR DOSE FROM NOBLE CASES 3.8.1 Method I (Continued)

b. Maximum off-site receptor location, ground-level release conditions:

D{t,c,3 = 6.0E

t-o.ste . (Qi e DFf) (3-7b)

I t

3 (mrad) = ( CP i-yr)3

( ) { (pCi e arad-m )

pCi-m PGi yr

c. Science & Nature Center receptor; elevated release conditions: 4 D{i,gg,3 = 1. 8E-09

t-o.ss * (Qi

DFI) (3-7c) 8 (arad) = (PCi-yr) 3

( ){ (pcie arad-m )

pCi-m PGi-yr

d. Science & Nature Center receptor; ground-level release conditions:

I D{ter(s) = 2.4E-08 + t-8 347 * (Qi

DF() ( -7d) 3 Ci-yr (arad) = ( PpCi-m3 ) * ( ) { (pCi e arad-m PCi-yr

)

e. Receptor at the " Rocks"; elevated release conditions:

D{i,ng,3 = 3. 9F.-08

t-o.24e * (Qi

DFf) (3-7e) 8 (arad) = (d.i~Y#) 3

( ) { (pCi e arad-m )

pCi-m PCi-yr

f. Receptor at the " Rocks"; ground-level release conditions:

D$trats = 4. 6E-07

t-o.as7 * (Qi*DFT) (3-7f) 3 Ci-yr (arad) = ( PpCi-m 3 ) * (

) { (pCi

mrad-m PGi-yr

)

I X-B.3-23 ODCM Rev. 18

3.8 METHOD TO CALCULATE THE BETA AIR DOSE FROM NOBLE GASES 3.8.1 Method I (Continued)

Equations 3-7a through 3-7f can be applied under the following 'itions (otherwise justify Method I or consider Method II):

1. Normal operations (nonemergency event), and

2. Noble gas releases via station vents to the atmosphere.

Method IA is implemented by the EMS sof tware as described in Appendix C.

Caseous release models are detailed in Section 6.7.3 of the EMS Software Requirements Specification (Attachment 3 of Appendix C).

3.8.2 Method II Method II consists of the models, input data (dose factors) and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific data or assumptions have been identified in the ODCM. The general equations (B-4 and B-5) taken from Regulatory Guide 1.109, and used in the derivation of the simplified l Method I approach as described in the Part B Bases Section 7.2.5, are also applied to Method II assessments. Concurrent meteorology with the release period may be utilized for the atmospheric dispersion factor identified in ODCM Equation 7-15, and l determined, as indicated in Part B, Sections 7.3.2 and 7.3.3 for the release point (either ground level or vent stack) from which recorded effluents have been discharged.

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X-B.3-24 ODCM Rev. 18

3.9- METHOD TO CALCULATE THE CRIT 8 CAL ORGAN DOSE FROM 10D1NES, TRITIUM AND {

PARTICUIATES l Part A Control C.7.3.1 limits the critical organ dose to a member of the J public from radioactive iodines, tritium, and particulates with half-lives greater than 8 days in gaseous effluents to 7.5 ares per quarter and 15 mres per year per !

l ur.it. Part A Control C.7.3.1 limits the total body and organ dose to any real member of the public from all station sources (including gaseous effluents) to 25 mrea in a year except for the thyroid, which is limited to 75 mrem in a year.

Use Method I or Method IA first to calculate the critical organ dose from gaseous effluent releases as it is simpler to execute and more conservative than j Method II. '

Use Method II if a more refined calculation of critical organ dose is needed (i.e., Method I or Method IA indicates the dose is greater than the limit). See l Part B, Section 7.2.6 for basis.

3.9.1 Method I D , = (X/Q)f{f/(X/Q)$1

t-* * (Qi

DFG3 ,,) (3-8)

(mram) = ( )/( )*( ) * [ (pC1) * ( )

where D., is the critical organ dose from iodines, tritium, and particulates, (X/Q)f{f is the 1-hour depleted atmospheric dispersion factor.

(X/Q)$1 is the annual average depleted atmospheric dispersion.

t

is a unitiess adjustment factor to account for a release with a total '

i duration of t hours, Qi is the total activity in pCi of radionuclide "i" released to the atmosphere during the period of interest (for strontiums, use the most recent measurement), and DFGi ,is the site-specific critical organ dose factor for radionuclide "i",

see Tables B.1-12, B.1-14, and B.1-15. (For each radionuclide, it is the age group and organ with the largest dose factor.)

l' l Incorporating receptor location-specific atmospheric dispersion factors j ((X/Q)fj'l and (X/Q)$1) and adjustment factors (t-*) for elevated and ground-level release conditions, and incorporating occupancy factors when. applicable (see Section 7.2.7), yields a series of equations by which the critical organ dose can be determined.

X-B.3 25 ODCM Rev. 18 l-

3.9 METHOD TO CALCULATE 11'I ';RITICAL ORGAN DOSE FRE 20 DINES, TRITIUM AND PARTICUIATES 1

3.9.1 Method I (Continued) 1 a. Maximum off-site receptor location, elevated release conditions:

l D .g 3 = 14. 8

t-o.2e7 * (Q3

DFGg ,,g,3) (3-8a)

(mrem) = ( )*( ) { ( Ci * )

b. Maximum off-site receptor location, ground-level release conditions:

D,.c,3 = 17. 7

t-o. ate * (Q

DFC ) (3-8b)

(area) = ( )*( ) { (pCi * )

c. Science & Nature Center receptor; elevated release conditions:

(3 8c)

Deo r(*) = 3. 3E-02

t-0 8'8 * { (Qi

DFCscos(*>)

(arem) = ( )*( ) { (pci * )

d. Scic.nce & Nature Center receptor; ground-level release conditions:

Deor(s) = 3. 3 E-02

t-0 8'7 * (Qi< DFGteer(s)) (3-8d)

(area) = ( )*( ) { (pci * )

e. Receptor at the " Rocks"; elevated release conditions:

Dcom(*) = 7. 3E-02

t-o.26e * (Qi

DFG3 ,,ag, ) (3-8e)

(mrem) = ( )*( ){ (pci* )

f. Receptor at the " Rocks"; ground-level telease conditions:

Desats) = 8.6E-02

t 887 * { (Qi

DFCscoats)) (3-8f)

(aren) = ( )*( ) { ( Ci * )

Equations 3-8a through 3-8f can be applied under the following conditions (otherwise, justify Method I or consider Method II):

1. Normal operations (nonemergency event),

2. Iodine, tritium, and particulate releases via station vents to the atmosphere, and

3. Any continuous or batch release over atv time period.

X B.3-26 ODCM Rev. 18

3.9 METHOD TO CALCULATE THE CRITICAL ORGAN DOSE FROM 10 DINES, TRITIUM AND PARTICULATES 3.9.1 Method I (Continued)

Method IA is implemented by the EMS software as described in Appendix C.

Gaseous release models are detailed in Section 6.7.3 of the EMS Software Requiremerts Specification (Attachment 3 of Appandix C).

3.9.2 Method II Method II consists of the models, input data and assumptions in Appendix C of Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific data or assumptions have been identified in the ODCM (see Tables B.7-2 and B 7-3). The critical organ dose will be determined based on the location (site boundary, nearest resident, or farm) of receptor pathways, as identified in the most recent annual land use census, or by conservatively assuming the existence of all pathways (ground plane, inhalation, ingestion of stored and leafy vegetables, milk and meat) at an off-site location of maximum potential dose. Concurrent meteorology with the release period may be utilized for determination of atmospheric dispersion factors l in accordance with Part B, Sections 7.3.2 and 7.3.3 for the release point (either ground level or vent stack) from which recorded effluents have been discharged. The maximum critical organ dose will consider the four age groups independently, and use a shielding factor (Sr) of 0.7 associated with residential structures.

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i X-B.3-27 ODCM Rev. 18 e

3.10 METHOD TO CALCULATE DIRECT DOSE FROM PLANT OPERATION l l Part A Control C 8.1.1 restricts the dose to the whole body or any organ to

any member of the public from all uranium fuel cycle sources (including direct radiation from station facilities) to 25 area in a calendar year (except the ;

thyroid, which is limited to 75 area). It should be noted that since there are no l uranium fuel cycle facilities within 5 miles of the station, only station sources l need be considered for determining compliance with Part A Control C.8.1.1.

3.10.1 Method i i

The direct dose from the station will be determined by obtaining the dose from l i TLD locations situated on-site near potential soarcos of direct radiction, as well i as those TLDs near the site boundary which are part of the environmental monitoring program, and subtracting out the dose contributior from background. Additional methods to calculate the direct dose may also be used to supplement the TLD information, such as high pressure ion chamber measurements, or analytical design calculations of direct dose from identified sources (such as solid waste storage facilities).

l The dose determined from direct measurements or calculations will be related !

to the nearest real person off-site, as well as those individuals on-site involved in activities at either the Education Center or the Rocks boat landing, to assess the contribution of direct radiation to the total dose limits of Part A Control C.8.1.1 in conjunction with liquid and gaseous effluents. l I

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i l X-B.3-28 ODCM Rev., 18 l

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3.11 DOSE PROJECTIONS' I Part A Controls C.6.3.1 and C.7.4.1 require that appropriate portions of tiquid and gaseous radwaste treatment systems, respectively, be used to reduce radioactive effluents when it is projected that the resulting dose (s) would exceed limits which represent small fractions of the "as low as reasonably achievable" criteria of Appendix I to 10CFR Part 50. The surveillance requirements of these l Part A Controls state that dose projections be performed at least once per 31 days when the liquid radwaste treatment systems or gaseous radwaste treatment systems are not being fully utilized.

Since dose assessments are routinely performed at least once per 31 days to account for actual releases, the projected doses shall be determined by comparing the calculated dose from the last (typical of expected operations) completed 31-day period to the appropriate dose limit for use of radwaste equipment, adjusted if appropriate for known or expected differences between past operational parameters and those anticipated for the next 31 days. ,

l 3.11.1 Liould Dose Proiections The 31-day liquid dose projections are calculated by the following:

a. Determine the total body Dtb and organ dose D,o (Equations 3-1 and 3-2, respectively) for the last typical completed 31-day period. The last typical 31-day period should be one without significant identified operational differences from the period being projected to, such as full power operation vs. periods when the plant is shut down,

b. Calculate the ratio (Rg) of the total estimated volume of batch releases expected to be released for the projected period to that actually released in the reference period,

c. Calculate the ratio (R 2) of the estimated gross primary coolant activity for the projected period to the average value in the reference period.

Use the most recent value of primary coolant activity as the projected value if no trend in decreasing or increasing levels can be determined.

d. Determine the projected dose from:

Total Body: Dt3 ,, - Dtb . R3 . R l

Max. Organ: D ,, ,, - D , . R3 . R The EMS sof tware can also be used to perform monthly projected dose calculations as described in Appendix C. The methodology applied by EMS in projecting liquid doses is outlined in Section 2.7 of Attachment 4 to Appendix C (EMS Technical Reference Manual).

3.11.2 Gaseous Dose Projections For the gaseous radwaste treatment system, the 31-day dose projections are calculated by the following:

a. Determine the gamra air dose DL, (Equation 3-6a), and the beta air dose D[,

(Equation 3-7a) from the last typical 31-day operating period.

X-B.3-29 ODCM Rev. 18 L

3.11 DOSE PROJECTIONS 3.11.2 Gaseous Dose Proiections

b. Calculate the ratio (R 3 ) of anticipated number of curies of noble gas to l be released from the hydrogen surge tank to the atmosphere over the next l 31 days to tha number of curies released in the reference period on l which the gamma and beta air doses are based. If no differences between the reference period and the next 31 days can be identified, set R3 to 1.

l c. Determine the projected dose from:

Gamma Air: Dji, ,, = Djt, . R 3 Beta Air: D{3, ,, = D{3, . R 3 For the ventilation exhaust treatment system, the critical organ dose from iodines, tritium, and particulates are projected for the next 31 days by the following:

l l a. Determine the critical organ dose D , (Equation 3-8a) from the last typical 31-day operating period. (If the limit of Part A Control C.7.4.1.c (i.e. , 0.3 mrem in 31 days) is exceeded, the projected controlled area annual total effective dose equivalent from all station sources should be assessed to assure that the 10CFR20.1301 dose limits to members of the public are not exceeded.)*

b. Calculate the ratio (R 4 ) of anticipated primary coolant dose equivalent I-131 for the next 31 days to the average dose equivalent I-131 level during the reference period. Use the most current determination of DE I-131 as the projected value if no tre* d can be determined,

c. Calculate the ratio (R 3 ) of anticipated primary system leahage rate to the average leakage rate during the reference period. Use the current value of the system leakage as an estimate of the anticipated rate for the next 31 days if no trend can be determined,

d. Determine the projected dose from:

Critical Organ: D , ,, - D , . R. . R3 i The EMS software can also be used to perform monthly projected dose l calculations as described in Appendix C. The methodology applied by EMS in projecting gaseous dose is outlined in Section 3.8 of Attachment 4 to Appendix C (EMS Technical Reference Manual).

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l X-B.3-30 ODCM Rev. 18

3.11 DOSE PROJECTIONS 3.11.2 Caseous Dose Proiections (Continued) l

Note: This action is based on the assumption that tritium is the controlling I nuclide for whole body exposures through the inhalation pathway. Maximum annual average on-site X/Q's for station effluent release points are approximately 100 times the values used for the site boundary dose calculations. However, the site boundary doses calculated by the ODCM for iodines, tritium, and particulates with half lives greater than 8 days, includes all potential off-site exposure pathways. For tritium, the inhalation pathway only accounts for 10% of the total dose contribution being calculated. As a result, if the monthly calculation indicates that l

the site boundary maximum organ dose reached 0.3 area, the on-site maximum

! dose due to inhalat "ould be approximately 3.0 mrom fo: this period. If this were projected antinue for a year with a 2000 hotr occupancy factor l applied, the projectec inhalation whole body dose would be approximately

! 8 mrem, or 8% of the 10CFR20.1301 limit. This is a reasonable trigger value l for the need to consider the dose contribution from all station sources to

! members of the public in controlled areas.

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X-B.3-31 ODCM Rev. 18 t

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4.0 . RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM The radiological environmental monitoring stations are listed in Table B.4-1.

The locations of the stations with respect to the Seabrook Station are shown or,the maps in Figures B.4-1 to B 4-6.

Direct radiation measurements are analyzed at the station. All other radiological analyses for environmental samples are performed at a contractor laboratory. The contractor laboratory participates in the U.S. Environmental Protection Agency's Environmental Intercomparison Studies Program for all relevant 4 species in an aqueous (eater) matrix. An independent vendor (Analytics) supplies the remaining cross check samples. These samples are presented on an air filter and in milk and water matrices.

l Pursuant to Part A Surveillance S.9.2.1, the land use census will be conducted "during the growing season" at least once per 12 months. The growing season is defined, for the purposes of the land use census, as the period from June 1 to October 1. The method to be used for conducting the census will consist of one oc more of the following, e.c appropriate: door-to-door survey, visual inspection from roadside, aerial survey, or consulting with local agricultural authorities.

l Technical Specification 6.8.1.3 and Part A, Section 10.1 of the ODCM requires that the results of the Radiological Environmental Monitoring Program be summarized in the Annual Radiological Environmental Operating Report "in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, 1979."

The general table format will be used with one exception and one clarification, as follows. The mean and range values will be based not upon detectable measurements only, as specified in the NRC Branch Technical Position, but upon all measurements.

This will prevent the positive bias associated with the calculation of the mean and range based upon detectable measurements only. Secondly, the Lower Limit of l Detection column will specify the LLD required by ODCM Table A.9.1-2 for that radionuclide and sample medium. ;

l !

l l

l X-B.4-1 ODCM Rev. 18

TABLE B.4-1 RADIOLOGICAL ENVIRONMENTAL MONITORING STATIONS (*)

i

Distance From i Exposure Pathway Sample Location Unit 1 Direction From and/or Samole and Desirnated Code Containment (km) the Plant ,

1. AIRBORNE (Particulate and Radioiodine)

AP/CF-01 PSNH Barge 2.7 ESE i Landing Area AP/CF 02 Harbor Road 2.7 E AP/CF-03 SW Boundary 0.8 SW AP/CF-04 W. Boundary 1.0 W AP/CF-05 Winnacunnet H.S. <b) 4.0 NNE AP/CF-06 Georgetown 24.0 SSW Substation (Control)

AP/CF 07 PSNH Substation (b) 5.7 NNW AP/CF 08 E&H SubstationUd 3.4 SSE

2. WATERBORNE

a. Surface WS-01 Hampton-Discharge Area 5.3 E WS-51 Ipswich Bay (Control) 16.9 SSE

b. Sediment SE-02 Hampton-Discharge Area th) 5.3 E )

SE-07 Hampton Beach (b) 3.1 E t SE-08 Seabrook Beach 3.2 ESE SE-52 Ipswich Bay (Control)(b> 16.9 SSE SE-57 Plum Island Beach 15.9 SSE (Control)<b)

3. INCES* TION

a. Milk TM-04 Salisbury, MA 5.2 SW TM-08 dampton Falls, NH 4.3 NNW TM-09 Hampton, NH 5.5 NNW TM-15 Hampton Falls, NH 7.0 NW TM-16 Kensington, belch' 7.7 WNW TM-20 Rowley, MA (Control) 16.3 S TM-21 North Andover, MADO 29.0 SW

b. Fish and Invertebrates (*)

FH-03 Hampton - Discharge 4.5 ESE Area FH-53 Ipswich Bay (Control) 16.4 SSE RA-04 Hampton - Discharge 5.5 E Area HA-54 Ipswich Bay (Control) 17.2 SSE MU-06 Hampton - Dischar8e 5.2 E Area MU-09 Hampton Harbor"') 2.6 E MU-56 Ipswich Bay (Control) 17.4 SSE MU-59 Plum IslandR4 15.8 SSE X-B.4 2 ODCM Rev. 18

TAELE B.4-1 RADIO 1DCICAL ENVIRONMENTAL MONITORING STATIONSW (Continued)

Exposure Distance From Direction Froa Pathway and/or Sample Location Unit 1 Direction From Samnie and Desirnated Code Containment (km) the Plant

4. DIRECT RADIATION TL 1 Brimmer's Lane, 1.1 N Hampton Falls TL-2 Landing Rd., Hampton 3.2 NNE TL 3 Glade Path, Hampton 3.1 NE Beach TL-4 Island Path, Hampton 2.4 ENE Beach TL 5 Harbor Rd., Hampton 2.7 E Beach TL 6 PSNH Barge Landing 2.7 ESE Area TL-7 Cross Rd., Seabrook 2.6 SE Beach TL-8 Farm Lane, Seabrook 1.1 SSE TL-9 Farm Lane, Seabrook 1.1 S TL-10 Site Boundary Fence 1.0 SSW TL-11 Site Boundary Fence 1.0 SW TL-12 Site Boundary Fence 1.0 WSW TL-13 Inside Site Boundary 0.8 U TL 14 Trailer Park, Seabroek 1.1 WNW TL-15 Brimmer's Lane, 1.4 NW Hampton Falls TL-16 Brimmer's Lane, 1.1 NNW Hampton Falls TL-17 South Rd., N. Hampton 7.9 N TL-18 Mill Rd., N. Hampton 7.6 NNE TL-19 Appledore Ave.., 7.9 NE N. !!ampton TL-20 Ashworth Ave., 3.4 ENE Hampton Beach ;

TL-21 Route 1A, Seabrook 2.7 SE Beach TL 22 Cable Ave., 7.6 SSE Salisbury Beach TL-23 Ferry Rd., Salisbury 8.1 S TL-24 Ferry Lots Lane, 7.2 SSW Salisbury TL-25 Elm St., Amesbury 7.6 SW TL-26 Route 107A, Amesbury 8.1 WSW X B 4-3 ODCM Rev. 18

TABLE B.4-1 RADIO 1.DGIcAL ENVIRONMENTAL MONITORING STATIONS (e)

(Continued)

! Exposure Distance From Direction From Pathway and/or Sample Location Unit 1 Direction From Samnie and Dahinnated Code Containment (km) the Plant TL-27 Highland St., 7.6 W S. Hampton TL-28 Route 150, Kensington 7.9 WNW TL-29 Frying Pan kne, 7.4 NW Hampton Falls TL-30 Route 27. Hampton 7.9 NNW TL 31 Alumni Drive, Hampton 4.0 NNE TL-32 Seabrook Elementary School 1.9 S TL-33 Dock Area, Newburyport 9.7 S TL-34 Bow St., Exeter 12.1 NW TL-35 Lincoln Ackerman School 2.4 NNW TL-36 R0ute 97, Geor$etown 22.0 SSW (Control)

TL-37 .Plaistow, NH (Control) 26.0 WSW TL-38 .Hampstead, NH (Control) 29.0 W TL-39 Fremont, NH (Control) 27.0 WNW TL-40 Newmarket, NH (Control) 24.0 NNW TL-41 Portsmouth, NH 21.0 NNE (Control)*)

TL-42 .Ipswich, MA (Control)*) 27.0 SSE (a) Sample locations are shown on Figures B.4 1 to B.4 6.

