ML051240100

Yankee Nuclear Power Station, Off-Site Dose Calculation Manual

ML051240100
Person / Time
Site:	Yankee Rowe
Issue date:	04/26/2005
From:	Yankee Atomic Electric Co
To:	Document Control Desk, NRC/FSME
References
BYR 2005-043
Download: ML051240100 (205)

Text

YANKEE NUCLEAR POWER STATION OFF-SITE DOSE CALCULATION MANUAL YANKEE ATOMIC ELECTRIC COMPANY PORC MEETING PREPARED BY/DATE REVIEWED BY/DATE NOJDATE C. L Albright Mark Strum Meeting No. 92-72 REVISION 8 August 19, 1992 August 19, 1992 August 19,1992 REVISION 9 Edward R. Cumming Mark Strum Meeting No. 93-22 May 12,1993 May 18, 1993 May 18, 1993 REVISION 10 R. Brad Harvey Mark Strum Meeting No. 93-28 June 22,1993 June 22,1993 June 22,1993 REVISION 1 March Strum G. M. Babineau Meeting No. 96-63 R October22, 1996 November12,1996 October31, 1996 REVISION 12 R. B. Harvey Mark Strum Meeting No. 97-3 February 7,1997 February 7,1997 January 9, 1997 REVISION 13 M. S. Strum John S. Gedutis Meeting No. 99-19 June 17, 1999 June 17, 1999 June 17,1999 REVISION 14 M. S. Strum G. M. Babineau Meeting No. 00-15 April 13,2000 April 13,2000 April 13,2000 REVISION 15 M. S. Strum G. M. Babineau Meeting No. 01-69 November 19, 2001 December 17,2001 November 19, 2001 REVIION16 ark tru y Mcrting No. 03-65 REVISION 16 August 6, 2003 August 14, 2003 Meeting No. 04-05

,-5er- ohs@ January 29, 2004

REVISION RECORD Revision Date Description 0 12/01/82 Initial printing. Approved by PORC 11/29/82. Submitted for USNRC approval 12/03/82.

1 03/30/84 Change in environmental monitoring sampling locations based on 1983 land use census. Errors in Table 4.1 corrected. Maps revised.

2 07/30/85 Addition of Intercomparison Program description to Section 4.0. Reviewed by PORC 07/30/85.

3 03/19186 Addition of a PVS 1-131 inspection limit to demonstrate compliance with Technical Specification 3.1'1.2.1 .b.

4 05/21/86 Change in milk sampling location. Samples no longer available at Station TM-i1.

5 09/30/86 Change in food product sampling location based on 1986 land use census.

6 02/18188 Change in liquid dose factors to reflect additional dose pathways. Change In gaseous dose factors to reflect five-year average meteorology. Change in gaseous dose rate factors to reflect a shielding factor of 1.0.

Deletion of food product location TF-12 (samples no longer required after 10/31/86). Update of fence line location and several building names and locations in l_ Figure 4-4.

7 05/21/90 Addition of Appendix A which documents the commitments for disposal of septage as provided in YNPS's Application For Approval to Routinely Dispose of Septage under 10CFR Part 20.302, and the NRC's acceptance as transmitted in their Safety Assessment, dated May 17, 1990.

8 08/19/92 a. The following changes were implemented in accordance with NRC Generic Letter 89-01, which provided guidance on the relocation of the Radiological Effluent Technical Specifications to the ODCM:

1. Addition of List of Controls Page (succeeds Table of Contents);

Revision 8

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REVISION RECORD Revision Date Description 8 08/19/92 2. Section 1.0, Introduction updated to reflect the change in scope of the ODCM;

3. Technical Specifications 3/4.0.1, 3/4.0.2, 3/4.0.3, and 3/4.0.4 listed in Section 1.2, Applicability of Controls and Surveillance Requirements (SR), and now referred to as Controls 1.1, 1.2, 1.3, and 1.4, respectively;