(b) This sample location is not required by monitoring program defined in Part A -

of ODCM; program requirements specified in Part A do not apply to samples taken at this location.

-l (c) Samples will be collected pursuant to ODCM Table A.9.1-1. Samples are not required from all stations listed during any sampling interval (FH = Fish; l HA - Lobsters; MU - Mussels). Table A.9.1-1 specifies that "one sample of three commercially and recreationally important species" be collected in the vicinity of the plant discharge area, with similar species being collected at a control location. (This wording is consistent with the NRC Final Environmental Statement for Seabrook Station.) Since the discharge area is off-shore, there is a great number of fish species that could be considered commercially or recreationally important. Some are migratory (such as striped bass), making them less desirable as an indicator of plant-related radioactivity. Some pelagic species (such as herring and mackerel) tend to j school and wander throughout a large area, sometimes making catches, of l significant size difficult to obtain. Since the collection of all species l would be difficult or impossible, and would provide unnecessary redundancy in terms of monitoring important pathways to man, three fish and invertebrate species have been specified as a minimum requirement. Samples may include marine fauna such as lobsters, class, mussels, and bottom-dwelling fish, such as flounder or hake. Several similar species may be grouped together into one sample if sufficient sample mass for a single species is not available after a reasonable effort has been made (e.g., yellowtail flounder and winter floun, der) .

X-B.4-4 ODCM Rev. 18

l. _ .

e TECURE 5.6 1 RADIO 7ACICAL ENvTRCNMENTAL MONTTCRINC TLCATIONS VTTHIN 4 KIIfMETERS OF SEARROOK STATION N

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' PADTOLOCICAL NkONwtsrfAL MOMTOR9NC TACATIONS BE-*,*rDT 4 KILOMETERS AND 12 KTLOMw nt WOM SEARROOK STATION l

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X-B.4-6 ODCM Rev. 18 1

l FICURE B.4-3 golocICAL ENVIRONMENTAL MONITORINc IDCATIONS

! OUTSIDE 12 KIIAMETERE OF SEABROOK STATION

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X-B.4-7 ODCM Rev. 16 i

FIGURE 5.4-4 DIRECT RADIATION MONITORING IACATIONS UTTHIN 4 KTV.0 METERS OF SEABROOK STATION NNW N NNE A TL-2 NE

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Eb 7 \-35A TL j 3 A TL-3 A TL-4 ENE y o A TL-16 ef L n t3 AR1 +p WNW .

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5 D 1 X-B.4 8 ODCM Rev. 18

FICURE B.4-5 l

DIRECT RADIATION MONITORING LOCATIONS BETWEEN 4 KILOMETERS AND 12 KIIDMETERS FROM SEABROOK STATION H NME N = ,N'NW /

g[ 0 5

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KLCMETERS

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l X-B.4-9 ODCM Rev. 18

I l

FIGURE B.4-6 f

DIRECT RADIATION MONITORING LOCATIONS OUTSIDE 12 KItDMETERS OF SEABROOK STATION i

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l X-B.4-10 ODCM Rev. 18

~

'5.0 SETPOINT DETERMfNATIONS l l Chapter 5 contains the methodology for the calculation of effluent monitor setpoints to implement the requirements of the radtoactive effluent monitoring j l systems Part A Controls C.5.1 and C.S.2 for liquids and gases, respectively. ]

1 l E:rataple setpoint calculations are provided for each of the required effluent monitors.

l 5.1 LIQUID EFFLUENT INSTRUMENTATION SETPOINTS Part A Control C.5.1 requires that the radioactive liquid effluent instrumentation in Table A.5.1-1 of Part A have alarm setpoints in order to ensure that Parc A Control C.6.1.1 is not exceeded. Part A Control C.6.1.1 limits the activity concentration in liquid effluents to the appropriate MPCs in 10CFR20 and a total noble gas MPC.

I 5.1.1 Liauid hzua 'Zgit Tank Monitor (RM-6509)

The liquid waste test tank effluent monitor provides alarm and automatic termination of release prior to exceeding the concentration limits specified in 10CFR20, Appendix B, Table II, Column 2 to the environment. It is also used to monitor discharges from various waste sumps to the environment.

5.1.1.1 Method to Determine the Setooint of the Liould Waste Test Tank Monitor (RM-6509)

The alarm setpoint is based on ensuring that radioactive effluents in liquid l waste are in compliance with Control limits which are based on the concentration limits in Appendix B to 10CF1t20. The alarm point depends on available dilution flow through the discharge tunnel, radwaste discharge flow rate from the test tanks, the isotopic composition of the liquid waste, and the monitor response efficiency and ,

background count rate applicable at the time of the discharge. The alarm / trip setpoint is determined prior to each batch release taking into account current values for each variable parameter. The following steps are used in determining the monitor setpoint:

First, the minimum required dilution factor is determined by evaluating the isotopic analysis of each test tank to be released along with MPC requirements for each radionuclide. The most recent analysis data for tritium and other beta emitters that are analyzed only monthly or quarterly on composite samples can be used as an estimate of activity concentration in the tank to be released. For noble l gases, the Control limit (C.6.1.1) is defined as 2E-04 pCi/ml total for all dissolved and entrained gases. Therefore, I

Dr,t. =

{ or { fg,whicheverislarger. (5 3)

Where:

DF,in - Minimum required dilution factor necessary to ensure that the sum of the ratios for ech nuclide concentration divided by its MPC value is not greater tn.c 1 (dimensionless).

l l

l X B.5 1 ODCM Rev. 18

l 5.1 LfQUID EFFLUENT TNSTRUMENTATION SETPOINTS l 5.1.1.1 Method to Determine the Setooint of the Liould Waste Test Tank Monitor (RM-6509) (Continued)

C3 -

Activity concentration of each radionuclide "i" (except noble gases) determined to be in the test tank (uc i/ml). This includes tritium and other non gamma emitting isotopes either measured or estimated from the most recent composite analysis. 3 C,

n The sum of all dissolved and entrained noble gases identified in l each test tank (pci/ml).

)

MPC i -

The concentration limit (above background) at point of discharge to the environment for radionuclide "i" taken from 10CFR20, Appendix B. Table II, Column 2 (pCi/ml) for all nuclides other than noble gases.

See ODCM, Appendix B, for a listing. In the event that no activity is expected to be discharged, or can be measured in the system, the liquid monitor setpoint should be based on the most restrictive MPC for an " unidentified" mixture or a mixture known ,

not to contain certain radionuclides as given in 10CFR20, j Appendix B, notes, j l 2E-04 -

The total dissolved and entrained noble gas Control concentration limit in liquid effluents from the plant (pci/ml).

Next, the available dilution flow through the discharge tunnel (F d), or a conservative estiutate for it, is divided by the minimum dilution factor (DF,in) to i determine the anximum allowable discharge flow rate (F ) that the test tanks could I be released St without exceeding the MPC limits, assuming no additional radioactive flow paths are discharging at the time of release of the test tanks. Therefore, r-e.

Where:

Fan -

The maximum allowable discharge flow rate from the test tank past l the monitor which would equate to the Control concentration limit for the radioactivity mixture determined to be in the test tank (gpm).

F4 -

The actual or conservative estimate of the flow rate out of the discharge tunnel (gpm).

The selection of the actual discharge flow rate (F.) from the test tanks compared to the maximum allowable discharge rate must satisfy the following:

F, 5 F., x fet X-B.5-2 ODCM Rev. 18

5.1 LIQUID EFFLUENT INSTRUMENTATZON SETPOINTS 5.1.1.1 Method to Determine the Setooint of the Liould Waste Test Tank Monitor (RM-6509) (Continued) where the fs t represents an administrative fraction of the maximum allowable discharge flow from the test tanks. This fraction provides additional margin in meeting MPC limits for non gamma emitters (such as tritium) at the discharge point to the ocean when other flow paths may contribute to the total site release at the time of tank discharges and minimum dilution flow conditions exist.

With the above conditions on discharge and dilution flow rates satisfied, the alarm / trip setpoint for the monitor which corresponds to the maximum allowable concentration at the point of discharge (conservatively assuming that any change in i

the expected gamma activity in a test tank is also reflective of the same change in non gamma emitters, such as tritium) is determined as follows:

R ,,eposar

f X i

y_ y y X{W U*O Where:

R eetptnt - The maximum allowable alarm / trip setpoint for an instrument response (pCi/ml) that ensures the limiting concentration at the point of discharge is not exceeded.

fg - The fraction of the total contribution of MPC at the discharge point to be associated with the test tank effluent pathway, where fa. fs, and f4 are the fractions for the Turbine Building Sump Steam Generator Blowdown, and Primary Component Cooling pathways contributions to the total, respectively (fg+f +fr&f. $ 1). Each of the fractions may be conservatively set administrative 1y such that the sum of the fractions is less than 1. This additional margin can be used to account for the uncertainty in setpoint parameters such as estimated concentration of non gamma emitters that are based on previous composite analyses of the waste stream.

1 5.1.1.2 Liould Vaste Test Tank Monitor Setooint Ex==nle i

The radioactivity concentration of each radionuclide, C i , in the waste test j tank is determined by analysis of a representative grab sample obtained at the i radwaste sample sink, and analyzed prior to release for gamma emitters, or as part of a composite analysis for non gamma emitters. This setpoint example is based on i the following data- I l

i Ci (pCi/ml) MPCi (pCi/al)

Cs-134 2.15E-05 9E-06 Cs-137 7.48E-05 2E-05 Co-60 2.56E 05 3E-05 H-3 1.50E-01 3E-03 The minimum required dilution factor for this mix of radionuclides is:

X-B.5-3 ODCM Rev, 18

5.1 LIQUID EFFLUENT INSTRUMENTATION SETPOINTS 5.1.1.2 Liould Waste Test Tank Monitor setooint Ex==nle (Continued)

C4_ 2.15E-05 , 7.48E-05 2.56E-05 1.50E-01 gg"i" ,

{ MFC,

, , , = 57 9 E-06 2E-05 3E-05 3E-03 The release flow rate (F,) from the waste test tanks can be set between 10 and 150 gpm. The cooling water tunnel discharge dilution flow rate (F ) 4can typically vary from approximately 8,800 to 412,000 gpa depending on the operating status of the plant. In this example, if the dilution flow (F ) is4 taken as 412,000 gpm, the maximum allowable discharge rate (F ) is:

r~ = or.'u

, 412,000 57

= 7228 ppm 1

l Vith the selected release rate from the test tank set at 150 gpa, and the administrative flow fraction (fst) assumed in this example to be 0.7, the condition for the technical Specification concentration limits is met since:

F. (equal to 150) < F,., (equal to 7228 gpa) x ft (set at 0.7) l 150 < 5060 and the monitor response due to the mix of the gamma emitters is:

i Cyt (pCi/ml) j Cs-134 2.15E 05 Cs-137 7.48E-05 l

Co-60 2.56E-05 l I

! {Cyi = 1.22E-04pel/ml Under these conditions, the alarm / trip setpoint for the liquid radwaste discharge monitor is:

R,,,,,3., = fg X y 7.a X { Cyi (5-1) pC1/ml () () pCi/ml X-B.5-4 ODCM Rev. 18

5.1 LIQUID EFFLUENT INSTRUMENTATION SETPOINTS !

5.1.1.2 Liauid Waste Test Tank Monitor Setooint Ers=nle (Continued) 1 R,,egw = 0.4 X '

0x x 1.22E-04

- 2.35E-03 pCi/mi '

In this example, the alarm / trip setpoint of the liquid radwaste dir.c arge l monitor can be put at 2.35E-03 pCi/ml above background. For the example, it is assumed that the test tank release pathway will be limited to only 40% of the total site discharge allowable concentration.

5.1.2 Turbine Buildine Drains Liould Effluent Monitor (RM-6521)

The Turbine Building drains liquid effluent monitor continuously monitors the Turbine Building sump effluent line. The only sources to the Sump Effluent System are from the secondary steam system. Activity is expected in the Turbine Building Sump Effluent System only if a significant primary-to-secondary leak is present. If a primary-to-secondary leak is present, the activity in the sump effluent system would be comprised of only those radionuclides found in the secondary system, with reduced activity from decay and dilution.

The Turbine Building drains liquid effluent monitor provides alarm and automatic termination of release prior to exceeding the concentration limits specified in 10CFR20, Appendix B Table II, Column 2 to the environment. The alarm setpoint for this monitor will be determined using the same method as that of the liquid waste test tank monitor if the total sump activity is greater than 10 percent of MPC, as determined by the most recent grab sample isotopic analysis. If the total activity is less than 10 percent of MPC, the setpoints of RM-6521 are calculated as follows:

High Trip Monitor Setpoint (pCi/ml) - f2 (DF') (" unidentified mix MPC" (pCi/ml)) (5-21) where:

DF" - Circulating water flow rate (gpm)

Flow rate pass-monitor (gpm) unidentified mix MPC - most restrictive MPC value (pCi/ml) for an ,

unidentified mixture or a mixture known not I to contain certain radionuclides as given in 10CFR20, Appendix B, Notes.

fa - 1 - (fx + f3+ f ); where the f values are described above.

In addition, a warning alarm setpoint can be determined by multiplying the high trip alarm point by an administrative 1y selected fraction (as an example, 0.25).

' Warning Alarm ' ' ' '

High Trip Monitor Setpoint = Monitor Setpoint 0.25 i

(pci/ml) , t 4 i 4 X-B.5-5 ODCM Rev. 18

i 5.1 LIQUID EFFLUENT INSTRUMENTATION SETPOINTS 1

l 5.1.3 Steam Generator Blowdown Liauid Samnle Monitor (RM-6519)

The steam generator blowdown liquid sample monitor is used to detect abnormal l

activity concentrations in the steam generator blowdown flash tank liquid discharge.

The alarm setpoint for the steam generator blowdown liquid sample monitor, when liquid is to be discharged from the site, will be determined using the same l approach as the Turbine Building drains liquid effluent monitor.

For any liquid monitor, in the event that no activity is expected to be discharged, or can be measured in the system, the liquid monitor setpoint should be based on the most restrictive MPC for an " unidentified" mixture or a mixture known not to contain certain radionuclides given in 10CFR20, Appendix B notes.

5.1.4 PCCW Head Tank Rate-of-Chance Alarm Setnoint A rate-of-change alarm on the liquid level in the Primary Component Cooling Water (PCCW) head tank will work in conjunction with the PCCW radiation monitor to alert the operator in the Main Control Room of a leak to the Service Water System

, from the PCCW System. For the rate-of-change alarm, a setpoint is selected based on detection of an activity level equivalent to 10-8 pCi/mi in the discharge of the Service Water System. The activity in the PCCW is determined in accordance with the l liquid sampling and analysis program described in Part A, Table A.6.1-1 of the ODCM and is used to determine the setpoint.

The rate-of-change alarm setpoint is calculated from:

RC,,e - 1x10-e , gyp

y g (5-23) mi i (gal) hr ,(pCi) al (gal) hr ( M )

where:

RC .t -

The setpoint for the PCCW head tank rate-of-change alarm (in gallons per hour).

lx10e -

The minimum detectable activity level in the Service Water System I due to a PCCW to SUS leak (pCi/ml).

1 SWF - Service Water System flow rate (in gallons per hour).

PCC -

Primary Component Cooling Water measured (decay corrected) gross l radioactivity level (pCi/ml).

l l As an example, assume a PCCW activity concentration of 1x10-5 pCi/mi with a

! service water flow rate of only 80 percent of the normal flow of 21,000 gpm. The j rate-of-change setpoint is then:

l RC,,e - 1x10-e pCi . 1.0x10e gph (1/1x10-3 pCi) l ml al i RC,,s - 1000 gph i'

X-B.5-6 ODCM Rev. 18

5.1 LIQUID EFFLUENT INSTRUMENTATION SETPOINTS 5.1.4 PCCW Head Tank Rate-of-Channe Alarm Setooint (Continued)

As a result, for other PCCW activities, the RC,,e which would also relate to a detection of a minimum service water concentration of 1x10-e pCi/ml can be found from:

1x10-8 e pCi/ml .1000 gph RC. 6 -

pcc (5-244, 5.1.5 PCCW Radiation Monitor The PCCW radiation monitor will alert the operator in the Main Control Room of a leak to the PCCW System from a radioactively contaminated system.

The PCCW radiation monitor alarm is based on a trend of radiation levels in the PCCW System. The background radiation of the PCCW is determined by evaluating the radiation levels over a finite time period. The alert alarm sett2 int is set at 1.5 x background, and the high alarm setpoint is set at 2 x background, per Technical Specification Table 3.3-6.

l

! 1 l \

i f

X-B.5-7 ODCM Rev. 18

i 5.2 GASEOUS CFFLUENT INSTRUMENTATION SETPOINTS Part A Control C.5.2 requires that the radioactive gaseous effluent I instrumentation in Table A.S.2 1 of Part A have their alarm setpoints set to insure that Part A Control C.7.1.1 is not exceeded.

5.2,1 Plant Vent Wide-Ranne Gas Monitors (RM-6528-1.2 and 3)

The plant vent wide-range gas monitors are shown on Figure B.6-2.

5.2.1.1 Method to Determine the Setnoint of the Plant Vent Wide Rance Gas Monitors (RM-6528-1.2 and 3)

The maximum allowable setpoint for the plant vent wide-range gas monitor (readout response in pCi/sec) is set by limiting the off-site noble gas dose rate to the total body or to the skin, and is denoted R .tytt. R .spt,t is the lesser of:

Rtb - 588 (5-5) a pCi/sec - ( mrem-pCi-m ) ( pCi-yr) 3 yr-pGi-sec mrem-m and:

R,gt, - 3,000 (5-6) pCi/sec - ( "#**) pCi-yr )

yr ( arem-sac) )

where:

I R tb - Response of the monitor at the limiting total body dose rate )

(pCi/sec) 3 588 - 500 (mrem-pCi-m )

(lE+06) (5.5E-07) yr-pGi-sec 500 - Limiting total body dose rate (arem/yr) 1E+06 -

Number of pCi per pCi (pCi/pC1) 1 8.5E-07 -

[X/Q)1, maximum off-site long-term average gamma atmospheric dispersion factor for primary vent stack releases (sec/m 8)

DFB, -

Composite total body dose factor (arem-m8 /pCi-yr)

X B.5-8 ODCM Rev. 18

\

5.3 CASEOUS EFFLUENT INSTRUMENTATION SETPOINTS 5.2 1.1 Method to Determine the Setooint of the Plant Vent Wide Range Gas Monitors (RM-6528-1.2 and 3) (Continued)

, Qi DFB i (5 7) 23 41 Qi

                    -   The release rate of noble gas "i" in the mixture, for each noble gas identified in the off-gas (pci/sec)

DFB i - Total body dose factor (see Table B.1-10) (arem-m 3/pci-yr) ) l Roan - Response of the monitor at the limiting skin dose rate (pci/sec)

                                                                           .           i 3,000       -   Limiting skin dose rate (ares /yr)

DF', - Composite skin dose factor (arem-sec/pCi-yr) j l I l

                    ,    ][ Qi DF'i                                              (5-8)
                            }C D1 DF't        -

Combined skin dose factor (see Table B.1-10) (arem-sec/pCi-yr) The following setpoint example for the plat.t vent wide range gas monitors demonstrates the use of equations 5-5 and 5-6 for determining setpoints. This setpoint example is based on the following data (see Table B.1-10 for DFBi and DF't): Qi DFB t DF~i 8 (pCi) ( aree-m ) (arem-sec) sec pCL-yr pCL-yr i Xe-138 1.03E+04 8.83E-03 1.20E-02 Kr-87 4.73E+02 5.92E-03 1.38E-02 Kr-88 2.57E+02 1.47E-02 1.62E-02 Kr-85m 1.20E+02 1.17E-03 2.35E-03 Xe-135 3.70E+02 1.81E-03 3.33E-03 Xe-133 1.97E+01 2.94E-04 5.83E-04 i i i X-B.5-9 ODCM Rev. 18 j

l 5.2 GASEOUS EFFLUENT INSTRUMENTATION SETPOINTS 5.2.1.1 Method to Determine the Setooint of the Plant Vent Wide Range Gas Monitors (RM-6528-1.2 and 3) (Continued) 1 1 DFB* - (5-7) E Qt I { Qi DFB i (1.03E+04)(8.83E 03) + (4.73E+02)(5.92E-03) l l l + (2.57E+02)(1.47E-02) + (1.20E+02)(1.17E-03)

                           + (3.70E+02)(1.81E-03) + (1.97E+01)(2.94E-04)
                           -    9.83E+01 (pci arem-m 3
                                                        /sec-pci-yr)

5.2.1.2 Plant Vent Wide Ranma Cas Monitor Setooint Evannie l

l { Qi - 1.03E+04 + 4.73E+02 + 2.57E+02

                           + 1.20E+02 + 3.70E+02 + 1.97E+01 l
                           -    1.15E+04 pCi/sec DFB,            -     9.83E+01 1.15E+04
                           -    8.52E-03 (arem-m 8
                                                   /pci-yr)

Rtb - 588 (5-5) l l

(8.52E-03)

                           -    6.90E+04 pCi/sec X-B.5-10                        ODCM Rev. 18 L

5.2. GASEOUS EFFLUENT INSTRUMENTATION SETPOINTS 5.2.1.2 Plant Vent Vide Range Gas Monitor Setooint Examole (Continued) and next; DF' - 1 i (5-8) [ 41 [ Q DF'i i (1.03E+04)(1.20E-02) + (4.73E+02)(1.38E-02)

                              + (2.57E+02)(1.62E-02) + (1.20E+02)(2.35E-03)
                              + (3.70E+02)(3.33E-03) + (1.97E+01)(6.83E-04)
                              -   1.38E+02 (pCi-arem-sec/sec-pCi yr)

1. 36E+02 DF'* 1.15E+04

                - 1.18E-02            (arem-sec/pci-yr)

R,gi,- 3,000 (5-6)

                  - (3,000)         (            )

g

                  - 2.54E+05 Ci/see The setpoint, R,.g,,io,,, is the lesser of Ru, and R, gin.        For the noble gas mixture in this example Ru, is les           than R, min, indicating that the total body dose rate is more restrictive. Theref ore, in this example the plant vent wide-range gas l monitor should be set at no more than 6.90E+04 pCi/see above background, or at some administrative fraction of the above value.

In the event that no activity is expected to be released, or can be measured in the system to be vented, the gaseous monitor setpoint should be based on Xe-133. 5.2 GASEOUS EFFLUEST INSTRUMENTATION SETPOINTS 5.2.2 Waste Gas System Monitors (RM-6504 and RM-6503) Process radiation monitors in the waste gas system provide operational information on the performance of the system before its discharge is combined and diluted with other gas flows routed to the plant vent for release ~to the environment. X-B.5-11 ODCM Rev. 18

5.2 CASEOUS EFFLUENT INSTRUMENTATZON SETPOINTS 5.2.2 Vaste Cas System Monitors (RM-6504 and RM-6503) (Continued) The setpoints for the waste gas system monitors are administrative 1y set as small multiples of tho' expected activity concentration to provide operational control over unexpected changes in gas discharges from the system. Typically, the alert alarm setpoint for both monitors is placed at 1.5 times the expected activity , I concentration passing the monitor, with the high alarm trip set at 2 times the expected concentration flow. Under all conditions, the maximum allowable alarm trip shall not exceed a concentration equivalent to 62.5 uCi/cm8 . This concentration limit, based on system design flow of 1.2 cfa, assures that any release from the waste gas system to the plant vent will not exceed the site boundary dose rate limits of Part A Control l C.7.1.1.a. 5.2.3 Main Condanser Air Evacuation Monitor (RM-6505) The process radiation monitor on the main condenser air evacuation system provides operational information about the air being discharged. The discharge occurs either directly from the turbine building during start up (hogging mode) or through the plant vent during normal operations. This process monitor is also used as an indicator of potential releases from the Turbine Gland Seal Condenser exhaust. Early indications of a potential release (i.e., monitor count rate at twice the normal background) should be evaluated by collecting a grab sample of the exhausts from both the main condenser and the Turbine Gland Seal Condenser. The operational setpoints for the air evs uation monitor are administrative 1y set as small multiples of the expected background response of the detector to provide operational control over unexpected changes in the activity discharged from the system. Typically, the alert setpoint is 1.5 times background, with the high alarm set at 2 times background. I Maximum allowable setpoint determinations assure that the site boundary dose rate l lia'ts of Part A Control C.7.1.1.a will not be exceeded. 3 For a typical air ev. cation detector efficiency of 6.0E+05 cpm-cm /pci, flow rates of 10 and 10,000 ch ior the normal and hogging modes of cperation, respectively, and assuming that all che response is due to thu most restrictive noble gas (Kr-89), the difference between the stack release and ground level release pathway setpoints for the two ; modes of operation (normal power and startup, respectively) are seen to be about j three orders of magnitude. This example also assumes 670 lbs/ hour of steam flow through the Turbine Gland Seal System, 1.5E+07 lbs/ hour of steam flow to the main condenser, and that the Turbine Gland Seal Condenser exhaust flow rate of 1,800 cfm goes directly to the Turbine Building Vents (does not directly pass RM-6505). For these conditions, the maximum allowable alarm should not exceed 3.2E+06 cpm when exhausting to the plant vent (assumes an administrative limit of 70% of the , calculated value to account for potential contributions from the Turbine Gland Seal l Condenser exhaust). Under hogging mode operations, the maxiuum allowable alarm j ahould not exceed 1.4E+02 cpm (assumes an administrative limit of 15% of tb j calculated value to account for potential contributions form the plant vent). The maximum allowable setpoints during startup and normal power operations may be recelculated based on identified changes in detector efficiency, discharge flow rats., radionuclide mix distribution, or administrative apportionment of potential contributions from the plant vent and ground level release points following the l methods identified in Part B, Section 8.5. X-B.5-12 ODCM Rev. 18 !

F 1 6.0 LIOUID AND GASEOUS EFFLUENT STREAMS. RADIATION MONITORS AND RADWASTE TREATMENT SYSTEMS Figure B.6-1 shows the liquid effluent streams, radiation monitors and the appropriate Liquid Radwaste Treatment System. Figure B.6-2 shows the gaseous effluent streams, radiation monitors and the appropriate Gaseous Radwaste Treatment System. For more detailed information concerning the above, refer to the Seabrook Station Final Safety Analysis Report, Sections 11.2 (Liquid Waste System), 11.3 l (Gaseous Waste System) and 11.5 (Process and Effluent Radiological Monitoring and Sampling System). The turbine gland seal condenser exhaust lodine and particulate gaseous releases will be determined by continuously sampling the turbine gland seal condenser exhaust. The noble gas releases will be determined by periodic noble gas grab samples. A ratio of main condenser air evacuation exhaust and turbine gland seal condenset exhaust noble gas will be determined periodically. J I i l i l l \ l X-B.6-1 ODCM Rev. 18

FIGURE B.4-1 LIOUID EFFLUENT STFIAMS . RADIATION MONITORS . AND f RADUASTE TREATMENT SYSTEM AT SEABROOK STATION 1 l MAKEUP STORAGE UMT TANK PAB $ l O M @ q E =

                                               --@                [=mi.       )
 <'-"        >;                         =

l 1 ,_ I 4 sum ! t., m, tw.s sanno 8 i EE 7on. systuu 8 i E,

                                                                     ,omo I                                            i i     l    I     _

g 1 , i " 2 ($ l , cucurses 1 Isnu $ l 8 l t $ 1 G5> "" " ' (- > e,.- ._ e---

 @ h*' " " .                               e = =u s'ia= =====
    "'"". lll"llllll'.

e-- O-- l X-B.6-2 ODCM Rev. 18

{ FIGURE B.6-3 CASEOUS EFFLUENT STREAMS. RADIATION MONIIQEgg RADUASTE TREATMENT SYSTEM AT SEABROOK STATION l I I i CONTAINMENT L~ SU8*C "CCC"G "008 0 " BUILDING i VENT 1LA70as 1 Tumaseg 9 9 9 Vacuuu ) tutchG l l l puup g I # A44C7 tnt MPERd M VACUUW N" r b 5 mas.se conotusem

                                                                                                  %]~
                                                    -g

_ .s

                                /
                                                                                             ,    CL v                 V                                                                       ,

L" ) 4

                -                CONTAse. MENT
                                 -A g:                        _

SLowcowN FLASM TAhK

SASEQUS waste .m0CESSese SYSftu 0 f d p_ d'

                                                                       ,y            n-LJ 95s7 ast I            ,

gap ,

                                               """                          ')    Y 0

9 9 e0 .. i i i CHAAcoALSes ' = l

                                                             ,                0    I i                     l                         Il
                  -                                9
                                                    -i 9-                           -
                                                                                                       =       l

[h m

                                                                                 =
                                                                                     ,3 9 e                                                  euna p
                   %)
                - - = >=                                   -

,,,,, i l M MP4 415L L' " .

C . CMARCOAA MTWL RAI . RADIAftops le0Nrf0M X-B.6-3 ODCM Rev. 18

7.0 RASES FOR DOSE CALCUMfl0N METHODS 7.1 LIQUID RELEASE DOSE CA!EUIATIONS This section serves: (1) to document the development and conservative nature of Method I equations to provide background information to Method I users, and (2) to identify the general equatJ.ons, parameters and approaches to Method II-type dose assessments. Appendix C pro ides the bases for the EMS software which is used to implement the dose and dose r3te calculations indicate (i as Method IA. Method I may be used to show that-the Part A RECP which limit off-site total body dose from liquids (C.6.2.1 and C.6.3.1) have haen met for releases over the appropriate periods. The quarterly and annual dose limits in Part A Control C.6.2.1 are based on the AMRA design objectives in 10CFR50,, Appendix I Subsection II A. l The minimum dose values noted in Part A Control C.(s.3.1 are " appropriate fractions," as determined by the NRC, of the design objective to ensure that radwaste equipment is used as required to keep off-site doses AIARA. Method I was developed such that "the actual exposure of an individual . . is unlikely to be substantially underestimated" (10CFR50, Appendix I). The definition, below, of a single " critical receptor" (a hypothetical or real individual whose behavior results in a maximum potential dose) provides part of the conservative margin to the calculation of total body dose in Method I. Method II allows that actual individuals, associated with identifiable exposure pathways, be taken into account for any given release. In fact, Method I was based on a Method II analysis for a critical receptor assuming all principal pathways present instead of any real individual. That analysis was called the " base case;" it was then recuced to form Method I. The general equations used in the base case analysis are also used as the starting point in Method II evaluations. The base case, the metnod of reduction, and the assumptions and data used are presented below. Tho steps performed in the Method I derivation follow. First, the dose impact to the critical receptor [in the form of dose factors DFLm (ares /pCL)] for a unit activity release of each radioisotope in liquid effluents was derived. The base case analysis uses the general equations, methods, data and assumptions in Regulatory Guide 1.109 (Equations A-3 and A-7, Reference A). The liquid pathwa j-contributing to an individual dose are due to consumption of fish and invertebrates, naar the discharge point. A nominal l shorelina activities, operating plant andflow discharge swimming rate of and boating /sec war used with a mixing ratio o 918 ft 0.10. The mixing ratio of 0.10 corresponds to the mitdaus expected prompt dilution or near-field mixing zone created at the ocean surface directly above the multiport

diffusers. (Credit for additional dilution to tne outer edge of the prompt mixing zone which corresponds to the l'F surface isotherm (mixing ratio .025) can be applied in the Method II cal:ulation for shoreline exposures only since the edge of this isotherm typically does not reach the shoreline receptor points during the tidal cycle. The mixing ratio for equatic food pathways in Method II assessments shall be limited to the same value (0.10) as &pplied in Method I for near-field mixing, or prompt dilution only. X B.7-1 ODCM Rev. 18

7.1 LIQUID RELEASE DOSE CALCULATIONS (Continusd) The requirements for the determination of radiological impacts resulting from releases in liquid effluents is derived from 10CFR50, Appendix I. Section III.A.2 of Appendix I indicates that in making the assessment of doses to hypothetical receptors, "The Applicant may taks account of cny real phenomenon or factors , actually affecting the estimate of radiation exposure, including the characteristics I of the plant, modes of discharge of radioactive materials, physical processes tending to attenuate the quantity of radioactive material to which an individual would be exposed, and the effects of averaging exposures over time during which determining factors may fluctuate." In accessing the liquid exposure pathways that I characterize Seabrook Station, the design and physical location of the Circulating I Water Discharge System needs to be considered within the scope of Appendix I. Seabrook utilizes an offshore submerged multiport diffuser discharger for rapid dissipation and mixing of thermal effluents in the ocean environment. The ! 22-port diffuser section of the Discharge System is located in approximately 50 to 60 feet of water with each nozzle 7 to 10 feet above the sea floor. Water is discharged in a generally eastward direction away from the shoreline through the multiport diffuser, beginning ar a location over one mile due east of Hampton Harbor inlet. This arrangement effectively prevents the discharge pluL (at least to the 1 degree or 40 to 1 dilution isopleth) from impacting the shoreline over the tidal cycle. Eleven riser shafts with two diffuser nozzles each form the diffuser and are spaced about 100 feet apart over a distance of about 1,000 feet. The diffusers are designed to maintain a high exit velocity of about 7.5 feet per second during power operations. Each nozzle is angled approximately 20 degrees up from the horizontal plane to prevent bottom scour. These high velocity jets passively entrain about ten volumes of fresh ocean water into the near field jet mixing region before the plume reaches the water surface. This factor of 10 mixing occurs in a very narrow zone of less than 300 feet from the diffuser by the time the thermally buoyant plume reaches the ocean surface. This high rate of dilution occurs within about 70 seconds of discharge from the diffuser nozzles. The design of the multiport diffuser to achieve a 10 to 1 dilution in the near field jet plume, and a 40 to 1 dilution in the near mixing zone associated with the 1 degree isotherm, has been verified by physical model tests (reference

" Hydrothermal Studies of Bifurcated Diffuser Nozzles and Thermal Backwashing -

Seabrook Station," Alden Research Laboratories, July 1977) . During shutdown periods, when the plant only requices service water cooling flow, the high velocity jet mixing created by the normal circulating water flow at the diffuser nozzles is reduced. However, mixing within the discharge tunnel water volume is significantly increased (factor of about 5) due to the long transit time (approximately 50 hours) for batch waste discharged from the plant to travel the three miles through the 19-foot diameter tunnels to the diffuser nozzles. Additional mixing of the thermally buoyant effluent in the near field mixing zone assures that an equivalent overall 10 to 1 dilution occurs by the time the plume reaches the ocean surface. l X-B.7-2 ODCM Rev. 18 l

7.1 LIQUID RELEASE DOSE CALCULATIONS (Continuad) The dose assessment models utilized in the ODCM are taken from NRC Regulatory Guide 1.109. The liquid pathway equations include a parameter (M,) to account for the mixit ; ratio (reciprocal of the dilution factor) of effluents in the environment at the point of exposure. Table 1, in Regulatory Guide 1.109, defines the point of I exposure to be the location that is anticipated to be occupied during plant I lifetime, or have potential land and water usage and food pathways as could actually l exist during the term of plant ope:.ation. For Seabrook, the potable water and land irrigation pathways do not exist since saltwater is used as the receiving water body for ths circulating water discharge. The three pathways that have been factored into the assessment models are shoreline exposures, ingestion of invertebrates, and fish ingestion. With respect to shoreline exposures, both the mixing ratios of 0.1 and 0.025 are extremely conservative since the effluent plume which is discharged over one mile offshore never reaches the beech where this type of exposure could occur. Similarly, bottom dwelling invertebrates, either taken from mud flats near the shoreline or from the area of diffuser, are not exposed to the undiluted effluent plume. The shore area is beyond the reach of the surface plume of the discharge, and the design of the upward directed discharge nozzles along with the thermal buoyancy of the effluent, force the plume to quickly rise to the surface without affecting bottom organisms. Consequentially, the only assumed exposure pathway which might be impacted by the near field plume of the circulating water discharge is finfish. However, the mixing ratio of 0.1 is very conservative because fish will avoid both the high exit valocity provided by the discharge nozzles and the high thermal temperature difference between the water discharged from the diffuser and the ambient water temperaeure in the near field. In addition, the dilution factor of 10 is achieved l within 70 seconds of discharge and confined to a very small area, thus prohibiting any significant quantity of fish from reaching equilibrium conditions with radioactivity concentrations created in the water environment. The mixing ratio of 0.025, which corresponds to the 1 degree thermal near field mixing zone, is a more realistic assessment of the dilution to which finfish might be exposed. However, even this dilution credit is conservative since it neglects the plant's operational design which discharges radioactivity by batch mode. Batch discharges are on the order of only a few hours in duration several times per week and, thus, the maximum discharge concentrations are not maintained in the environment long enough to allow fish to reach equilibrium uptake concentrations ar assumed in the dose assessment modeling. Not withstanding the above expected driution credit afforded at the 1 degree isothers, all Method II aquatic food pathway dose calculations shall conservatively assume credit for prompt dilution only with an M, - 0.10. When dose impacts from the fish and invertebrate pathways are then added to the conservative dose impacts derived for shoreline exposures, the total calculated dose is very unlikely to have underestimated the exposure to any real individual. , The recommended value for dilution of 1.0 given in NUREG-0133 is a simplistic i I assumption provided so that a single model could be used with any plant design and i physical discharge arrangement. For plants that utilize a surface canal-type ( discharge structure where little entrainment mixing in the environment occurs, a j dilution factor of 1.0 is a reasonable assumption. However, in keeping with the ; ) guidance provided in Appendix I to 10CFR50, Seabrook has determine site-specific mixing ratios which factor in its plant design. i l X B.7-3 ODCM Rev. 18 l l t

l l ! 7.1 LIC ID REIEASE DOSE CALCULATIONS (Continusd) , I j The transit time used for the aquatic food pathway was 24 hours, and for shoreline activity 0.0 hours. Table B.7 1 outlines the human consumption and use factors used in the analysis. The resulting, site-specific, total body dose factors { l appear in Table 9.1-11. Appendix A provides an example of the development of a i Method I liquid dose conversion factor for site-specific conditions at Seabrook. ! 7.1.1 Dose to the Total Body l For any liquid release, during any period, the increment in total body dose l from radionuclide "i" is: ADa - k Qi DFlia rem' (mrem) () (pCi)

                                #Ci (7-1) where:

DFLi u, - Site-specific total body dose factor (mrem /pC1) for a liquid release. It is the highest of the four age groups. See Table B.1-11. I Qt - Total activity (pci) released for radionuclide "i". k - 918/F4 (dimensionless); where F4 is the average dilution flow of the Circulating Water System at the point of discharge from the multiport diffuser (in ft 3/sec). l Method I is more conservative than Method II in the region of the Part A dose limits because the dose factors DFL a used in Method I were chosen for the base i case to be the highest of the four age groups (adult, teen, child and infant) for that radionuclide. In effect each radionuclide is conservatively represented by its own critical age group. l 7.1.2 Dose to the Critical Ornan The methods to calculate maximum organ dose parallel to the total body dose l methods (see Part B, Section 7.1.1) .

l I l [ 1 l X B.7-4 ODCM Rev. 18

7.1 LIQUID RELEASE DOSE CALCUIATIONS (Continuad) , 7.1.2 Dose to the Critical Orran (Continued) l For each radionuclide, a dose factor (area /pC1) was determined for each of seven organs and four age groups. The largest of these was chosen to be the maximum DFL . also includes the external organ dose factor (DFL i .) for that radionuclide. i dose contribution to the critical organ. For any liquid release, during any period, the increment in dose from l radionuclide "i" to tha maximum organ is: l l AD, = k Q. DFL%

                           'mres,                                                    (7-2)

(mrem) () (4C1) , pCi l where: DFL . 3

                 -    Site-specific maximum organ dose factor (mrem /pC1) for a liquid release. See Table B.1-11.