4. Table 1.6, Definition of Terms, modified to include definitions pertinent to the relocated Technical Specifications;

5. Tables 1.9, OPERATIONAL MODES, and 1.10, FREQUENCY NOTATIONS, added to Section 1.0;

6. Technical Specification 3/4.1 1.1.1, now referred to as Control 2.1, relocated to Section 2.0;

7. Technical Specifications 3/4.11.1.2, 3/4.11.4, 3/4.11.2.1, 3/4.11.2.2, and 3/4.11.2.3, now referred to as Controls 3.1, 3.2, 3.3, 3.4, and 3.5, respectively, relocated to Section 3.0;

8. Technical Specifications 3/4.12.1, 3/4.12.2, and 3/4.12.3, now referred to as Controls 4.1, 4.2, and 4.3, respectively, relocated to Section 4.0;

9. Technical Specification 3/4.3.3.6, now referred to as Control 5.1, relocated to Section 5.0;

10. Technical Specification 3/4.3.3.7, now referred to as Control 5.2, relocated to Section 5.0 (Existing requirements for explosive gas monitoring instrumentation retained in Technical Specification 314.3.3.7);

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REVISION RECORD Revision Date Description 8 08/19/92 11. Technical Specifications 3/4.11.1.3 and 3/4.11.2.4, now referred to as Controls 6.1 and 6.2, respectively, relocated to Section 6.0;

12. Section 7.0 created to contain reporting details for the Annual Radiological Environmental Monitoring Operating (Control 7.1) and Semiannual Effluent Release Reports (Control 7.2), and Major Changes to the Liquid and GASEOUS RADIOACTIVE WASTE TREATMENT SYSTEMS (Control 7.3); and

.13. Corresponding Technical Specification Bases relocated with Technical Specifications to become part of controls.

b. All pages renumbered.

9 05/18/93 Replacement of milk sampling location TM-12 with TM-14 in Table 4.4 and Figure 4-2.

10 06/22/93 Technical Specification 3/4.3.3.3 (now referred to as Control 5.5) and its Bases relocated to Section 5.0; Technical Specification 3/4.3.3.3 nominal sensor elevations revised to reflect actual measurement heights; Technical Specification 3/4.3.3.3 Bases revised to eliminate reference to protective action recommendations.

11 10/31/96 Surveillance and analyses schedules for both the in-plant Gaseous and Uquid Effluent Monitoring Programs and the off-site REMP have been reduced. These reductions reflect changes in plant configuration due to plant dismantlement and decommissioning activities, and the elimination of radioactive source terms due to the cessation of the fission process with the shutdown of power operations.

12 02/07/97 The requirement to submit an annual summary of hourly meteorological data with the Semi-annual Radioactive Effluent Release Report due 60 days after January 1 of each year was eliminated.

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REVISION RECORD Revision Date Description 13 6/17199 Elimination of the Turbine Building composite sampler and analysis requirements due to change In plant configuration. Elimination of milk sampling location due to changes identified in the annual land use census.

Clarification of actions necessary to perform maintenance and testing on an in-line rad-monitor, plus expanded flexibility of operating Aux. Service Water for Spent Fuel Pit cooling while the effluent rad-monitor is out of service. Correction to block diagram (Figure 6-1) to reflect current liquid waste processing system configuration. Change in the reporting requirement for the Radioactive Effluent Release Report from semiannual to annual based on Technical Specification Amendment No. 151. Editorial changes to improve readability and correct or eliminate unnecessary text.

14 4/13/00 Clarifications to liquid effluent monitoring ACTION STATEMENTS I and 4 (Table 5.1) are included to provide clearer guidance on the existing intent of actions needed if the liquid radiation effluent monitor is out of service.

15 11/19/01 1. Eliminates the requirement to maintain the 200 foot On-Site Meteorological Monitoring System (Section 5.5).

2. References changed to reflect the relocation of Technical Specification Section 6.7 into the Yankee Decommissioning Assurance Program, Appendix D, Administrative Controls.