Qt - Total activity (pci) released for radionuclide "i". k - 918/F4 (dimensionless); where F4 is the average dilution flow of the Circulating Water System at the point of discharge from the multiport diffuser (in ft 3/sec). l X-B.7-5 ODCM Rev. 18 l l

l ll1 i l ll1jlli)ji i!II l) , 8

                                                                   "                             1 y                  .            .
                *         )

r t v t r e G R 0 0 0 0 s o R N Y 0 0 0 0 u p I / d e T R 2 2 9 0 ny R M A H 5 5 2 I e . C O ( h s l D B r t y a O e a t w e w n

o ch e

                *         )                                       P    nt           m G          R     0 0   0      0                    r i a s p n
          )     N         Y      0 0   0      0                                     o s    I
                          /                                        a                r t     M         R      8 5   8      0                    e   d    r      i s    M         H        4   2                         l e e              v i     I

( c t h n x W u a t E e S N u o l n N y a E ma e o vh o O

i I w N )
r e t t T h I R
0 0 0 f a A t L. Y 0 0 0 0 eh t T a k / 0 sb t S S p k R 7 t i O H 4 4 0 ct w r o H ( 3 6 1 e o e K n S 3 f nd w O f e o O e E d t P R r ) e a B e R l ei c A h E Y a n c i E L R S w B E / 0 0 0 0 c o m A T R 0 0 0 0 i s s o T o T A E g y s t A r OW T 0 0 0 0 o a a A e P I l w S Z L oh n e

( i t o e Y d ai k A n W d

                   .     )      0  0  0      0                   R a p t u   a H     e    T         R      0  8  7      0 1T       t    R         Y                                       l s i eb       Y A     o    E         /      5  3  1      0                    a e r            e

7. P n V G nh t n 6 BD N K o t n i -

s I ( i o a 7 I EU a g , c M e s B LO t ) 0 0 0 0 R e e - BI p R 0 0 9 0 X AL H Y s s T e S 1 6 6 0 g y o s S c I / nl d t U x F C 2 1 i a a O e K t nl l ( I , a a a f R m t d A * ) i I o u 5 t I t m V - T R 0 0 0 0 s E A Y 0 0 0 0 E d e R E / h O e C oh t F l M K 0 0 0 0 rh t i ( o t w S b f ef R a M o d O T ) e e T R d r n t C A Y o o o a A K / 0 0 0 0 C f i i R c F e I I E 0 0 0 0 r ,tc o E c M ' 0 0 0 0 e e a s G n i w t t r s A e L- u of a S r ( . pN U e 9 m l e f 0 o l s e ) 1 C . a u R Y R 1 m F G . Y 0 0 0 0 1 l 7 s e m A E / 0 0 0 0 a9 n o E C e t 1 a i r L V K 0 0 0 0 d i l F ( i g r y e ( u i el r G D b n o

y m o h

                  .      R      0  0  0      0                   A e               s G         Y      0  0  0      0          r        " c t E         /                              o               e n       l V         C      0  0  0      0        t          ;

D e a K a S s n ( l E , e o u M L r i _ g R D p g t e E E e e E t d n R H H r R G l n l a A u e i f d e h n

A T C I *
l l l

7.3 GASEOUS RELEASE DOSE CALCULATIONS 7.2.1 Total Body Dose Rate From Noble Cases l This section serves: (1) to document the development of the Method I equation, (2) to provide background information to Method I users, and (3) to l identify the general equations, parameters and approaches to Method II-type dose I rate assessments. I l Method I may be used to show that the Part A Cor.trols which limit total body l'.doserate.fromnoblegasesreleasedtotheatmosphere(PartAControlC.7.1.1)has been met for the peak noble gas release rate. 1 Method I was derived from general equation B-8 in Ragulatory Guide ?.109 as follows: ce - 1E46 [X/Q]Y Q, DFB, (7-3) l 3

              'mres'  ,
                        'pC1'   sec  'pCi '    mrem-s yr     ,

pci m3 , sec , pCi-yr , where: [X/Q]' - Maximum off-site receptor location long-term average gamma atmospheric dispersion factor. Qi - Release rate to the environment of noble gas "i" ( Ci/sec). 3 DFB i - Gamma total body dose factor, mrem-m . See Table B.1-10.

                                                               . pCi-yr ,

j (Regulatory Guide 1.109, Table B-1). i l Elevated and ground level gaseous effluent release points are addressed separately through the use of specific [X/Q)y. For an elevated gaseous effluent release poin*. and off site receptor, Equation 7-3 takes the form: i i X B.7-7 ODCM Rev. 18 l

l i ! 7.2 GASEOUS RELEASE DOSE CALCULATIONS I 7.2.1 Total Body Dose Rate From Noble Cases (Continued) i

                ,3<,3 = (lE+06) * (8.5E-07) + [ (Q, . DF8i )                                l i

3

            ' mrem'  ,
                        'pCi'  ,

sec , pCi , mrem-m yr ,pci m 3 sec pCi-yr { l l which reduces to. 6,w,3 = 0.85 [ (Q +i DFB i ) I 3

            ' mrem'  ,

pCi-sec 'gCi' , mrem-m yr #Ci-m 3 sec pCi-yr ! For a ground level gaseous effluent release point and off-site receptor, Equation 7-3 takes the form: i 6,g,3 = (IE+06) + (3.4E-06) * [ (Q i. DFB,) I which reduces to: 1 l tw,3 = 3.4 * [ (Q +i DFB i ) i ( pCi-sec mrem-m3

              'mres'                            _gCi' yr
                     =

pCi-m3 { , sec pCi-yr , 1 l The selection of critical receptor, outlined in Part B Section 7.3 is inherent in the derived M thod I, since the maximum expected off-site long-term average atmospheric dispersion factor is used. The sum of dosos from both plant vent stack and ground level releases must be considered for determination of i Technical Specification compliance. All noble gases in Table B.1-10 should be considered. X-B.7-8 ODCM Rev. 18

7.2 GASEOUS RELEASE DOSE CALCUIAT80NS 7.2.1 Total Body Dose Rate From Noble Cases (Continued) A Method II analysis could include the use of actual concurrent meteorology to assess the dose rates as the result of a specific release. 7.2.2 Skin Dose Rate From Noble Cases This section serves: (1) to document the development of the Method I equation, (2) to provide ba.ckground information to Method I users, and (3) to identify the general equations parameters and approaches to Method II-type dose rate assessments. The methods to calculate skin dose rate parallel the total body dose l rate methods in Part B, Section 7.2.1. Only the differences are presented here. Method I may be used to show that the Part A Controls which limit skin dose rate from noble gases released to the atmosphere (Part A Control C.7.1.1) has been mec for the peak noble gas release rate. The annual skin dose limit is 3,000 aren (from NBS Handbook 69, Reference D, pa5es 5 and 6, is 30 rem /10). The factor of 10 reduction is to account for nonoccupational dose limits. It is the skin dose commitment to the critical, or most limiting, off-site receptor assuming long-term site average meteorology and that the release rate reading remains constant over the entire year. Method I was derived from the general equation B-9 in Regulatory Guide 1.109 as follows: D' - 1.11 D!i, + 3.17E44 Qi (X/Q) DFS i (7-4)

             ' mrem'      ,
                                   ' mrem' ' mrad'         'pCi-yr' Ci    sec    mrem-m3 yr mrad             yr    Ci-sec yr m3   , PCi -yr ,

whe:o: 1.11 - Average ratio of tissue to air absorption coefficients (will convert arad in air to area in tissue). DFS $ - Beta skin dose factor for a semi-infinite cloud of radionuclide "i" which includes the attenuation by the outer " dead" layer of the skin. X-B.7 9 ODCM Rev. 18

P- 3 l ) 7.2 CASEOUS RELEASE DOSE CALCULAT20NS 7.2.2 Skin Dose Rate From Noble Cases (Continued) l l D!,, - 3.17E44 { Q (X/Q) i DF Y i (7-5) i l

               ' mrad'            'pCi-yr'                                        mrad d (Ci) ( see) t yr Ci-sec    .

yr d . pCi-yr DFI - Gamma air dose factor for a uniform semi-infinite cloud of radionuclide "i". Now it is assumed for the definition of (X/Q1) from Reference 8 that: Djinit, = D!i, [X/Q)T/[X/Q] (7-6)

               ' mrad'      ,
                                   ' mrad'        sec          m3 yr                yr            d          sec and Qi      M.54 Q, Ci'
                 -      =
                             'Ci -sec'          ' sci' yr pCi-yr           ,

sec so: y (7-8)- 1 Dat, - 1.11 1E+06 (X/Q)? E Qi

DF i i aren ,

arem pCi see pCi arad-m3,

             , yr ,            %,           ,M, f,                 ,sec,   ,pci-yr,
                              +1E+06 X/Q                   Qi DFS g pCi' 'arem-m3, pCi       see                                                        l
                               ,M, [,                  ,s e c, ,pCL-yr, l

4 t L X B.7-10 ODCM Rev. 18

l l l l 7.2 GASEOUS RELEASE DOSE CALCU1ATIONS l 7.2.2 Skin Dose Rate From Noble Cases (Continued) Substituting atmospheric dispersion factors for an elevated gaseous effluent l release point, Equation 7-8 takes the following form: D,xing,3 = (1.11 e 1E+06

8.5E-07 * { (Qi
DFI)) + (1E+06
8.2E-07 * { (Q
DFS1 )]i i i which yields:

D. king.) = [0.94 { (Qi

DFi)] + [0.82 (Qi
DFS )]

i (7-9a)

         ' mrem', 'pCi-see-ares'     g     'pCi , arem-m8' , pCi see pe 'gCi , mrem-m'   3 yr ,    ,

3 pci-m -arad , ,s e c pCL-yr, ' pCi-m3

                                                                        ,s e c     pGi-yr, defining:

DF[g,3 = 0.94 DFl + 0.82 DFS i (7-10a) l Then the off-site skin dose rate equation for an elevated gaseous effluent release point is: 0.kinc.) " E Q

DF[g,,

i (3-4a) i mrem

                  ,p    pCi , mrem-sec
         , yr ,        ,s e e    pGi-yr ,

For an off-site receptor and a ground level gaseous effluent release point, Equation 7-8 becomes: D,gioc,3 = [1.11*1E+06 *3.4E-06 *[ (Qi+DFI) ] + [1E+M *1.0E-05 * (Qi +DFS3 )] which yields: 0.ktn(s) = [3.8 { (Q

DFi)] + (10 { (Qi
DFS )]

i i (7-9b)

Qi (3.8 DFi + 10 DFS ] i X-B.7-11 ODCM Rev. 18

l J 7.3 CASEOUS RELEASE DOSE CALCULATIONS 7.2.2 Skin Dose Rate From Noble Gases (Continued) defining: DFh:3 = 3.8 DF? + 10 DFSi (7-10b) Then the off-sfte skin dose rate equation for ground level gaseous effluent release points is: D a tnc ,3 = Qt

DFh,3 (3-4b) l The selection of critical receptor, outlined in Part B Section 7.3, is inherent in the derived Method I, as it is based on the determined maximum expected off site atmospheric dispersion factors. All noble gases in Table B.1-10 must be considered.

7.2.3 Critical Orran Dose Rate From Iodines. Tritium and Parriculates With HA,1f-Lives Creater Than Eight Days This section serves: (1) to document the development of the Method I equation, (2) to provide background information to Method I users, and (3) to identify the general equation's parameters and approached to Method II type dose rate assessments. The methods to calculate skin dose rate parallel the total body l dose rate methods in Part B, Section 7.2.1. l Method I may be used to show that the Part A Controls which limit organ dose rate from iodines, tritium and radionuclides in particulate form with half lives l greater than 8 days released to the atmosphere (Part A Control C.7.1.1) has been met fer the peak above-mentioned release rates. The annual organ dose limit is 1500 mrea (from NBS Handbook 69, Reference D, pages 5 and 6). It is evaluated by looking at the critical organ dose commitment to the most limiting off-site receptor assuming long-term site average meteorology. l The equation for 6,, is derived from a form of Equation 3 8 in Part B, Section 3.9 by applying the conversion factor, 3.154E+07 (sec/yr) and converting Q to Q in pCi/sec: i X-B.7 12 ODCM Rev. 18

7.2 CAREOUS RELEASE DOSE CALCU1ATIONS 7.2.3 Critical Orean Dose Rate From Iodines. Tritium and Particulates With Half-Lives Creater Than Einht Days (Continued) 6,,- 3.15E+07 * .(Qi*DFG,,)3 (7-12). mres , see p ', sci , arem

       , yr ,       ,yr,          ,s e cj     K Equation 7-12 is rewritten in the form:

(7-12a) D., = { (Q

DFG[,,) -

i , ares

                  ,g      pCi    ,    arem-sec
       , yr ,            ,s e c,     ,  pci-yr ,

where: DFG[,, = 3.154E+07

DFGi ., (7 13)

        'arem-sec'l , 'sec' , 'aren'    -
        ,  pci-yr J         ,yr,        ,pTI-The dose conversion factor, DFGs., has been developed for both elevated gaseous effluent release points and ground level gaseous effluent release points (DFG.c.3 and DFG ..c 3), respectively. These'dese factors are used to determine i                   i accumulated doses over extended periods and have been calculated with the Shielding Factor (SF) for ground plane exposure set equal to 0.7, as referenced in Regulatory Guide 1.109. In the case of the dose rate conversion factors (DFG' t..c.3 and DFG'i..c,3), the dose conversion factors from which they were derived were calculated with the Shielding Factor (SF) for ground plane exposure set equal ec 2.0, For an off site receptor and elevated effluent release point, the critical       j organ dose rate equation is:

(3-Sa) De c.) " ( (Qi

DFC' 3,.c,3) ares

         , yr ,
                  ,y pCi , arem-sec
                          ,s e c       pci-yr ,

For an'off-site receptor and ground level effluent release point, the critical organ dose rate equation is: X B.7-13 ODCM Rev. 18

l 7.2 CASEOUS RELEASE DOSE CALCULATIONS l 7.2.3 Critical Orran Dose Rate From Iodines. Tritium and Particulates With Half-Lives Greater Than Eirht Days (Continued) j 6,,,,3 = { (Qi DFG' g,,c,3) (3-5b) mrem

                      ,g    pCi , mrem-sec
             , yr ,        ,s e e   pci-yr ,

l The selection of critical receptor, outlined in Part B, Section 7.3 is inherent in Method I, as are the expected atmospheric dispersion factors. In accordance with the Basis Statement 3/4.11.2.1 in NUREG-0472, and the l base's section for the organ dose rate limit given for Part A Control C.7.1.1 a Method II dose rate calculation, for compliance purposes, can be based on restrieting the inhalation pathway to a child's thyroid to less than or equal to 1,500 mrem /yr. Concurrent meteorology with time of release may also be used to . assess compliance for a Method II calculation. 7.2.4 Cam == Dose to Air From Noble Cases This section serves: (1) to document the development and conservative nature of Method I equations to provide background information to Method I users, and (2) to identify the general equations, parameters and approaches to Method II-type dose assessments. l Method I may be used to show that the Part A Control C.7.2.1 which limits off-site gamma air dose from gaseous effluents has been met for releases over l appropriate periods. This Part A Control is based on the objective in 10CFR50, Appendix I, Subsection B.1, which limits the estimated gamma air dose in off-site unrestricted areas. NUREC/CR 2919 presents a methodology for determining atmospheric dispersion factors (CHI /Q values) for intermittent releases at user specified receptor locations (intermittent releases being defined as teleases with durations netween 1 and 8,760 hours). The CHI /Q values for intermittee. releases are determinec by linearly interpolating (on a log-log basis) between an hourly 15-percentile OUf/Q value and an annual average CHI /Q value as a function of release duration. nis methodology has been adopted to produce a set of time-dependent atmospheric dispersion factors for Method I calculations. For any noble gas release, in any period, the increment in dose is taken frota Equations B-4 and B-5 of Regulatory Guide 1.109 with the added assumption. chat D'rtnite - DT [X/Q)?/[X/Q): X-B.7-14 ODCM Rev. 18

7.2 CASEOUS RELEASE DOSE CALCULATIONS 7.2.4 G==== Dome to Air From Noble Cases AD't, - 3.17E+4 [X/Q)5 { Qi DFl (7-14) I PCi-yr sec mrad-m3 (mrad) . (C1)

                      ,E e c,     ,7,           ,pGi-yr.

where: l 3.17E+04 - Number of pCi per Ci divided by the numbe of seconds per year. [X/Q)1 - Annual average gamma atmospheric dispersion factor for the receptor location of interest. Qi - Number of curies of noble gas "i" released. DFT i - Gamma air dose factor for a uniform semi-infinite cloud of radionuclide "1". Incorporating a unitiess release duration adjustment term t'* (where "a" is a constant and "t" is the total release duration in hours), and the conversion factor for Ci to pCi (to accommodate the use of a release rate Q in pCi), and substituting the 1 hour gamma atmospheric dispersion factor in place of the annual average gamma atmospheric dispersion factor in Equation 7-14 leads to:

                                                                                          ~(3-6)

D't, - 3.17E-02 * [X/Q]'a, e t-* * (Qi

DFi)

P Ci-yr ' Isac' 3 (arad) .

                      ,ci-sec, p
                                       ,m a
                                               *{   'u Ci e arad-m' pCL-yr, For an elevated release, the equation used for an off-site receptor is:

{g, pp Dair(e) - 3.17E-02 * [1.0E-05]

t-o.27s .

X-B.7-15 ODCM Rev. 18

7.2 CASEC 21S REi. EASE DOSE CALCUIAT20NS

7. 2.4 Ga-= Dose to Air From Noble Gases which leads to:

D!i,c,3 - 3.2E-07 + t-0 275 * { (Q

DFJ) ( 6a) 3 (mrad) - 'pci-yr'*{

p 3 pCi e mrad-m'3 pGi-yr,

                      , C 1 -m ,                                                             ;

For a ground-level release, the equation used for an off-site receptor is: ' 7 D,gr(g) - 3.17E-02 * [4.9E-05) e t-o.2e3

fQi*DFf which leads to:

D]i,<,3 - 1.6E-06

t-o.2s3 . { {gt . pp7) (M) pCi
mrad-m'8 (mrad) - 'pC1,-yr7 pCi- * { pGi-yr The major difference between Method I and Method II is that Method II would use actual or concurrent meteorology with a specific noble gas release spectrum to determine [X/Q)? rather than use the site's long-term average meteorological I dispersion values.

7.2.5 Beta Dose to Air From Noble Cases This section serves: (1) to document the development and conaervative nature of Method I equations to provide background information to Method I users, and (2) to identify the general equations, parameters and approaches to Method II-type dose assessments. l Method I may be used to show that Part A Control C.7.2.1, which limits off-site beta air dose from gaseous effluents, has been met for releases over ; l appropriate periods. This Part A Con *rol is based on the objective in 10CFR50, ' Appendix I, Subsection B.1, which limits the estimated beta air dose in off-site unrestricted area locations. For any noble gas release, in any period, the increment in dose is taken from Equations B 4 and B-5 of Regulatory Guide 1.109: X-B,7 16 ODCM Rev. 18

7.2 GASEOUS RELEASE DOSE CALCUMTIONS I 7.2.5 Beta Dose to Air From Noble Cases (Continued) l AD{i, - 3.17E-02 X/Q { Q DFf i (7-15) A l , , P C i-yr 'sec' rarad-m3, (mrad) - ( Ci) p

                     .Gi-sec     s  ,7,                 pGi yr.

where: DF{

                  -          Beta air dose factors for a uniform semi-infinite cloud of radionuelide "i".

Incorporating the term t-" into Equation 7-15 leads to: D{i, - 3.17E-02

X/Q
t-*

                                                    * { (Qi

Dd) (3-7) i 3

P C i-yr sec pCi

mrad-m (arad) .

                                       ,7, * ( ) * {

p pGi-yr

                     .Gi-sec.                                 .              -

Where X/Q - average 1-hour undepleted atmospheric dispersion factor. For an elevated release, the equation used for an off-site receptor is: D{irc.3 - 3.17E-02

1. 3E-05
t-8 8 * (Qi*Dd)

PCi-yr s_e c pCi * "#8d~" (mrad) - ,

                                                 *()*{

p

                     ,CL-sec,             mI ,
                                                              ,      PCi-yr, which leads to:

D{irc.3 - 4.1E-07

t-0 8 * (Qi
Dd) (3-7a)

(arad) . PC i-yr'3 * ( )

8' Ci
mrad-m

                      ,pci-m ,                              PC1-yr, For a ground-level release, the equation used for an off-site receptor is:

l I X-B.7-17 ODCM Rev. 18

7.2 ' CASEOUS RELEASE DOSE CALCUIATIONS 7.2.5 Beta Dose to Air From Noble Games (Continued) D{t,cp - 3.17E-02

1.9E-04
t-8 818 * (Qi
DFT)

Ci-yr ' 'sec' 8 (arad) . P ,

                                             * ( )

E sci e erad-m '

pct-sec, { , pci-yr, i which leads to: D{i,cp - 6.0E-06

t-o.sts '* (Q*DFT) i

            . (arad) - 'PCi-yr'                 Ci e arad-m' 8

pCi-m3,

                                    *()*{              Pc1-yr, 7.2.6    Dome to critical or mn From Iodines. Tritium and Particulates With Half-Lives Creater Than Eirht Days This section serves: (1) to document the development and conservative nature of Method I equations to provide background information.to Method I users, and (2)
    -to identify the general equations, parameters and approaches to Method II-type dose assessments.

Method I may be used to show that the Part A Controls which limit off-site organ dose from gases (C.7.3.1 and C.8.1.1) have been met for releases over.the appropriate periods. Part A Control C.7.3.1 is based on the AIARA objectives in 10CFR50, Appendix I',-Subsection II C. Part A Control C.8.1.1 is based on Environmental Standards for Uranium Fuel Cycle in 40CFR190, which applies to direct

   . radiation as well as liquid and gaseous effluents. There methods apply only to iodine, tritium, and particulates in gaseous effluent contribution.

Method I was developed such that "the actual exposure of an individual ... is

   .unlikely to be substantially underestimated" (10CPR50, Appendix I). The use below of a single " critical receptor" provides part of the conservative margin to the calculation of critical organ dose in Method I. Method II allows that actual individuals, associated with identifiable exposure pathways, be taken into account for any given release.         In fact, Method I was based on a Method II analysis of a critical receptor assuming all pathways present. That analysis was called the " base case"; 'it was then reduced to form Method I.              The base case, the method of reduction, and_the assumptions and data used are presented below.

X-8.7-18 ODCM Rev. 18

7.2 GASEOUS RELEASE LOSE CALCULAT10NS i 7.2.6 Dose to Critical Ornan From Iodines. Tritium and Particulates With Half-Lives Greater Than Eirht Days The steps performed in the Method I derivation follow. First, the dose impact to the critical receptor (in the form of dose factors DFG ., (mrem /pCi)) for a unit i activity release of each iodine, tritium, and pari:iculate radionuclide with half lives greater than eight days to gaseous effluents was derived, Six exposure pathways (ground plane, inhalation, stored vegetables, leafy vegetables, milk, and meat ingestion) were assumed to ex.ist at the site boundary (not over water or marsh areas) which exhibited the highest long-term X/Q. Doses were then calculated to six organs (bone, liver, kidney, lung, GI-LLI, and thyroid), as well as for the whole body and skin for four age groups (adult, teenager, child, and infant) due to the seven combined exposure pathways. For each radionuclide, the highest dose per unit activity release for any organ (or whole body) and age group was then selected to become the Method I site-specific dose factors. Tte base case, or Method I analysis, uses the general equations methods, data, and assumptions in Regulatory Guide 1.109 (Equation C-2 for doses resulting from direct exposure to contaminated ground plane; Equation C-4 for doses associated with inhalation of all radionuclides to different organs of individuals of different age groups; and Equation C-13 for doses to organs of individuals in different age groups resulting from ingestion of radionuclides in produce, milk, meat, and leafy vegetables in Reference A). Tables B.7-2 and B.7-3 outline human consumption and environmental paraineters used in the analysis. It is conservatively assumed that the critical receptor lives at the " maximum off-site atmospheric dispersion factor location" as defined in Section 7.3. The resulting site-specific dose factors are for the maximum organ which { combine the limiting age group with the highest dose factor for any organ with each nuclide. These critical organ, critical age dose factors are given in Table B.1 12. Appendix A provides an example of the development of Method I gaseous dose conversion factor for site-specific conditions at Seabrook. For any iodine, tritium, and particulate gas release, during any period, the increment in dose from radionuclide "i" is: AD ie - QiDFG., i (7-16) where DFG i , is the critical dose factor for radionuclide "i" and Qi is the activity of radionuclide "i" released in microcuries. j Applying this information, it follows that the general form for the critical organ dose equation is: l X B.7-19 ODCM Rev. 18

s 7.2 GASEOUS RELEASE DOSE CALCUIXIIONS 7.2.6 Dosa'to critical Orean Fros Iodines. Tritium and Particulates With' Half-Lives Greater Than Elaht Dava (Continued) D , - (X/Q){T[/(X/Q)'gl e t** * (Qg

DFG3 ,,) (3 8) ares' ares - 'sec'/sec'*()*E

                     ,7, ,7                901
                                                  *7   .

Substituting specific values associated with the maximum off-site receptor location and elevated release condition yields: D ,g,3 - (1.12E-05)/(7.55E-07)

t-o.as7 * (Qt
DFGs .c 3) which reduces to:

D,,c,3 - 14. 8

t-o.as7 * (Qt
DFGi ..g.3) (3-8a)

For the maximum off-site receptor location and ground-level release conditions, the equation is:

          - D,,g,3 - (1.71E-04)/(9.64E-06) e t*0 318 *     (Qg

DFGt ,,g,3) which reduces to:

          ' D,,(,3 - 17. 7

t-o,31s * (Qg
DFGg .c 3) (3-8b) 7.2.7 Snacial Racantor Caseous Releasa Desa Calculatione l Part A Section 10.2 requires that the doses to individuals involved in recreational activities within the site boundary are to be determined and reported in the Annual Radioactive Effluent' Release Report.

The gaseous dose calculations for the special receptors parallel the bases of l_ the gaseous dose rates and dosas in Part B, Sections 7.2.1 through 7.2.5. Only the differences are presented here. The special receptor XQs are given in Table B.7-5. 7.2.7.1 Total Body Dome Rata From Noble Casas Method I was derived from Regulatory Guide 1.109_as follows: 1 X-B.7-20 ODCM Rev. 18

7.2' CASEOUS RELEASE DOSE CALCUIATIONS l 7.2.7.1 Total Body Dose Rate From Noble Cases (Continued) ba - 1E+06 [X/Q]5 Qi DFB i (7-3) I General Equation (7-3) is then multiplied by an Occupancy Factor (OF) to j account for the time an individual will be at the on-site receptor locations during ! the year. There are two special receptor locations on-site. The " Rocks" is a boat j landing area which provides access to Browns River and Hampton Harbor. The Seabrook i Station UFSAR, Chapter 2.1, indicates little boating activity in either Browns River or nearby Hunts Island Creek has been observed upon which to determine maximum or conservative usage factors for this on site shoreline location. As a result, a default value for shoreline activity as provided in Regulatory Guide 1.109, Table E-5, for maximum individuals was utilized for determining the " Rocks" occupancy factor. The 67 hours / year corresponds to the usage factor for a teenager involved l in shoreline recreation. This is the highest usage factor of all four age groups l listed in Regulatory Guide 1.109, and has been used in the ODCM to reflect the ! maximum usage level irrespective of age. Regulatory Guide 1.109 does not provide a maximum individual usage factor for activities similar to those which would be associated with the Seabrook Station Science & Nature Center. Therefore, the usage factor used in the ODCM for the

 -Science & Nature Center reflects the observed usage patterns of visitors to the facility. Individuals in the public who walk in to look at the exhibits on display and pick up available information stay approximately 1.5 hours each. Tour groups who schedule visits to the facility stay approximately 2.5 hours. For conservatism, it was assumed tnat an individual in a tour group would return five times in a year, and stay 2.5 hours on each visit. These assumptions, when multiplied together, provide the occupancy factor of 12.5 hours / year used in the ODCM for public activities associated with the Science & Nature Center.

For the Science & Nature Center, and the " Rocks", the occupancy factors (OFs) are: Science & Nature Center "1 5 hrs /yr'" -0.0014 T/60 hrs /yr The " Rocks" - 67 hrs /yrtu - 0.0076 5760 hrs /yr l l (O Taken from Seabrook Station Technical Specifications (Figure 5.1-1). X-B.7-21 ODCM Rev. 18

f-7.2 GASEOUS RELEASE DOSE CALCUIATIONS t '7.2.7.1 Total Body Dose Rate From Noble Cases (Continued) l substituting in the annual average gamma X/Qs: i (X/Q]5 - 1.1E-06 sec/m3 (Science & Nature Center) for primary vent stack l releases.

                       - 5.3E-06 seem3 (Science & Nature Center) for ground level releases.
                       - 5.0E-06 sec/m3 (The " Rocks") for primary vent stack releases.
                       - 2.6E-05 sec/m3 (The " Rocks") for ground level releases and multiplying by:

OF - 0.0014 (Science & Nature Center)

                 - 0.0076 (The " Rocks")

l gives: Dtbet ) = 0. 0015 * (Qi

DFBi ) (area /yr) (3-3c) 6 thz(s) = 0.0074 * (Qi
DFB3 ) (ares /yr) (3-3d) btbate) = 0.038 * (Qi
DFBi ) (ares /yr) (3-3e)

Debats) = 0.2 + { (Qi

DFB ) i (area /yr) (3-3f)

  .where:

btbute), bebsts) , bebat), and 6 tbm(s) - total body dose rates to an individual at the Science & Nature Center and the

                                                            " Rocks" (recreational site),

respectively, due to noble gases in an elevated (e) and ground level (g) release. Qi and DFBi are as defined previously. 7.2.7.2 Skin Dose Rate From Noble Cases Method I was derived from Equation (7-8): 0,mi - 1.111E+06 [X/Q]5 { Q DFi + i (7-8) , l 1E+06 X/Q Qi DFS i l X-B.7-22 ODCM Rev. 18

7.2 CASEOUS RELEASE DOSE CALCUIATIONS 7.2.7.2 Skin Dose Rate From Noble Cases (Continued) substituting in the annual average gamma X/Qs: [X/Q]' - 1.lE-06 sec/m 8(Science & Nature Center) for primary vent stack releases.

                  - 5.3E-06 sec/m8 (Science & Nature Center) for ground level release           ,

points.

                  - 5.0E 06 sec/m8 (The " Rocks") for primary vent stack releases.
                  - 2.6E-05 ser ' a (The " Rocks") for ground level release points.

and the arnual average undepleted X/Qs: 1 X/Q - 1.6E-06.sec/m3 (Science & Nature Center) for primary vent stack releases.

              - 2.3E-05 sec/m3 (Science & Nature Center) for ground level release points.

l

              - 1.7E-05 sec/m3 (The " Rocks") for primary vent stack releases.
              - 1.6E-04 sec/m8 (The " Rocks") for ground level release points.

and multiplying by: OF - 0.0014 (Science & Nature Center)

            - 0.0076 (The " Rocks")

gives: b,gggg,3 = 0.0014 Qi (1.22 DFZ + 1.60 DFS g ] for an elevated release point. 6,gggg,3 = 0.0014 Qi (5.88 DFI + 23 DFSg] for a ground level release point. 6,gigg,3 = 0.0076 4g (5.55 DFl + 17.0 DFSg] for an elevated release point. 6,gigg,3 = 0.0076 Qi (28.9 DFI + 160 DFS g ] for a ground level release point. and the equations can be_ written: D. misc.) = 0. 0014 * (Qg

DF'ac.3) i (3-4c)

X-B.7-23 ODCM Rev. 18

r 7.2 GASEOUS RELEASE DOSE CALCULATIONS } .7.2.7.2 Skin Dose Rate From Noble Cases (Continued) l bkint(s) = 0.0014 * { (Qi

DF'gg,3) i (3-4d) i b,gigg,3 = 0.0076 * { (Qi
DF'at ))

i (3-4e) A b,wigg,3 = 0.0076 * (Qi

DF'ac,3) i (3-4f) where:

b.ktnet.). bkinz(s)* baktet.)' . and 6.gigg 3 - the skin dose rate (mrem /yr) to an individual at the Science & Nature Center and the " Rocks", respectively, due to noble gases in an elevated (e) and I ground level (g) release, i Qi - defined previously, and DFirt ), DF[gg,3, DIat.)and t DIgg i 3 - the combined skin dosa factors for radionuclide "i" fot ,he Science & Nature Center and the " Rocks", respectively, for elevated (e) and ground level (g) release points (see Table B.1-13). 7.2.7.3 Critical Orman Dose Rate From Iodines. Tritium and Particulates With Half-Lives Creater Than Eiaht Days l The equations for 6 are derived in the same manner as in Part B, Section 7.2.2, except that the occupancy factors are also included. Therefore: Dcort.)

0.0014 * (Qi
DFG$ cort.)) for an elevated release. (3-Sc) be .gg,3 = 0.0014 * { (Q i* DFG$,,gg,3) for a ground level release. (3-5d) i beant.) = 0.0076 * ${ (Qi
DFG$eont.)) for an elevated release. (3-Se) 1 l

l I X-B.7-24 ODCM Rev. 18

f 7.2 GASEOUS RELEASE DOSE CALCULATIONS 7.2.7.3 Critical Organ Dose Rate From Iodines. Tritium and Particulates With Half-Lives Greater Than Eirht Days (Continued) beents) = 0.0076 * ][ (Qi

DFGl ,at,3) for a ground level release. (3-5f) i where:

b ea st.). beor(s). b ea nt.), and 6 eo n(s) - the critical organ dose rates

(ares /yr) to an individual at the Science & Nature Center and the

                                                          " Rocks", respectively, due to ivuine, tritium, and particulates in elevated (e) and ground level (g) releases, Qi - as defined previously, and DFG$corte), DFG[. gc,3, DFG$conce). and DFGIcoat.)        - the critical organ dose rate factors for radionuclide "1" for the Science & Nature Center and      J the " Rocks", respectively, for      j elevated (e) and ground level (g) release points (see Tables B.1-14 and B.1-15).                         j l

7.2.7.4 G---- Dose to Air From Noble Cases Method I was derived from Equation (3-6): Dji, - 3.17E-02 * [X/Q)72 ,

t-*

                                                 * { (41

DFl) (3-6) ;

i l where all terms of the equation are as defined previously. Incorporating the specific 0F and the atmospheric dispersion factor, the gamma air dose equation for the Science & Nature Center for elevated releases: D!i, c.3 - 3.17E-02 + 1.1E-05 t-o.252 , 0.0014 * (Qt

DF])

which reduces to: l l f-f I l X-B.7-25 ODCM Rev. 18

+ 7.2 GASEOUS RELEASE DOSE CALCU1ATIONS

7.2.7.4 C=--- Dose to Air From Noble Cases (Continued) Dlts:(*) = 4. 9E-10

t-o.252 * { (Qi
DFj) (3-6c)

(arad) . PC i-yr'3 * ( )*{ wCi e arad-m3'

                      ,9Ci-m ,                      PCI-yr, For ground level releases, the gamma air dose equatiari for the Scier. e & Nature Csnter becomes:

Djirct(s) - 3.17E-02 e 1.0E-04 t-o.321

0.0014 * { (Qi
DFJ) 1 which reduces to:

Dji,gg,3 = 4.4E-09

t-o.321 * { (Q
DFJ) i (3-6d)

(arad) . P Ci-yr' mrad-

                      ,pci-m3,  * ( ) * { (Ci* F y[!,

Incorporating the specific 0F and atmospheric dispersion factors for the

 " Rocks" yields the Samma air dose equation for elevated releases:

D] irate) - 3.17E-02

2.1E-05
t-o.iss
0.0076 * { (Qi
DFI) which reducts to:

Dji,,c.3 = 5.1E-09

t-0 135 * (Qi
DFI) (3-6e)

(arad) . Ci-yr' "#*b"

                                * ( )*{ uCi

PCi-yr, P

                      ,pci-m3,              ,

For ground-level releases, the gamma air dose equation for the " Rocks" becomes: D!trats) - 3.17E-02

1.7E-04 t-o.zo'
0.0076 * { (Q
DF])

i which reduces to: i X B.7-26 ODCM Rev. 18 i

7.2 CASEOUS REI. EASE DOSE CALCU1ATIONS 7.2.7.4 C=-- Dose to Air From Noble Cases (Continued) ! l Dli,gg,3 - 4.1E-08

t-o.20' * { (Qi *DFi) (3-6f)

(arad) - pC P i-yr' * ( )* Ci * "#*d~" 3 PGi-yr

                        ,CL-m    s            >                 -

7.2.7.5 Beta Dose to Air From Noble Cases l Method I was derived as described in Part B Section 7.2.5. The general form of the dose equation is: D{3, - 3.17E-02

X/Qy"df
t-* * { (Qi
Dfi) (3-7) l l

l where all terms in the aquation are as defined in Part 5, Secticn 7.2.5. Incorporating the specific 0F and atmospheric dispersion factor for elevated release 6 into Equation 3-7 yields the following beta dose equation for the Science & Nature Center: D{tret.) - 3.17E-02

4.0E-05
t-0 35
0.0014 *a { (Qi
DFf) l which reduces to:

D{i,gg,3 - 1.8E-09

t-0 35 * { (Q
Dfi) i (3-7c)

(arad) - P Ci-yr' Ci

mrad-o f

                                   * ( )*{             pGi 's : ,

pCi-m', For ground-level releases, the beta air dose equation for the Science & Nature Center becomes: l di,gg,3 - 3.17E-02

5.5E-04
t-0 3'7
0.0014 * { (Q
Dfi) i I

which reduces to: l l l X-B.7-27 ODCM Rev. 18 I

r 1 7.2 CASEOUS RELEASE DOSE CALCULATIONS l 7.2.7.5 Beta Dose to Air From Noble Cases (Continued) D{i,gcp - 2.4E-08

t-0 3'? * [ (Qi
Dd) (3-7d)

{ l 3 PCi-yr (mrad) - pci-m3, ( )*{ uCi

arad-m pci-yr, Incorporating the specific 0F and atmospheric dispersion factors for the

   " Rocks" yields the beta air dose equation for elevated releases:            .

D{trat*) - 3.17L-02

1.6E-04
t-o.zes
0.0076 * (Q
DFf) i which reduces to:

I l D{irac.3 - 3.9E-08 + t-o.24s , { (qt pg) (3,7,) f (mrad) 'PCi-yr' * ( )* Ci

arad-a

                        ,ci-m p    3
                               ,                    pCL-yr, For greund-level releases, the beta air dose equation for the " Rocks" becomes:

l D{trats) - 3.17E-02

1.9E-03
t-o.ast
0.0076
t{ (Qi
DF{}

which reduces to: D{irag,3 - 4.6E-07

t-o.as7 * { (Qi
Dd) (3 7f)

                        'PCi-yr'             pCi

mrad-m3 '

(mrad) . 3

                        ,pci-m ,
                                  *()*{             pGi-yr, 7.2.7.6 Critical Orean_ Pose From Iodines. Tritium e:nd Particulates With Half-Lives Creacet Than Einht Days l          Method I was derived as described in Part 3, Section 7.2.3.      The critical Organ Dose equations for receptors at the Science & Nature Center and the " Rocks" were derived from Equation 3-8. The following general equation incorporates (i) a X-B.7-28                        ODCM Rev. 18

7.2 CASEOUS RELEASE DOSE CALCUIATIONS I l 7.2.7.6 critical Orman Dose From Iodines. Tritium and Particulates With Half-Lives Greater Than Eight Davs (Continued) l l ratio of the average 1-hour depleted atmospheric dispersion factor to the average l annual depleted atmospheric dispersion factor, (ii) the unitiess t** term, and (iii) l the OF: D,, - (X/Q)U[j/(X/Q)d*Pl e t**

OF
1{ (Q
DFGm) i (mrem) - / *( )*( )*{ Ci
l Applying the Science & Nature Center-specitic factors for elevated release l conditions produceu the equation:

Deert*> - (3.72E-05)/(1.56E-06)

t-0 3*8
0.0014 * { (Q e DFG ,,g,3) i i 1

which reduces to: Deorte) - 3. 3E-02

t~0 3'8 * (Q,
DFGgeog,3) (3-8c)

(arem) - ( )*( )*{ pCi

For a ground-level release, the equation for a receptor at the Science & Nature Center is ,

i l D,,gg,3 - (5.21E-04)/(2.23E-05) e t-0 3'?

0.0014 * { (Q *i DFG ,,c,3) i which reduces to:

l D,org,3 - 3.3E-02

t-0 3'? * { (Qi
DFGi ,g,3) (3-8d) 1 (area) - ( )*( )*{ Ci e l

l The specific Critical Organ Dose equation for a receptor at the " Rocks" under elevated release conditions is: ) X-B.7-29 ODCM Rev, 18

f 7.2 CASEOUS RELEASE DOSE CALCULATIONS 7.2.7.6 critical Ornan Dose From Iodines. Tritium and Particulates With Half-Lives Greater Than Eight Days (Continued) D,ong,3 - (1.54E-04)/(1.61E-05) e t-o.zas e 0.0076 * [ (Q i* DFG ,g,3) i which reduces to i s Deont.) - 7. 3E-02

t*8 8'8 * (Qi
DFCi ,,g,3) (3-8e)

(aren) - ( )*( )*{ Sci

For a ground-level release, the equation for a receptor at the " Rocks" is: 1 D, oats) - (1.80E-03)/(1.59E-04) e t-o.2e7
0.0076 * { (Q
DFG i. g,3) i which reduces to:

1' D,.ag,3 - 8. 6E-02

t-o.as7 * [ (Q
DFG .,c,3) i i (3 8f)

(area) - ( )*( )*{ pCi

The special receptor equations can be applied under the following conditions (otherwise, justify Method I or consider Method II):

1. Nermal operations (nonemergency event) .

2. Applicable radionuclide releases via the station vents to the atmosphere.

If Method I cannot be applied, or if the Method 1 dose exceeds this limit, or if a more refined calculation is required, then Method II may be applied. X-B.7-30 ODCM Rev. 18

l3 8 1 d . t s . . . 0 . . . . r 2 0 0 0 0 0 0 v o 4 8 4 4 6 5 e t 2 4 1 4 1 R t S 3 1 2 1 a M e 0 0 C M e 7 . 5 D r . 0 . O u . . t 0 0 0 0 0 O 0 0 1 s 4 8 4 2 5 2 4 1 7 a 3 P 1 d . e . . . 0 . . . r 2 0 8 0 0 0 6 k o 4 4 4 4 6 l t 2 1 4 1 i S 3 1 2 M 1 t 0 0 N a e 7 . 5 O o r . . 0 . . . . I C u 0 0 8 0 0 O 6 0 1 T t 4 4 4 2 A s 2 1 7 T a 3 S P 1 K O O d . R e . . . 0 . . . B r 2 0 8 0 0 0 0 A k o 4 4 4 4 6 5 E t 2 1 4 1 S l i S 3 1 1 2 - T M A w e 0 0

o r 7 .