3. Change the auxiliary service water flow rate to 120 gpm to reflect current operating performance and deletes the use of pump and valve curves to estimate flow rate.

Revision 16 I

REVISION RECORD Revision Date Description 16 8/14/03 1. Eliminates detection requirements for radionuclides which, by natural decay, are no longer a significant effluent or of environmental concern.

2. Eliminates milk sampling requirements from REMP.

3. Adds new liquid waste treatment and discharge conditions for decommissioning of structures including the SFP.

4. Updates, site maps and monitoring programs to recognize the new ISFSI operations.

5. Eliminates Control limits and measurement requirements for dissolved and entrained noble gases in liquid waste.

6. Updates surveillance requirements for liquid effluent releases.

I 17 1/29/04 1. Elimination of Noble Gas effluent dose and dose rate calculation methodology and Control Limit requirements due to removal of source potential from plant systems.

2. Elimination of Tritium In gaseous effluent dose and dose rate calculation methodology due to removal of source potential from plant systems.

3. Elimination of gaseous ventilation exhaust treatment system operability and gaseous monitoring instrumentation surveillance requirements (primary vent stack) due to dismantle of system equipment.

4. Update and expansion of dose and dose rate conversion factors for use in Method I dose projections of airborne particulates released to the Revision 17 I

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REVISION RECORD Revision Date Description atmosphere during building demolition.

5. Update site boundary figure (Fig. 1-2) to reflect changes in effluent release points.

6. Update of site specific historical atmospheric dispersion factors to reflect removal of the plant vent stack (mixed mode release height) and accommodate ground level release conditions with no building wake credits.

7. Elimination of the need for gaseous waste sampling and analysis requirements due to the removal of the plant vent stack release point.

8. Inclusion of supplemental environmental air particulate monitoring in the vicinity of building demolition activities to validate effluent release model assumptions and modeling.

9. Minor editorial changes to improve referencing in the ODCM.

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LIST OF AFFECTED PAGES Page IRev. No.I Page I Rev. No.I Page I Rev. No.I Page I Rev. No.

. = ., = - = - = . = - =

Cover 17 1-10 17 3-9 17 3-38 17 ii 8 1-11 17 3-10 17 3-39 17 iii 8 1-12 17 3-11 17 3-40 17 iv 12 1-13 17 3-12 17 3-41 17 V 16 1-14 11 3-13 17 3-42 17 vi 17 1-15 17 3-14 16 vii 17 1-16 17 3-15 16 4-1 15 viii 17 1-17 16 3-16 11 4-2 16 ix 17 1-18 17 3-17 11 4-3 15 x 17 1-19 11 3-18 11 4-4 11 xi 17 1-20 17 3-19 16 4-5 16 xli 17 1-21 17 3-20 16 4-6 16 xiii 17 3-21 17 4-7 16 xiv 17 2-1 16 3-22 17 4-8 15 xv 17 2-2 16 3-23 17 4-9 15 xvi 17 2-3 16 3-24 17 4-10 15 Xvii 17 2-4 16 3-25 17 4-11 17 xviii 17 2-5 16 3-26 17 4-12 15 xix 17 2-6 16 3-27 17 4-13 17 2-7 16 3-28 17 4-14 16 1-1 17 3-29 17 4-15 17 1-2 17 3-1 15 3-30 17 4-16 17 1-3 17 3-2 15 3-31 17 4-17 17 1-4 17 3-3 15 3-32 17 4-18 17 1-5 17 3-4 11 3-33 17 4-19 17 1-6 17 3-5 17 3-34 17 4-20 17 1-7 17 3-6 17 3-35 17 4-21 17 1-8 17 3-7 17 3-36 17 4-22 17 1-9 17 3-8 17 3-37 17 Revision 17 I

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LIST OF AFFECTED PAGES

• (Continued)