5 S ) , 0 . . T C u O N A t 0 0 8 0 4 0 0 0 1 E s 4 4 2 5 U e a 2 1 7 L c P 3 r n 1 F e 2 r

-  t.

e y . 0 1 7 f . . 0 . . S e f 3 s a 2 0 0 0 4 1 B UO R e m 4 4 4 2 7 2 1 4 E L S E m o l b I 3 1 B 1 B A r a - AG T f t e 6 X R d g d . 7 O e e e . . 0 . . F v V r 2 0 0 0 0 0 i o 4 4 4 4 S r t 2 1 4 4 K. e S 3 1 1 2 D 1 1 ( t M ) A ) ) Y ) R z z A 3 A M M ) ) ) ) D m P / / S S S S / y g g R R R R g L K K H H H H K g _ A ( ( ( ( ( ( ( ( - T N __ E M n N i n 5 _ O - i . n i R z 0 _ I < w n - I e e o w -

   "l E                 y                  m u

r u G r o r e N 1 e t l t n G n 0 . l i P s o . i 6 v b v r a g . d e i e o P n e g o 5 R a t y s m t t e V e I i r c t U s n h V - , a u e e e d o w d l 9 V d'i o n s t o i m i m v r e e r e e r f y t a y t 0 1 r e T T a F a r o a n i P D e H y e u t e e d 1 - m e e Y t S L m i l e i r r r l s e m e a c T u u e i f a f f l u d r a s s t a o P o o E H i u f t o o f D u t r r p p A n nm n n n e G l u o x x s o o e o o o t y u S p E E p l i i r i n i n i u c s u a t t u t e t e t l r i l n l p d m c c t cd cd c o o r i a i o l i a a s a r a r a s t g o r o r o n r r a r a r a r b a A S T S C H A F F P F G F G F A l u g F F e V B H P S G L I R Y P T T T T Q F F F F F H

s 8 a c n 1 m i e . e u f i i ) v r b c t s e l d i s R i a e o n e d e _ o d h s a p n r _ s n t l l s oP _ e a c M e l r e r n C h a o yh o t o D t cF a t f ce m O h m f 0 l m a r t . P o . f g a r) F 1 s e r e u n t ns u r Pe y o o r f s . a a n o e a0 p h e t , n r e t c 0 o o s1 s d 2 a s e e3 - n t gt s s s a s n ad e u u81 h ri e l t i wt t d e3 f o e r o n b N s . e g r s op t a e O s s l l g I e y h n e e t l a T l a t ob r e e aP A w n sh h ; T f h g( l d t s 0 S o t n l n e a i oi a b f ) F K s p e r u e r w d y o 3 m1 0 O t / O s l d z) e a. R al e a g s F f P i mra m g) B e p ( A l ni( i t E e r i t r s S r o f r e e n r r r e

                                        -i    s a

6 ug T y uu ud S v a o ap pa so T i e N t y s I o( a l l 2 S u B U d h r f m O in e od o r o e e f i vnr 7 E E t i 0 AG C t 8 n e s i i ms t r t u n i J d X T R ( e t r a c s f e rh r rf t ai p ui h s S f e e a T s e sl t A R l a l n ri P a n m a a e t n a e a e r c e a s m a o f m r a A ei ul t t q T t e N al a V /t /t af E s r s a i s s d a I e r re t r p e t ) ) o 1 2 es 5 u A t ) c t t i ad A i r c s s p e o( nr t t 9 o op l 3 E c e n n s a E a1 s o ei e e u . l r ql F i e e e n i e o 2 F d s l b moi h d V E )d a r e i w na e e r uo r o t t t a d u 3 7. ed a a - f e L S f 6 m nf f y u B n d u o n B A o ei t o i 7 - t e i s d a o ai s d mc M i t o y R o s p ef f s eh E sh i o t P _ s u E eb e a s ni l h M s y y sm yi s o n o i n o t _ s l _. t o b a A t a ci a t a e H - r a n L e o e , _ m h f s c E r e rf r r t n e n M r o a a i s e N e e d f , y r O s t o a s nn o o e c s s l e R e i af I d h d i l i sh n e - yt aR N e e c ci l ( c af I o o o rd e ni I t d f d f r p a s - I eh I n o e I m I h t r I o h h i t o i t t d t d o ,dn f d t e r o o a M o h e h s ad h c N t r t r e . t o l e u e ed t s el l e M s M t n s u M ah r ar cjno u s o r o x o u ed e F q s a c o n o n F e F E F i s ) ) 3 4 N ( ( ( (

~_

TABLE B.7-3 USAGE FACTORS FOR VARIOUS CASEOUS PATHWAYS AT SEABROOK STATION l (from Reference A, Table E-5)* Maximum Racentor: l Age Leafy h Veretables Veretables gilk Haag Inhalation i (kg/yr) (kg/yr) (1/yr) (kg/yr) (m3 /yr) l Adult 520.00 64.00 310.00 110.00 8000.00 Teen 630.00 42.00 400.00 65.00 8000.00 1 child 520.00 26.00 330.00 41.00 3700.00 l l Infant 0.00 0.00 330.00 0.00 1400.00 l \ l The ' Rocks" and Science & Nature Center: l Age Leafy Group Veretables Verstables gilk gang Inhalation (kg/yr) (kg/yr) (1/yr) (kg/yr) (m3 /yr) i Adult 0.00 C.00 0.00 0.00 8000.00 l Teen 0.00 0.00 0.00 0.00 8000.00 Child 0.00 0.00 0.00 0.00 3700.00 1 Infant 0.00 0.00 0,00 0.00 1400.00

Regule ; v Guide 1.109 X-B.7 33 ODCM Rev. 18

7.3 RECEPTOR POINTS AND AVERAGE ATMOSPHERIC DISPERSION FACTORS FOR IMPORTANT EXPOSURE PATHWAYS The gaseous effluent dose equations (Method I) have been simplified by assuming an individual whose behavier and living habits inevitably lead to a higher dose than anyone else. The followin; exposure pathways to gaseous effluents listed in Regulatory Guide 1.109 (Reference A) have been considered:

1. Direct exposure to contaminated air;

2. Direct exposure to contaminated ground;

3. Inhalation of air;

4. Ingestion of vege' bles;

5. Ingestion of goat's milk; and

6. Ingestion of meat.

l Part B, Section 7.3.1 details the selection of important off-site and on-site l locations and receptors. Part B, Section 7.3.2 describus the atmospheric model used l to convert meteorological data into atmospheric dispersion factors. Part B, Section 7.3.3 presents the maximum atmospheric dispersion factors calculated at each of the off-site receptor locations. 7.3.1 Retector Locations The most limiting site boundary location in which individuals are, or likely to be located as a place of residence was assumed to be the receptor for all the gaseous pathways considered. This provides a conservative estimate of the dose to an individual from existing and potential gaseous pathways for the Method I analysis. This point is the west sector, 974 meters from the center of the reactor units for undepleted, depleted, and gamma X/Q calculations, and the northwest section, 914 meters for calculations with D/Q tse dispersion parameter. l The site boundary in the NNE through SE sectors is located over cidal marsh (e.g., over water), and consequently are not used as locations for determining maximum off-site receptors (Reference NUREG 0133). X-B.7-34 ODCM Rev. 18

7.3 RECEPTOR POINTS AND AVERAGE ATMOSPHERIC DISPERSION FACTORS FOR IMPORTANT EXPOSURE PATHWAYS 7.3.1 Recentor Locations (Continued) Two other locations (on-site) were analyzed for direct ground plane exposure and inhalation only. They are the " Rocks" (recreational site) and the Editcation Center shown on Figure 5.1-1 of the Technical Specifications. 7.3.2 Seabrook Station Atmosoherie Dinnersion Model The time average atmospheric dispersion factors for use in both Method I and Method II are computed for routine releases using the AEOLUS-2 Computer Code (Reference B). i AEOLUS-2 produces the following average atmospheric dispersion factors for l each location:

1. Undepleted X/Q dispersion factors for evalus. ting ground level concentrations of noble gases;

2. Depleted X/Q dispersion factors for evaluating ground level concentrations of iodines and particulates;

3. Gamma X/Q dispersion factors for evaluating gamma dose rates from a sector averaged finite noble gas cloud (multiple energy undepleted sourca); and

4. D/Q deposition factors for evaluating dry deposition of elemental ,

radiolsdines and other particulates. Gamma dose rate is calculated throughout this ODCM usint the finite cloud ; model presented in " Meteorology and Atomic Energy - 1968" (Reference E, Section 7-5.2.5). That model is implemented through the definition of an effective gamma atmospheric dispersion factor, [X/Q1] (Reference B, Section 6), and the replacement of X/Q in infinite cloud dose equations by the [X/Q5]. 7.3.3 Averara Armannharic Dinnersion Factors for Recentors The calculation of Method I and Method II atmospheric diffusion factors (undepleted CHI /Q, depleted CHI /Q, D/Q, and gamma CHI /Q values) utilize a methodology generally consistent with US NRC Regulatory Guide 1.111 (Revision 1) criteria and the methodology for calculating routine release diffusion factors as represented by the XOQDOQ computer code (NUREG/CR-2919). The primary vent stack is treated as a " mixed mode" release, as defined in Regulatory Guide 1.111. Effluents are considered to be part-time ground X-B.7 35 ODCM Rev. 18

P 7.3 NECEPTOR POINTS AND AVERAGE ATMOSPHERIC DISPERSION FACTORS FOR ZMPORTANT EXPOSURE PATHWAYS 7'. 3. 3 Aversea Ae=ascherie Dinnersion Factors for Recentors (Continued) l

level /part-time elevated releases depending on the ratio of the primary vent stack l effluent exit velocity relative to the speed of the prevailing wind. All other release points (e.g., Turbine Building and Chemistry lab hoods) are considered ]

ground-level releases.

          ~In addition, Regulatory Guide 1.111 discusses the concept that constant mean wind direction models like AEOLUS-2 do not describe spatial and temporal variations in airflow such as the recirculation of airflow which can occur during prolonged l

periods of atmospheric stagnation. For sites near large bodies of water like Seabrook, the onset and decay of sea breezes can also result in airflow reversals l and curved trajectories. Consequently, Regulatory Guide 1.111 states that . adjustments to constant mean wind direction model outputs may be necessary to account for such spatial and temporal variations in air flow trajectories. Recirculation correction factors have been applied to the diffusion factors. The recirculation correction factors used are compatible to-the " default open terrain" recirculation correction factors used by the XOQD0Q computer code. The relative deposition rates, D/Q values, were derived using the relative l deposition rate curves presented in Regulatory Guide 1.111 (Revision 1). These l curves provide estimates of deposition rates as a function of plume height, l stability class, and plume r~. vel distance. l Recentor Locations For ground-level releases, the downwind location of "The Rocks" (244m NE/ENE) at.d the Science & Nature Center (406m SW) were taken as the distance from the nearest point on the Unit 1 Administrative Building / Turbine Building consplex. For the site boundary, the minimum distances from the nearest point on the Administration Building / Turbine Building complex to the site boundary within a 45-degree sector centered ot the compass direction of interest as measured from UFSAR Figure 2.1-4A were used (with the exception that the NNE-NE-ENE-E-ESE SE site boundary sectors were not evaluated because of their over-water locations). For primary vent stack releases, the distances from the Unit 1 primary vent stack to "The Rocks" (244m NE) and the Science & Nature Center (488m SW) as measured from a recent site aerial photograph were used. For the site boundary, the minimum distances from the Unit 1 primary vent stack to the atte boundary within a 45-degree sector centered on the compass direction of interest as measured from UFSAR Figure 2.1-4A were used (with the exception that the l i X-B.7-36 ODCM Rev. 18 l j

, 7.3 RECEPTOR POINTS AND AVERAGE ATMOSPHERIC DISPERSION FACTORS FOR IMPORTANT EXPOSURE PATHWAYS 7 3.3 Aversea Armaanharie Dinnersion Factors for Recentors (Continued) NNE NE-ENE-E-ESE SE site boundary sectors were not evaluated because of their over water locations). Meteorolonical Data Bases For "The Rocks" and Science & Nature Center receptors, the diffusion factors reprasent six-year averages during the time period January 1980 through December 1983 and January 1987 through December 1988 (with the exception that, because of low data recovery, April 1979 and May 1979 were substituted for April 1980 and May 1980). For the site boundary receptors, both six-year average growing season (April through September) and year-round (January through December) diffusion factors were generated, with the higher of the two chosen to represent the site boundary. The meteorological diffusion factor used in the develcpment of the ODCM Method I dose models are summarized on Tables B.7-4 through B.7-6. i X.B.7-37 ODCM Rev. 18

TABLE B.7-4 SEABROOK STATION LONG-TERM AVERAGE DISPERSION FACTORS

PRIMARY VENT STACK Dose Rate to Individual Dose to Air Dose to critical Orgsn Total Skin Critical Camma Beta Thyroid Body Organ 7.5E-07 - --

7.5E-07 X/Q depleted 's e c' L "' .

                                    -      8.2E-07            -           -

8.2E-07 - X/Q undepleted "s e e' "3 .

             '1'                    -         -

1. 5 E - 0 8** - -

1.5E-08 D/Q _ 8.5E-07 8.5E-07 - 8.5E-07 - - X/Q1 'sec]

               ,   )

l l s l l 1

West site b0ondary, 974 meters from Containment Building i

 ** Northwest site boundary, 914 meters from Containment Building                                  1 X-B.7-38                             ODCM Rev. 18  l l

TABI.E B.7 5 SEABROOK STATION TANG-TERM AVERAGE DISPERSION FACTORS FOR SPECIAL (ON-SITE) RECEPTQRg PRIMARY VENT STACl( l l Dose to Critical Dose Rate to Individual Dose to Air Organ Total Skin Critical Gamma Beta Thyroid Body Organ Education Center: (SW - 480 meters)

                                -        -        1.5E-06    -         -

1.5E-06 X/Q depleted 1.6E-06 - X/Q undepleted "I'

                                -        -        2.7E-08     -        -          -

D/Q p

      's . c '              1.1E 06   1.1E-06        -

1.1E-06 - - gj , 7 The " Rocks": ; (ENE 244 meters)

                  's e c' 1.6E-05     -        -       1.6E 05 X/Q depleted
                  ,7 1.7E 05        -

X/Q undepleted s , l

     '1'                         -       -

1.1E-07 - - - D/Q I

       's e c'              5.0E-06   5.0E-06         -

5.0E-06 - - yf ,

,T.

l X-B.7-39 ODCM Rev. 18

I TABLE B 7-6 SEABROOK STATION LONG-TERM ATMOSPHERIC DIFFUSION AND DEPOSITION FACTORS GROUND-LEVEL PFf FASE PATHWAY R E C E P T O R(*) Diffusion Factor The Rocks Science & Nature Off-Site Center Undepleted CHI /Q, 1.6 x 10-4 2.3 x 10-5 1.0 x 10-5 3 sec/m (244m ENE) (406m SW) (823m W) Depleted CHI /Q, 1.5 x 10-' 2.1 x 10-s 9.4 x 10-s sec/m3 (244m ENE) (406m SW) (823m W) D/Q , m-2 5.1 x 10-7 1.0 x 10-7 5.1 x 10-e (244m ENE) (406m SW) (823m W) Gamma CHI /Q, sec/m3 2.6 x 10-5 5.3 x 10-s 3,4 x to-a (244m ENE) (406m SW) (823m W) (*) The highest site boundary diffusion and deposition factors occurred during the April through September growing season. Note that for the primary vent stack release pathway, none of the off-site receptor diffusion and deposition factors (located at 0.25-mile increments beyond the rite boundary) cxceeded the site boun(ary diffusion and deposition factors. X-B.7-40 ODCM Rsv. 18

8.0 BASES FOR LIQUID AND CASEOUS MONITOR SETPOINTS 8.1 BASIS FOR THE LIQUID WASTE TEST TANK MONITOR SETPOINT The liquid waste test tank monitor setpoint musc ensure that the limits of l Part A Control C.5.1 are not exceeded in combination with any other site discharge pathways. The liquid waste test tank monitor is placed upstream of the major source of dilution flow. The derivation of Equation 5-1 bagins with the general equation for the response of a radiation monitor: R-[ C yi Su (8-1) (cps) = ( ml ) ( cps-el) pCi where: R - Response of the monitor to radioactivity (cps). Sti - Detector counting efficiency for radionuclide "i" (eps/(pCi/ml)). C,1 - Letivity concentration of each gamma e. sitting radionuclide "i" in the mixture that the monitor has a response efficiency sufficient to detect (pci/ml). I The detector calibration procedure for the liquid waste test tank monitor at Seabrook Station establish-s counting efficiency by use of a known calibration source standard and a line -ity response check. Therefore, in Equation 8-1 one may substitute S t for S ig, where St is the detector counting efficiency determined from the calibration procedure. Therefore, Equation 8-1 becomes: R- Si [ Cy (8-2) (eps)- ( CE"I ) ( #ml' ) pCi X-B.8-1 ODCM Rev. 18

r 1 8.1 BASIS FOR THE LIQUID WASTE TEST TANK MONITOR SETPOINT (Continusd) l The MFC for a given radionuclide must not be exceeded at the point of discharge to the environment. When a mixture of radionuclides is present, 10CFR20 specifies that the concentration (excluding dissolved and entrained noble gases) at the point of discharge shall be limited as follows: (8-3) [ Cd' s 1 MPC, where: Cat - Activity concentration of radionuclide "i" determined to be present in the mixture at the point of discharge to the environment (pci/ml). MPCg - The maximum permissible concentration (pCi/ml) for radionuclide "i" ; from 10CFR20, Appendix B. Table II, Column 2 for all radionuclides j except noble gases. The limit for the sum of all noble gases in the waste discharge is 2E-04 pCi/ml (See ODCM Appendix B for listing). The activity concentration of radionuclide "i" at the point of discharge is related to the setivity concentration of each radionuclide at the monitor as follows: F C,= g %., + Q,) (8 4) Fg (gCi) , gpm) #Ci ) ml gpm ml l and with equivalence of Ct - (Cyi + C$i), Equation 8-4 can be written as l Cd, = -F" Ci ! Fa l where: i F. - Flow rato past monitor (gpm) Fd - Flow rate out of discharge tunnel (gpm) l t C$$

                -     Activity concentration of non gamma emitting radionuclide "i" in        ;

the mixture at the monitor for which the monitor response is -! inefficient to detect (pCi/ml). l Ci - The activity concentration of each radionuclide "i" in the waste stream. This includes both gamma and non gamma emitters, such as tritium. X F 8-2 ODCM Rev. 18

8.1 BASIS FOR THE LIQUID WASTE TEST TANK MONITOR SETPOINT (Continuad) Substituting the right half of Equation 8-4 for Cd i in Equation 8-3, and solving for F 4/F, yields the dilution factor needed to. complete Equation 8-3: l F C' DF,in s

                   ' F, 2 i[MPC,                                                                        (8-5)   )

gpm) pCi-el) gpm ml-pCi where: MPC i - The maximum permissible concentration (pCi/ml) for radionuclide "i" from 10CFR20, Appendix B, Table II, Column 2 for radionuclides, except dissolved and entrained noble gases. For noble gases, a value of 2E-04 pCi/ml is used for the limit of the sum of noble gases in the waste stream. If F 4/F, is less than DF,in, then the tank may not be discharged until either Fd or F, or both are adjusted such that: (8-5) DFi, $ The maximum allowable discharge flow rate past the monitor can be found by setting Fm to F.,, and its equivalents, i.e: l r ** . '< DF,in Usually F4 /F, is greater than DF,in (i.e., there is more dilution than necessary to comply with Equation 8-3), but must be satisfied since the monitor can only detect the gamma emitt.ing portion of the waste stream. It is assumed that changes in the expected gamma concentration seen by the monitor from that determined in laboratory analysis are also reflected proportionally in the concentration of non gamma emitters. For tritium, this is conservative since changes in tritium are not affected by those mechanisms, such au crud burst, which could increase particulate gamma emitters. The response of the liquid waste test tank monitor at the setpoint is therefore: ft x d (0~0) R,,gpi,, - x St{C i cps-al pCi (eps} () () g pi ,7 { or with F , substituted into Equation 8-6 for the maximum allowable discharge flow

            #d rate                      ,   the setpoint equation can be stated also as:                                     l t

DYetn , a.. .,. .. - fi x ' '" x s,E cy, i 7 X-B 8-3 ODCM Rev. 18

l 1 8.1 BASIS FOR THE LIQUID WASTE TEST TANK MONITOR SETPOINT (Continu2d) where ft is equal to the fraction of the total concentration of MPC at the discharge l point to the environment to be associated with the test tank effluent pathway, such l that the sum of the fractions of the four liquid discharge pathways is equal to or less than one (ft+ f2+ f3 + f s 1) . 1 The monitoring system is designed to incorporate the detector efficiency, S ,t into its software. This results in an automatic readout in pCi/mi or pCi/cc for the monitor response. Since the conversion for changing cps to pCi/mi is inherently done by the system software, the monitor response setpoint can be calculated in terms of the total waste test tank activity concentration in pCi/ml determined by > the laboratory analysis. Therefore, the setpoint calculation for the liquid waste test tank is: F R ,,3g - ft x x { Cyg (pci) () () (pC1) al (5-1) ml

I l l l X-B.8-4 ODCM Rev. 18

v 8.2 BASIS FOR THE PIANT VENT WIDE RANGE GAS MONITOR SETPOINTS The.setpoints of the plant vent wide range gas monitors must ensure that Part A Control C.7.1.1.a is not exceeded. Part B, Sections 3.4 and 3.5 show that Equations 3-3 and 3 4 are acceptable methods for determining compliance with that l Part A Control. Which equation (i.e., dose to total body or skin) is more limiting depends on the noble gas mixture. Therefore, each equation must be considered separately. The derivations of Equations 5-5 and 5-6 begin with the general equation for the response R of a radiation monitor: 1 R- S,g C,5 (8-7) 3 (cpn) - ( cpm-cm ) ( pCi) pCL cm 3 where: R - Response of the instrument (cpm) S,i - Detector counting efficiency for noble gas "i" (cpm /(pci/cm 3)) C ,1 - Activity concentration of noble gas "i" in the mixture at the noble gas activity monitor (pCi/cm 3) C,i, the activity concentration of noble gas "i" at the noble gas activity monitor, may be expressed in terms of Qi by dividing by F, the appropriate flow rate. In the case of the plant vent noble gas activity monitors the appropriate flow rate is the plant vent flow rate. C ,1 - gi (8-8) (pCi) ,(yCi) (sec) cm 3 sec cm 3 where: 1 Qt

                         -      The release rate of noble gas "i" in the mixture, for each noble gas listed in Table B.1 10.