Page Rev. No. I Page IRev. No. I Page I Rev. No. I Page Rev. No 5-1 15 6-8 17 A-20 B-1 ii 5-2 16 A-21 B-2 11 5-3 16 7-1 17 A-22 B-3 11 5-4 16 7-2 17 A-23 B-4 11 5-5 16 7-3 13 A-24 B-5 11 5-6 16 7-4 17 A-25 B-6 11 5-7 16 7-5 17 A-26 B-7 Ii 5-8 16 A-27 B-8 11 5-9 16 8-1 17 A-28 B-9 11 5-10 17 A-29 B-10 11 5-11 17 A-1 7 A-30 B-il 11 5-12 17 A-2 7 A-31 5-13 17 A-3 7 A-32 5-14 17 A-4 7 A-33 5-15 17 A-5 7 A-34 5-16 11 A-6 7 A-35 5-17 11 A-7 7 A-36 5-18 16 A-B 7 A-37 5-19 11 A-9 7 A-38 5-20 11 A-10 7 A-39 5-21 17 A-1I 7 A-40 A-12 7 A-41 6-1 15 A-13 7 A-42 6-2 15 A-14 7 A-43 6-3 17 A-15 7 A-44 6-4 17 A-16 7 A-45 6-5 17 A-17 7 A-46 6-6 17 A-18 7 A-47 6-7 16 A-19 7 A-48

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DISCLAIMER OF RESPONSIBILITY This document was prepared for use 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 respect to anV 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 Inthis document.

Revision 17 I

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ABSTRACT The Yankee Nuclear Power Station (YNPS) OFF-SITE DOSE CALCULATION MANUAL (ODCM) contains the methodology and parameters used in the calculation of off-site doses resulting from radioactive gaseous and liquid effluents, in the calculation of liquid effluent I

monitoring alarmn/trip setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM also contains (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Appendix D, Section B.4, of the Yankee Decommissioning Quality Assurance Program (YDQAP) and (2)descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release required by controls 7.1 and 7.2, respectively. With initial approval I

by the U.S. Nuclear Regulatory Commission and the YNPS Plant Operation Review Committee (PORC) and approval of subsequent revisions by the ISR and the Site Manager, this manual Is I

suitable to show compliance where referred to by the Yankee Decommissioning Quality Assurance Programs and controls listed in this document.

Revision 17 I

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TABLE OF CONTENTS REVISION RECORD ........................................................... ii LIST OF AFFECTED PAGES .......................................................... viii DISCLAIMER OF RESPONSIBIUTY ........................................................... x ABSTRACT .......................................................... xi TABLE OF CONTENTS .......................................................... xii LIST OF CONTROLS .......................................................... xvi LIST OF TABLES .......................................................... xvii LIST OF FIGURES ............ ;xx

1.0 INTRODUCTION

.1-1 1.1 Summary of Methods, Dose Factors, Constants, Variables, and Defintions. 1-1 1.2 Applicability of Controls and Surveillance Requirements (SR).1-2 2.0 RADIOACTIVE LIQUID EFFLUENTS .. 2-1 2.1 Off-Site Concentrations .2-1 2.2 Method to Calculate Off-Site Liquid Concentrations .2-5 2.3 Method to Determine Radionuclide Concentration for Each Uquid Effleunt Pathway 2-6 2.3.1 Test Tank Pathway .2-6 2.3.2 Auxiliary Service Water System Pathway ........................................ ;.2-7 2.3.2 Remaining Pathways .2-7 3.0 DOSE/DOSE RATE CONTROLS AND CALCULATIONS .3-1 3.1 Dose Due to Radioactive Liquid Effluents .3-1 3.2 Total Dose Due to Radioactive Uquid and Gaseous Effluents .3-3 3.3 Dose Rate Due to Radioactive Gaseous Effluents .3-5 3.4 Dose Due to Noble Gases Released in Radioactive Gaseous Effluents (Deleted)3-p 3.5 Dose Due to Radionuclides in Particulate Form With Half-Lives Greater than Eight Days .3-11 3.6 Dose Calculation Concepts .3-13 3.7 Method to Calculate the Total Body Dose from Liquid Releases .3-14 3.7.1 Method I ......................................................................... 3..................

3-14 3.7.2 Methodll.3-15 3.7.3 Basis for Method I.3-15 Revision 17

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TABLE OF CONTENTS (Continued)

Page 3.8 Method to Calculate Maximum Organ Dose from Liquid Releases ................. 3-19 3.8.1 Method I.................................................. 3-19 3.8.2 Method 11 .................................................. 3-20 3.8.3 Basis for Method I.................................................. 3-20 3.9 Method, to Calculate the Total Body Dose Rate from Noble Gases . ............... 3-21 3.9.1 Method I (Deleted) .................................................. 3-21l 3.9.2 Method II (Deleted) .................................................. 3-22 3.9.3 Basis for Method I (Deleted) .................................................. 3-22 3.10 Method to Calculate the Skin Dose Rate from Noble Gases . ....................

3-24 3.10.1 Method I (Deleted) .................................................. 3-24 3.10.2 Method II (Deleted) .................................................. 3-25 3.10.3 Basis for Method I (Deleted) .................................................. 3-25 3.11 Method to Calculate the Critical Organ Dose Rate from Particulates with Half-Lives Greater Than Eight Days ... 3-27 3.11.1 MethodI .................... . 3-27 3.11.2 Method 11 .3-28 3.11.3 Basis for Method I.3-28 3.12 Method to Calculate the Gamma Air Dose from Noble Gases (Kr-85) ............. 3-30 3.12.1 Method I (Deleted) .................................................. 3-30 3.12.2 Method II (Deleted) .................................................. 3-30 3.12.3 Basis for Method I (Deleted) .................................................. 3-30 3.13 Method to Calculate the Beta Air Dose from Noble Gases . .....................3-31 3.13.1 Method I (Deleted) .......................... l3-31..

3.13.2 Method II (Deleted) ......................... 3-31 3.13.3 Basis for Method I (Deleted) ......................... 3-31 Revision 17

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TABLE OF CONTENTS (Continued)

Page 3.14 Method to Calculate the Critical Organ Dose from Particulates ...................... 3-32 3.14.1 Method I........................................................ 3-32 3.14.2 Method11 ........................................................ 3-33 3.14.3 Basis for Method I........................................................ 3-33 3.15 Critical Receptors and Long-Term Average Atmospheric Dispersion Factors for Important Exposure Pathways .................................................. 3-38 3.15.1 Critical Receptors .................................................. 3-38 3.15.2 Yankee Atmospheric Dispersion Model .............................................. 3-39 3.15.3 Long-Term Average Dispersion Factors for Critical Receptors ........... 3-40 3.16 Method to Calculate Direct Dose from Plant Operation .- 3-42 4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING . ...............................

4-1 4.1 Monitoring .4-1 4.2 Land Use Census .4-10 4.3 Intercomparison Program .4-12 4.4 Environmental Monitoring Locations .4-12 5.0 INSTRUMENTATION .5-1 5.1 Radioactive Liquid Effluents .5-1 5.2 Radioactive Gaseous Effluents (Deleted) .. .................................... . 5-10 5.3 Liquid Effluent Instrumentation Setpoints .5-16 5.3.1 Method (Deleted) .5-16 5.3.2 Liquid Effluent Setpoint Example (Deleted).........................................5-17 5.3.3 Basis (Deleted) .5-18 5.4 Gaseous Effluent instrumentation Setpoints . .5-21 5.4.1 Method (Deleted)..........? ..... 5-21 5.4.2 Gaseous Effluent Setpoint Example (Deleted) .5-21 5.4.3 Basis (Deleted) .5-21 Revision 17

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TABLE OF CONTENTS (Continued)