F - Appropriate flow rate (cm 3/sec) Substituting the right half of Equation 8-8 into Equation 8-7 for C,t yields: R- S,g Qi (8-9) 3 (epm) - ( cpm-cm ) ( pCi) ( sec) pci sec em 3 X-B.8-5 ODCM Rev. 18

8.2 BASIS FOR THE PLANT VDIT WIDE RANGE CAS MONITOR SETPCINTS (Continued) As in the case before, for the liquid waste test tank monitor, the plant vent wide range gas monitor establishes the detector counting efficiency by use of a ) calibration source. Therefore, S, can b'e substituted for 5,i in Equation 8-9, where j S, is the detector counting efficiency determined from the calibration procedure. J Therefore, Equation 8-9 becomes: R- (8-10) S, f { Qi 3 (cpm) - ( cpm-cm ) 4 m) ( yCi) pCi cm sec ; I The total body dose rate due to noble gases is determined with Equation 3-3: i l i Dtb - 0.85

EL(R)
Qg DFB i (3-3) !

l l 3 ( mrem) , ( pCi-sec) () (9C1) (mrem-m ) yr pCi-m a see pci-yr where: b s3 - total body dose rate (mres/yr) 0.85 - (1.0E+06) x (8.5E-07) (pCi-sec/pci-m 3 ) 1E+06 - number of pCi per pCi (pCi/pci) 8.5E-07 - [X/Q)1, maximum off-site average gamma atmospheric *ispersion factor (sec/m 3 ) for primary vent stack releases EL(R) - Release point correction factor - 1.0 for primary vent stack Qt - As defined above. DFB t - total body dose factor (see Table B.1-10) (arem-m /pCi-yr) 3 1 X-B.8-6 ODCM Rev. 18

8.2 BASIS FOR THE PIANT VENT WIDE RANGE GAS MONITOR SETPOINTS (Continusd) A composite total body gamma dose factor, DF3,, may be defined such that: Di DFB i (8-11) DFBc 8 ( Di 8 mrem-m (pC1) ,, (pCi) (mrem-m ) pGi-yr see sec pCl-yr Solving Equation 8-11 for DFB, yields: E1 4 iDFB 3 DFB,- (5-7) [D i i l- Part A Control C.7.1.1.a limits the dose rate to the total body from noble gases at any location at or beyond the site boundary to 500 l mrem /yr. By setting b t3 equal to 500 arem/yr and substituting DFB, for DFBi in Equation 3-3, one may solve for {Di at the limiting whole body noble gas dose rate: Di - 588 1 (8-12) 8 4 pCi) , ( arem-pci-m ) ( pCi-yr3 ) sec yr-pci-sec arem-m Substituting this result for Qi in Equation 8-10 yields Rtd. the response of the monitor at the limiting noble gas total body dose rate: 1 1 Rtb - 588 S, 7 g* 8 8 (8-13) (cpx) - ( arem-pCi-m ) ( cpa-cm ) ( ) ( pCi-yr)3 yr-pGi-sec pGi cm arem-m X-B.8-7 ODCM Rev. 18

8 .' 2 BASIS FOR THE PLANT VENT WIDE RANGE GAS MONITOR SETPOINTS (Centinusd) The skin dose rate due to noble gases-is determined with Equation 3-4: 0,gi, - EL(R) * (3-4) Qi DF'i (mrem) ,( ) {pCi) { mrem-sec) yr sec pCL-yr where: EL(R) - 1.0 for primary vent stack release (dimensionless) b,gio - Skin dose rate (area /yr) Qi

                     -       As defined above.

Combined skin dose factor (see Table B.1 10) (arem-sec/pci-yr) ! DF'i - I 1 A composite combined skin dose factor, DF',, may be defined such that: (8-14) DF', {Q i i {i 41 DF'i (mrem-sec) (pCi) _ ( pCi) ( mrem-sec) pGi-yr see sec pGi-yr l l Solving Equation 8-14 for DF', yields: l [D DF't t DF', - (5-8) hD i X-B.8-8 ODCM Rev. 18

8.2 BASES FOR THE PZANT VENT WIDE RANCE GAS MONITOR SETPOINTS (Continued) Part A Control C.7.1.1.a limits the dose rate to the skin from noble gases at

y location at or beyond the site boundar-f to 3,000 arem/yr.

By setting D,gi, equal to 3,000 arem/yr and substituting DF', for DF'i in Equation 3-4 one man solve for {Q at the limiting skin noble gas dose i rate: { Q - 3,000 i i * (8-15) (pCi) (area) (arem-yrpCi sec yr sec ) W s Substituting this result for {Q in Equation 8-10 yields R,gio, the -~ i ! ' response of the monitor at the limiting noble gas skin dose rate: R,yg, - 3,000 5, 3 f pCi-yr (cpa) ("#**) yr ( CPa-cm pC1

                                                                                                              )  (sec) cm 3   ( arem-sec)

(8-16) As with the liquid monitoring system, the gaseous monitoring system is also designed to incorporate the detector efficiency, S,, into its software. The monitor also converts the response output to a release rate ( Ci/sec) by using a real time stack flow rate measurement input. Therefore, multiplying by the stack flow rate measurement (F), the Equations 8-13 and 8 16 become: P ., - 588 (5-5) 3 ( pCi) , ( arem-pCi-m ) ( pCi-yr3 ) see yr-pGi-see urem-m R,mi - 3000 (5-6) pCi ( sac

                                                                           #E) . ; area) yr { arem sec yr )

t X-B.8-9 ODCM Rev. 18

f-i 8.3 BASIS FOR PCCW HEAD TANK RATE-OF-CHANGE ALnRM SETPOINT I l l The PCCW head tank rate-of-change alarm will work in conjunction with the PCCW radiation monitor to alert the operacor in the Main Control Room of a leak to the Service Water System from the PCCW System. For the rate-of-change alarm, a setpoint based on detection of an activity level of 10**pci/cc in the discharge of the Service Water System has been selected. This activity level was chosen because it is the minimum detectable level of a service water monitor if such a monitor were installed. The use of rate-of-change alarm with information obtained from the l liquid sampling and analysis commitments described in Table A.6.1-1 of Part A ensure that potential releases from the Service Water System are known. Sampling and analysis requirements for the Service Water System extend over various operating ranges with increased sampling and analysis at times when leakage from the PCCW to the service water is occurring and/or the activity level in the PCCW is high. l l l X-B.8-10 ODCM Rev. 18

8.6 BASIS FOR WASTE CAS FR0 CESSING SVETEM MONfTORS (RM-6504 AND RM-6503) The maximum allowable setpoint for the waste gas system monitors (response in uCi/cm8 ) can be determined by equating the limiting off-site noble gas dose rate from the plant vent to the total body or skin dose rate limits of Pcrt A Control C.7.1.1.a assuming that all the activity detected by the vent wide-ranbe gas l monitors is due to waste gas system discharges. l I l By evaluating the noble gas radionuclide with the most limiting dose factor as given on Table B.1-10, a conservative activity release rate from the plant vent for both whole body and skin dose rate conditions can be calculated. From Table B.1-10, Kr-89 is seen to be the most restrictive noble gas if it were present in the l effluent discharge. Applying plant vent setpoint equation 5-5 for the whole body, , t and equation 5-6 for the skin, the maximum allowable stack release rate can be l I calculated as follows: l l Rt3 - 588 1/DFB, (5 5) I where: 1 Rtb - Pl ant vent maximum release rate (uCi/sec) based on the whole body does rate limit of 500 mres/yr i 8 DFB, - 1.66E-02 (arem m /pci-yr), whole body dose factor for Kr-89 588 - conversion factor (mrem-uci-m 8 /yr-pCi-sec) Therefore: R tb - 588 1/1.66E-02

              -    35,421 uCi/see maximum release rate at plant vent i

Next, the skin dose rate limit is evaluated from equation 5-6 in a similar fashion l as follows: R,gio - 3000 1/DF', (5-6) where: R,gt, - plant vent maximum release rate (uCi/sec) based on skin dose rate limit of 3000 arem/yr. I DF', - 2.45E-02 mrem-sec/uci-yr skin dose factor for Kr-89 l 3000 - Site boundary skin dose rate limit (arem/yr) { X B.8 11 ODCM Rev. 18

r

8.4 BASIS FOR WASTE GAS PROCESSING SYSTEM MONITORS (RM-6504 AND RM-6503) (Continued) i I therefore: Raun

                     -   3000(ares /yr) 1/2.45E-02(arem-sec/uci-yr)
                     -   122,449 uCi/see from the plant vent i

Comparing the release rate limit for the whole body to that for the skin l (i.e., 35,421 uCi/sec vs 122,449 uCi/sec, respectively) it is determined that the release rate for the whole body is limiting. Next, to get the maximum plant vent release rate from the waste gas system discharge, equate the plant vent maximus release rate limit for the whole body equal to the waste gas system activity concentration times its flow rate to the plant vent, i.e.: R t3 - 35,421(uCi/sec) - R,,,(uCi/cm8 ) F.,,(cm8 /sec) or solving for R,,,: 8 R,,,(uci/cm ) - 35,421(uCi/sec) / F,,,(cm3/sec) where: R ,, - maximurn concentration (setpoint limit) at the waste gas . system monitors F,,, - waste gas design flow of 566.4 cm8 /sec (1.2 cfa) therefore: R,,,(uCi/cm8 ) - 35,421(uC1/sec) / 566.4(cm 8/sec) 62.5 uCi/cm8 This represents the maximum waste gas discharge concentration which would equal the site boundary whole body dose rate limit for plant vent releases. Administrative controls may set alert alarm and high alarm (waste gas isolation) setpoints on the waste gas monitors 'as some multiple of expected activity concentration, such as 1.5 and 2 times, respectively, as long as the maximum setpoint does not exceed 62.5 uCi/cm8 This provides operational controls to be l exercised before any waste gas discharges could equate to the Part A Control C.7.1.1.a. X-B.8-12 ODCM Rev. 18 ] j

8.4 3 ASIS FOR WASTE GAS PROCESSfNG SYSTEM MONITORS (RM-6504 AND RM-6503) (Continued) l l The primary process monitor noted in Part A Control C.5.2 is RM-6504, which is downstream of the waste gas discharge compressor at the end of the process system. Monitor RM-6503 is on the inlet side of the compressor downstream of the charcoal delay beds, and iv considered as an alternate monitor if RM-6504 is inoperable. For the purpose of sotting the maximum discharge setpoint, RM-6503 is treated the same I as RM-6504, which assumes no additional source reduction before discharge to the l plant vent. 4 l l l t X-B.8-13 ODCM Rev. 18

i 8.5 BAS 1S FOR THE MAZN CONDENSER AIR EVACUAT80N MONITOR SETPOINT (RM-6505) The maximum allowable setpoint for the main condenser air evacuation monitor must be evaluated for two modes of operation. For normal operations the monitor is ! responding to a low flow rate that is released through the plant vent stack. During ) start up (hogging mode), the monitor response must be related to a high flow rate that is being released from the turbine building which is considered a ground level release. In both instances, the setpoint can be determined by equating the limiting off-site noble gas dose rate from the release point to the total body or skin dose rates of Part A Control C.7.1.1.a. In a manner similar to that for the waste gas monitoring system in Part B, Section 8.4 the most restrictive radionuclide in Table B.1-10, Kr-89, can be used to calculate a conservative activity release rate condition for both total body and skin dose rates. More realistic or actual l radionuclide distributions in condenser air can be used to calculate the maximum allowable alarm setpoint. In addition to monitoring the main condenser air, the air evacuation monitor response is also used as an indicator for Turbine Gland Seal Condenser exhaust. Since this is a potential release pathway during both the normal and the hogging modes of operation, the impact is considered in the setpoint calculations. I l 8.5.1 Examnle for the Air Evacuation Monitor Setnoint During Normal Ooerations During normal power operation the maximum allowable setpoint for the air evacuation monitor is determined by applying plant vent setpoint equation 8-13 for the total body, and equation 8-16 for the skin. Therefore, the maximum allowable stack release rate can be calculated as follows: Rtb - (588) (S,) (1/F) (1/DFB,) (8-13) (cpm) - (arem pCi-m8 /yr-pCi-sec) (cpm cm3/pCi) (sec/cm )(pCi-yr/ 8 mrem-m 3) l where: l R tb - count rate (cpm) for the plant vent maximum release rate based on the total body dose rate limit of 500 mrea/yr 588 - conversion factor (arem pCi-m 8 /yr-pCi-sec) S, - the detector response efficiency (cpm-cm /pci) 8 as determined from monitor calibration. For the air evacuation monitor, a typical l value is 6.0E+05 cpm-em /pci. 8 l l F = release flow rate. During normal operations, a typical flow value

is 4.72E+03 cc/sec (10 cfm) for the air evacuation pathway.

l l DFB, - the composite total body dose factor, For Kr-89 alone, the value is i l 1.66E-02 (mrem m 8

                                      /pCi-yr). For different gat mixes, the composite can     l be found from:                                                               l l

DFB, - DFB i / Qi (5-7) {Qi l X-B.8-14 ODCM Rev. 18

8.5 BASfS FOR THE MAIN CONDENSER AIR EVACUATION MONITOR SETPOINT (RM 6505) l 8.5.1 Examole for the Air Evacuation Monitor Setooint During Normal Operations I l (Continued) Therefore, Ra - 588 6.0E+05 (1/4.72E+03) (1/1.66E-02)

                 -   4.50E+06 cpm detector count rate for a maximum release rate at the plant vent based on the total body dose rate.

Next, the off-site skin dose rate limit is eve'uated frem equation 8-16 in a similar fashion as follows: 1 R,mi - 3000 S (1/F) (1/DF',) (8-16) 1 (epm) - (arem/yr) (cpm-cm3 /pci) (sec/cm 3) (pCi-yr/arem-sec) j I where:

R,mg. - count rate (cpm) for a plant vent maximum release rate based on the skin dose rate limit of 300 mrem /yr DF', - the elevated release skin dose factor for Kr-89 of 2.45E-02 (mrem-sec/pci-yr). Therefore, R,mi - 3000 6.0E+05 (1/4.72E+03) (1/2.45E 02)

                  -  1.56E+07 cpm detector count rate for a maximum release rate at the plant vent based on the skin dose rate.

Comparing the release rate limit for the total body to that of the skin (i.e., 4.50E+06 cpm versus 1.56E+07 cpa, respectively) it is determined that the release rate for the total body is limiting in this case. l l Since during normal operations the Turbine Cland Seal Condenser exhaust has the potential to be a minor additional contribution to the plant vent release, the effective contributior from the main condenser exhaust must be limited to some frar. tion of the calculated value. The contribution from the Turbine Cland Seal Condenser exhaust is expected to be minor because this system handles only 670 l lbs/ hour of steam which is a very small fraction of the 1.5E+07 lbs/ hour of l secondary side steam that the main condenser handles. Therefore, the maximum alarm

!.s set at 3.2E+06 cpm, which is 70% of the calculated value, to ensure that the contribution of the two does not exceed the dose rate limit of Part A Control C.7 1.1.a.      During normal operations, this would represent the maximum allowable          J count rate on the air evacuation monitor that would equate to the site boundary total body dose rate limit or less.

X-B.8-15 ODCM Rev. 18

8.5 BASIS FOR THE MAIN CONDENSER AIR EUACUATION MONITOR SETPOINT (RM-6505) 8.5.2 Examole for the Air Evacuation Monitor Setooint Durine Start Uo (Horcine Mode) l During start up (hogging mode), the determination of the air evacuation setpoint i must take into account a larger air flow rate that is also released as a ground level effluent. The flow rate must also include the contribution from the Turbine l Gland Seal Condenser exhaust, which is a potential release pathway which the air evacuation monitor response must also take into account. For ground releases, the general equation 8-10 is used to represent the monitor count rate. R - (S ) (1/F) Di (8-10) (cpm) - (cpm-cm3

                                  /pci) (sec/cm3 ) (pCi/sec)                                !

where: 1 R - detector count rate (epm) S, - the detector efficiency (cpm-cm 3 /pci) 1 F - release flow rate (cm 8/sec) Qi - the release rate of noble gas "i" in the mixture, for each noble gas listed in Table B.1-10. For a ground release, the off-site total body dose rate is based on: D tbts) "34 (3-3b) ((41

DFB )

i A composite total body dose factor, DFB, can be defined such that: DFB, {Q s i {(Qi DFB ) t i (8-11) 1 l l l l l X-B.8-16 ODCM Rev. 18 l

8.5 BASSS FOR THE MA2N CONDENSER AIR EVACUATION MONITOR SETPOINT (RM-6505) 8.5.2 Examnie for the Air Evacuation Monitor Setooint Durina Start Uo (Horring Mode) (Continued) By substit" ting 8-11 into 3-3b and rearranging to solve for Qi the following equation is obtained: Qi - (D ctb<,3 / 3.4) (1/DFB,) By inserting a limiting value of 500 mrea/yr as 6tb(s) this simplifies to: Qi - 147 (1/DFB,) Insertion of this equation into equation 8-10 yields: Rth(s) - 147 Sc,3 (1/F) (1/DFB,) (cpm) - (arem-pCi-m 3

                                     /yr-pCi-sec) (cpm-cm3 / Ci) (sec/cm 3) (pCi-yr/ mrem m3 )

where: Rtb(s) - count rate (cpa) for the maximum ground release rate based on the total body dose rate limit of 500 are/yr. 147 - conversion factor (arem-nci-m 3/yr-pci-sec) S.

                 -   the detector response efficiency for the air evacuation monitor (a typical value of 6.0E+05 cpm-cm3 /pci is applied in this example).

I i X-B.8-17 ODCM Rev. 18

[ 8.5 BASIS FOR THE MAIN CONDENSER AIR EVACUATION MONITOR SE.fPOINT (RM-6505) 8.5.2 Examnle for the Air Evacuation Monitor Setooint Durine Start Uo (Horrine Mode) (Continued) F - release flow rate. During the hogging mode of operation, a value 3 of 5 57E+06 cm /sec (1.18E+04 cfa) is assumed. This value l represents the sum of the 1.0E< 'e4 cfm from the main condenser discharge and the 1.8E+03 cfm exhaust rate from the Turbine Gland Seal Condenser. Both discharges are to the Turbine Building roof. DFB, - the total body dose factor from Table B.1-10. For Kr-89, this factor ! is 1.66E-02 (arem m 8/pCi yr). Therefore: Rac,3 - 147 6.0E+05 (1/5.57E+06) (1/1.66E-02) 9.54E+02 cpm detector count rate for a maximum ground release rate based on the total body dose rate. Next, the off-site skin dose rate limit for a ground release is evaluated from equation 3-4b in a similar fashion as follows: (3-4b) ((Dt DF c,3) D.ktn(s) i A composite skin dose factor, DF',g,3 can be defined such that: DF',c,3 [Qi - (Q, DFic ,3) (8-17) By substituting 8-17 into 3-4b and rearranring to solve for Qi the following equation is obtained: E4t A

0. kin <s) (1/DF',c,3 )

By inserting a limiting value of 3000 mres/yr as 6,gtc,3 this simplifies to: X-B.8-18 ODCM Rev. 18

I l -- 8,5 BASIS FOR THE MAIN CONDENSER AIR EVACUATION MONITOR SETP0ZNT (RM-6505) 8.5.2 Example for the Air Evacuation Monitor Setooint During Start Un (Horrine Mode) I (Continued) Qi 3000 (1/DF',c,3) l Insertion of this equation into equation 8-10 yields: R* kin - 3000 S, (1/F) (1/DF',g,3) (cpm). - (arem/yr) (cpa-cm3 /pci) (sec/cm 8) (pci-yr/arem-sec) t

where: R,ging,3 - Count rate (cpm) for the maximum ground release rate based on the skin dose rate limit of 3000 mrem /yr. D F ' , g ,3 - the ground release skin dose factor from Table B.1-10. For

                                -   Kr-89, this factor is 1.67E-01 (arem-sec/p/Ci-~yr).

Therefore: R,ging,3 - 3000 6.0E+05 (1/5.57E+06) (1/1.67E-01) 1.94E+03 cpm detector count rate for a maximum ground release rate based on the skin dose rate. Comparing the release rate limit for the total body to that of the skin (i.e., 9.54E+02 cpm versus 1.94E+03 cpa, respectively) it is determined that the release rate for the total body is limiting in this case. During start up (hogging mode), this represents as a ground level release the maximum allowable count rate on the air evacuation monitor that would equate to the site boundary total body dose rate limit. Since during startup, the plant vent still constitutes a primary release pathway., the effective contribution from the hogging exhaust must be limited to some fraction of the calculated value to ensure that the combination of all gaseous 1 l releases from the station do not exceed the dose rate limits of Part A Control C.7.1.1.a. In this example, the maximum alarm point is set at 15% of the calculated value, or 1.4E+02 cpm. i 1 X-B.8-19 ODCM Rev. 18

REFERENCES i A. Regulatory Guide 1.109, " Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compli'.nce with 10CFR50, Appendix I", U.S. Nuclear Regulatory Commission, Revision 1, October 1977. B. Hamawi, J. N. , "AEOLUS-2 -- A Computer Code for the Determination of Continuous and Intermittent-Release Atmospheric Dispersion and Deposition of Nuclear Power Plant Effluents in Open-Terrain Sites, Coastal Sites, and Deep-River Valleys for Assessment of Ensuing Doses and Finite-Cloud Gamma Radiation Exposures," Entech Engineering, Inc., March 1988. C. Regulatory Cuide 1.111 " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases From Light-Water Cooled Reactors". U.S. Nuclear Regulatory Commission, March 1976. D. National Bureau of Standards, " Maximum Permissible Body Burdens and Maximum Permissible Concentrations of Radionuclides in Air and in Water for Occupational Exposure", Handbook 69, June 5, 1959. E. Slade, D. H., " Meteorology and Atomic Energy - 1968", USAEC, July 1968. l F. Seabrook Station Technical Specifications. I l-X-R-1 ODCM Rev. 18

10 CFR 50.59 EVALUATION l

1. IDENTIFICATION NUMBER: O QC.b4 SHEET I OF )\

2. TITLE: Offsite Dose Calculation Manual Rev.18

3. INITIATOR: L. R. Tardif DATE: 12/2/97

4. DETERMINATION OF SAFETY EVALUATION APPLICABILITY I

Does the proposed change: A. Make changes in the facility as described in the UFSAR? O YES X NO l Make changes in procedures as described in the UFSAR? X YES O NO B. C. Involve tests or experiments not described in the UFSAR? O YES X NO BASIS (sepporting information 's required for each answer; attach additional pages as necessary): See Attachment O Check this block if the answers to Questions 4A,4B and 4C are NO md a Safety Evaluation is being perfonned for conservatism. Safety Evaluations performed for consecvatism do not require SORC review.

5. OPERATING LICENSE Does the proposed change require a change to the existing Operating License (including the Technical Specifications) or are additional Operating License requirements needed? X YES O NO BASIS (supporting information is required for each answer; attach additional pages as necessary):

See Attachment

6. REVIEWS INDF/ENDENT REVA AffdD3lgkIN WjiW REVIEWER ([z?,14 6 f

hf REVIEWER b /M [(7 - ~ DATE ! DATE /2- o2-77 - 7,j,, i NARC FORM 3-3A Rev. 50 Page 1 of 2 y- l ODC.H StJ I V

A 10 CFR 50.59 EVALUATION (Continued) IDENTIFICATION NUMBER

7. SAFETY EVALUATION:

These questions shall be answered if any question in block 4 or 5 is answered YES. A. Will the probability of an accident previously evaluated in the UFSAR be increased? O YES XNO B. Will the consequences of an accident previously evaluated in the UFSAR be increased? O YES XNO C. Will the probability of a malfunction of equipment important to safety be increased? O YES XNO D. Will the consequences of a malfunction of equipment imponant to safety be increased? O YES XNO E. Will the possibility of an accident of a different type than any previously evaluated in O YES XNO the UFSAR be created? F. Will the possibility of a malfunction of a different type than any previously evaluated O YES X NO in the UFSAR be created? G. Will the margin of safety as defined in the basis for any technical speification be O YES XNO reduced? BRIEF

SUMMARY

AND BASIS (supponing information is required for each answer, attach additional pages as necessary): See attached. ,

Based on the above does an unreviewed safety question exist? O YES XNO e is a UFSAR update required? X YES O NO Explain. (Protected: Ref. ACR 96-0152)Section 1.8 of the UFSAR references TS 3/4.11.1,3/4.1.2,3/4.11.3,3/4.3.4.8 and 3/4.3.4.9 which will be changed to ODCM sections. UFSAR change required when LAR 97-08 implemented.

8. SORC REVIEW NSARC REVIEW MEETING # n Q)[-Ol(p MEETING #

DATE M 42./ 6[9 h DATE

          /m/                                                         NSARC CHAIRMAN CONCURRENCE
          ' STATION DIRIf[OR APPROVAL M                                                                 NARC FORM 3-3A Rev. 50 Page 2 0f 2 y-1                                  OD C. M Rw &

3

l l Seabrook Station 10 CFR 50.59 Evaluation Relocation of Radioactive Emuent and Environmental Monitoring Technical Specifications to the Offsite Dose Calculation Manual (ODCM) Per NRC Generic Letter 89-01 i

Introduction:

The NRC staff examined (Generic letter 89-01) the contents of Radiological Emuent Technical Specifications (RETS) in relatic,n to the Commission's Interim Policy Statement on Technical Specification Improvements. The staff determined that programmatic controls can be implemented in the Administrative Controls section of the Technical Specifications (TS) to satisfy existing regulatory requirements for RETS. At the same time, the procedural details of the current TS on radioactive effluents and radiological environmental monitoring can be relocated to the Offsite Dose Calculation Manual (ODCM). These actions simplify the RETS, I meet the regulatory requirements for radioactive effluents and radiological environmental monitoring, and are provided as a line-item improvement of the TS, consistent with the goals l of the NRC Policy Statement. Future changes to the procedural details in the ODCM are l performed under Technical Specification Administrative Controls for changes to the ODCM. Deerietion: l- , The proposed line-item improvements to the TS relative to the RETS have been made in I accordance with the guidance of NUREG-1301, "Offsite Dose Calculation Manual Guidance: Standard Radiological Effluent Controls for Pressurized Water Reactors - Generic Letter 89-l 01, Supplement 1". The amendment to the TS also includes the addition of programmatic l controls for RETS consistent with regulatory requirements, and allows the relocation of procedural details from the TS consisting of current limiting conditions for operations, their l applicability, remedial actions, surveillance requirements and Bases sections for these TS, to other appropriate licensee-controlled document (s), namely the ODCM. Ilo substar.tive changes in the content of the procedural details is being made. He addition of the programmatic controls to the Administrative Control Sections of the TS reflect the current effluent control and environmental monitoring requirements that exist now in the TS. i Relocation of the procedural details to the appropriate licensee-controlled documents is justified since they no longer meet the criteria of 10 CFR 50.36 for inclusion in TS. file: sb8901. doc y '3 on c M Ro iF

                                                                                                     &1 :

i ' The NRC revised 10 CFR 50.36 on July 19,1995 to meet the intent of a previously issued policy statement, " Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors," (58FR39312), which provides a specific set of four (4) objective criteria to determine which of the design conditions and associated surveillance's should be located in the TSs as limiting conditions for operations. The NRC recognized that implementation of these additional criteria may cause some requirements presently in the TSs to no longer merit inclusion in TSs. The proposed RETSs to h relocated are candidates for no longer meriting inclusion in TSs since they do not meet the four objective criteria specified in 10 CFR 50.36. The design conditions contained within the proposed RETSs to be relocated are not: (1) installed instrumentation that is used to detect, and indicate in the control room, a significant aonormal degradation of the reactor coolant pressure boundary; (2) a process variable, design feature, or operating restriction that is an initial condition of a design basis < accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; (3) a structure, system or component (SSC) that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; and, (4) a SSC which operatmg experience or probabilistic risk assessment has shown to be significant to public health and safety. Future changes to the relocated RETSs in the ODCM will be govemed by the programmatic and administrative controls specified in the Administrative Controls section of the TS. Therefore, a comparable level of administrative control will continue to be applied to those design conditions and associated surveillance's being relocated to the ODCM. One other significant change is made to the ODCM, Part A. The current wording requires prior NRC review and approval of all proposed changes made to Part A. Part A currently consists of the Radioactive Liquid Waste Sampling and Analysis Program, the Radioactive Gaseous Waste Sampling and Analysis Prey-u, and the Radiological Environmental Monitoring Program. These programs are to assure that the amount of radioactive effluents released to the environment and resultant exposures to members of the public are mamtained as low as is reasonably achievable and within the limits of the respective Technical Specifications. The requirement to obtain NRC review and approval prior to implementation of the proposed eh==** to Part A of the ODCM predates the NRC initiative in GL 89-01 to place the responsibility for procedural details of the required effluent and environmental l i monitoring programs directly with the licensee. The information contained in the current ODCM, Part A reflect the same program details that typically were part of RETS before the revision to 10 CFR 50.36 noted above. Generic Letter 89-01 states that the NRC staff does not intend to repeat technical reviews of the relocated procedural details bran => their consistency ! with the applicable regulatory requirements is a matter of record from past NRC reviews of RETS. The procedural details associated with these programs, which are currently contained in the ODCM, are addressed in GL 89-01, Supplement 1, as procedural details which merit l relocation to the ODCM. Therefore, since the procedural details for sampling and analysis associated with the aforementioned programs are presently contained in the ODCM, the l requirement to obtain NRC approval prior to any changes to Part A is no longer applicable and will be removed so as to meet the intent of GL 89-01. file: sbs901. doc V-4 01C.M Rulf

Seabrook Station License Amendment Request (LAR) 97-08 contains markups of each of the line-item changes proposed for the TS. Table 1 of the LAR outlines the sections of the Technical Specifications that are affected by the proposed change and the final disposition of each item. Proposed revision 18 to the ODCM reflects the changes to these documents for the inclusion of the procedural details of RETS removed from the TS. Conclusion The proposed changes transfer procedural requirements for radiological effluent and environmental monitoring from the Technical Specifications to other licensee-controlled documents (ODCM), along with the addition of programmatic and admmistrative controls for these procedural requirements being placed in the Admmistrative Controls Section of the Technical Specifications. The content of the existing technical requirements concerning RETS is not affected by the proposed changes. Therefore, the proposed changes will not reduce the accuracy or reliability of the dose calculations or setpoint determinations performcd under the existing ODCM since none of these methods are affected by the proposed changes. l l I l l kgg l2 file: sb8901. doc .h QD C {1

p _ l ~. r . l l l t t i

4. DETERMINATION OF SAFETY EVALUATION APPLICABILITY 4.A. Does the nronosed chance make channes in the facility as daccribed in the UFS AR?

u

         . The proposed change'does not modify any of the station facilities as described in the UFSAR.

The relocation of RETS procedural commitments from the TS to the ODCM is an , administrative action that simply moves radioactive effluent and environmental monitoring procedural details from one control document to another. There are no changes being made in

Station facilities, systems, or components as described in the UFSAR. The requirements to maintain programmatic controls on radioactive effluents and environmental monitoring are: added to the Admiuistrative Controls sections of the Technical Specifications (TS) to ensure that regulatory requirements conceming these programs are maintained. Existing procedural

[ details currently in the TS are being moved into the ODCM without substantive modification l to ensure that the existing level of controls over radioactive effluents and environmental monitoring is retained. Changes have been made only to nomenclature as various section numbers need to be edited to reflect that TS sections are now part of the ODCM. The are I. considered editorial changes for clarity only. l l-j- 4.B.' Does the y w-d chmane raaka chaneau in nrn-adures as damerihad in the UFEAR? L The relocation of RETS procedural requirements from the TS to the ODCM is an l administrative action which does not materially change the content of those requirements. l The UFSAR (Sections 1.6,1.8, all of Chapters 11,12,13 and 16) has been reviewed to identify information that could be affected by the relocation of RETS In Section 1.8, compliance with Regulatory Guide 1.21 (page 1.8-7), reference is made to Technical

Specifications 3/4.11.1,3/4.1.2,3/4.11.3,3/4.3.4.8 and 3/4.3.4.9 for the procedural

           . requirements concerning the control and monitoring of radioactive effluents. With the l             relocation of procedural details for these programs to the ODCM editorial changes are needed in the UFSAR Section 1.8 to direct the reader to the new corresponding sections that have been p

created in the ODCM. ' As such, changes are needed in the UFSAR. l N-4 ObCMRNIf( m.:*m m r

l l l 4.C. Does the crocosed change involve tests or exoeriments not described in the UFS AR? The proposed change does not involve tests or experiments not described in the UFSAR. The relocation of RETS procedural commitments from the TS to the ODCM is an administrative action that simply moves radioactive effluent and environmental monitoring procedural details from one control document to another. There are no changes being made in Station facilities, systems, or components. Similarly, no changes are being made in the content of the control and surveillance requirements that currently exist. No new tests or experiments are included in the proposed change. The requirements to maintain programmatic controls on radioactive effluents and environmental monitoring are added to the Admmistrative Control sections of j the TS to ensure that regulatory requirements conceming these programs are maintained. l Existing procedural details currently in the TS are being moved into the ODCM without substantive modification to ensure that the existing level of controls over radioactive effluents and environmental monitoring is retained. Changes have been mule only to nomenclature as various section numbers need to be edited to reflect that TS sections are now part of the ODCM. The are considered editorial changes for clarity only. I

5. OPERATTNG LICENSE l Does the propose change require a change to the existing Operating License (including the Technical Specifications) or are additional Operating License requirements needed?

The proposed change does require changes in the Technical Specifications with the addition of l new sections to the Admmistrative Controls portions (Section 6) that identify the current l regulatory requirements necessary for the control and monitoring of radioactive effluent and radiological environmental monitonng. The proposed change also relocates existing TS procedural details for the control of radioactive effluent and environmental monitoring to other Station control documents, i.e., the ODCM. (LAR 97-08 has been prepared to request NRC review and approval of thae proposed changes.) 1

7. SAFETY EVALUATION -

7;A. Will the probability of an accident previously evaluated in the UFSAR be increased? The probability of an accident previously evaluated in the UFSAR will not be increased. The proposed relocation of procedural details for radioactive emuents and environmental monitoring programs from the technical specifications to the ODCM does not change the existing control and surveillance requirements associated with these programs. There are no physical chcnges or modifications being made to any station system, component, or structure, nor any change in the current implementation methods or operating limits associated with the Radiological Emuent Technical Specifications (RETS). The change is an administrative rearrangement of existing RETS requirements. The current procedural detaib moved from the TS are only edited to the extent that revised sections numbers are assigne6s necessary to sequence the material into the ODCM. Programmatic requirementre maintain radiological environmental and emuent monitor in compliance with Federal regulations is added to the Administrative Controls section of the Technical Specificatioru. 7.B. Will the consequences of an accident previously evaluated in the UFSAR be increased? The consequences of an accident previously evaluated in the UFSAR will not be increased. The proposed relocation of procedural details for radioactive emuents and environmental monitoring programs from the technical specifications to the ODCM does not change the existing control and surveillance requirements associated with these programs. There are no .l physical changes or modifications being made to any station system, component, or structure, l nor any change in the current implementation methods or operating limits associated with the i Radiological Emuent Technical Specifications (RETS). The change is an administrative ! i rearrangement of existing RETS requimments. The cunent procedural details moved from the TS are only edited to the extent that revised sections numbers are assigned as necessary to sequence the materialinto the ODCM. Programmatic requirements to maintain radiological environmental and emuent monitor in compliance with Federal regulations is added to the Administrative Controls section of the Technical Specifications.

me: sbs901.4.c

g i l 7.C. Will the probability or e malfunction of equipment important to safety be increased? ! l l l L j The probability of a malfunction of equipment important to safety will not be increased. The i proposed relocation of procedural details for radioactive effluents and environmental monitoring programs, and solid waste handling, from the technical specifications to the ODCM does not change the existing comrol and surveillance requirements associated with i these programs. There are no physical changes or modifications being made to any station system, component, or structure, nor any change in the current implementation methods or operating limits associated with the Radiological Effluent Technical Specifications (RETS). The change it an administrative rearrangement of existing RETS requirements. The current procedural details moved from the TS are only edited to the extent that revised sections l numbers are assigned as necessary to sequence the materialinto the ODCM. Programmatic requirements to maintain radiological environmental and effluent monitor in compliance with Federal regulations is added to the Adnunistrative Controls section of the Techmcal Specifications. i 1 7.D. Will the consequences of a malfunction of equipment important to safety be increased? j The consequences of a malfunction of equipment important to safety will not be increased. The proposed relocation of procedural details for radioactive effluents and environmental monitoring programs from the technical specifications to the ODCM does not change the existing control and surveillance requirements associated with these programs. There are no physical changes or modifications being made to any station system, component, or stmeture, nor any change in the cunent implementation methods or operating limits associated with the Radiological Effluent Technical Specifications (RETS). The change is an administrative rearrangement of existing RETS requirements. The current procedural details moved from the TS are only edited to the extent that revised sections numbers are assigned as mwy to sequence the material into the ODCM. Programmatic requirements to maintain radiological environmental and effluent monitor in compliance with Federal regulations is added to the Administrative Controls section of the Technical Specifications. s_ va eu

O 7.E Will the possibility of an accident of a different type than any previously evaluated in the UFSAR be created? The possibility of an accident of a different type than previously evaluated in the UFSAR will not be created. The proposed reloca don of procedural details for radioactive effluents and environmental monitoring programs from the technical specifications to the ODCM does not change the existing control and surveillance requirements associated with these programs. - There are no physical changes or modifications being made to any station system, component, or structure, nor any change in the current implementation methods or operating limits associsted with the Radiological Effluent Technical Speci5 cations (RETS). The change is an administotive rearrangement of existing RETS requirements. The current procedural details j moved fron; the TS are only edited to the extent that revised sections numbers are assigned as . j necessary to sequence the material into the ODCM. Programmatic requirements to maintain i radiological envirenmental and effluent monitor in compliance with Federal regulations is added to the Adminisetive Controls section of the Technical Specifications. l 7.F. Will the possibility of a malfunction of a different type than any previously evaluated in the UFSAR be created? The possibility of a malfunction of a different type than any previously evaluated in the l UFSAR is not created. The proposed relocation of procedural details for radioactive effluents and environmental monitoring programs from the technical specifications to the ODCM does not change the existing control and surveillance requirements associated with these programs. l There are no physical changes or modifications being made to any station system, component, i or structure, nor any change in the current implementation methods or operating limits associated with the Radiological Effluent Technical Specifications (RETS). The change is an administrative rearrangement of existing RETS requirements. The current procedural details moved from the TS are only edited to the extent that revised sections numbers are assigned as necessary to sequence the material into the ODCM. Programmatic requirements to maintain radiological environmental and effluent monitor in compliance with Federal regulations is added to the Administrative Controls section of the Technical Specifications. m, \l-IO ODCM ko1F n

i 7.G. Will the margin of safety as defined in the basis for any technical specification be reduced? - No safety margin as defined in the basis for any technical specification is reduced. The basis for all Technical Specifications Limiting Conditions of Operations and Surveillance - Requirements remain the same. The proposed relocation of procedural details for radioactive effluents and environmental monitoring programs from the technical specifications to the ODCM does not change the existing control and surveillance requirements, nor their bases, associated with thesc programs. There are no physical changes or modifications being made to any station system, component, or structure, nor any change in the current implementation methods or operating limits associated with the Radiological Effluent Technical Specifications (RETS). The change is an administrative rearrangement of existing RETS requirements. The current procedural details moved from the TS are only edited to the extent that revised sections numbers are assigned as necessary to sequence the materialinto the ODCM. Programmatic requirements to maintain radiological environmental and effluent monitor in compliance with Federal regulations is added to the Administrative Controls section of the Technical Specifications. The systems, instrumentation, surveillance's and monitoring programs associated with these proposed changes do not meet the design criteria in 10 CFR 50.36 for inclusion in Technical Specifications since they do affect nuclear safety or the ability of the operators to respond to abnormal conditions. m l

4 l

OD tti Reo,t8' m.w,ow y-ti}}

ML20217Q111

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