Paqe 6.0 RADIOACTIVE WASTE TREATMENT SYSTEMS, EFFLUENT PATHWAYS, AND RADIATION MONITORS .................................................. 6-1 6.1 Liquid Radioactive Waste Treatment .6-1 6.2 Gaseous Radioactive Waste Treatment (Deleted) .. 3 6.3 Uquid Effluent Streams, Radiation Monitors, and Radioactive Waste Treatment Systems .6-5 6.4 In-Plant Uquid Effluent Pathways .6-5 6.5 In-Plant Gaseous Effluent Pathways (Deleted) .6-6 7.0 REPORTING REQUIREMENTS .7-1 7.1 Annual Radiological Environmental Operating Report .7-1 7.2 Annual Radioactive Effluent Release Report .7-2 7.3 Major Changes to Uquid Radioactive Waste Treatment Systems. 7-4 7.4 Special Reports .7-5

8.0 REFERENCES

.8-1 APPENDIX A: DISPOSAL OF SEPTAGE ............................... A-1 APPENDIX B CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND .................. B-1 Revision 17 I

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LIST OF CONTROLS Control Title Paae 1.1 Applicability of Controls and Surveillance Requirements 1-2 1.2 Applicability of Controls and Surveillance Requirements 1-2 2.1 Off-Site Concentrations 2-1 3.1 Dose Due to Radioactive Liquid Effluents 3-1 3.2 Total Dose Due to Radioactive Liquid and Gaseous Effluents 3-2 3.3 Dose Rate Due to Radioactive Gaseous Effluents 3-5 3.4 Dose Due to Noble Gases Released in Radioactive Gaseous Effluents 3-9 (Deleted) 3.5 Dose Due to Radionuclides In Particulate Form With Half-Lives Greater than Eight Days 3-11 4.1 Monitoring Program 4-1 4.2 Land Use Census 4-10 4.3 Intercomparison Program 4-12 5.1 Radioactive Liquid Effluents 5-1 5.2 Radioactive Gaseous Effluents (Deleted) 5-10 6.1 Uquid Radioactive Waste Treatment 6-1 6.2 Gaseous Radioactive Waste Treatment (Deleted) 6-3 7.1 Annual Radiological Environmental Operating Report 7-1 7.2 Annual Radioactive Effluent Release Report 7-2 7.3 Major Changes to Liquid and Gaseous Radioactive Waste Treatment Systems 7-4 7.4 Special Reports 7-5 Revision 17

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LIST OF TABLES Table Title Page 1.1 Summary of Concentration and Setpoint Methods, and Method I Dose Equations for Normal Operations at the Yankee Plant 1-3 1.2 Dose Factors Specific to the Yankee Plant for Noble Gas Releases (Deleted) 1-6 1.3 Summary of Radiological Effluent Controls and Implementing Equations 1-7 1.4 Summary of Constants (Deleted) 1-9 1.5 Summary of Variables 1-10 I

1.6 Definition of Terms 1-14 1.7 Dose Factors Specific to the Yankee Plant for Liquid Releases 1-17 1.8 Dose and Dose Rate Factors Specific to the Yankee Plant for Particulate Gaseous Releases 1-18 1.9 Frequency Notation 1-19 2.1 Radioactive Liquid Waste Sampling and Analysis Program 2-3 3.1 Radioactive Gaseous Waste Sampling and Analysis Program 3-7 3.2 Environmental Parameters for Liquid Effluents at Yankee Rowe (Derived from Reference A) 3-17 3.3 Age-Specific Usage Factors for Various Liquid Pathways at Yankee .Rowe 3-18 3.4 Age-Specific Usage Factors 3-35 3.5 Environmental Parameters for Gaseous Effluents at the Yankee Plant 3-36 3.6 Yankee Nuclear Power Station Five-Year Average Atmospheric Dispersion Factors 3-41 I

4.1 Radiological Environmental Monitoring Program 4-4 4.2 Reporting Levels for Radioactivity Concentrations in Environmental Samples 4-6 Revision 17 I

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LIST OF TABLES (Continued)

Table Title Page 4.3 Detection Capabilities for Environmental Sample Analysis 4-7 4.4 Radiological Environmental Monitoring Stations 4-13 5.1 Radioactive Uquid Effluent Monitoring Instrumentation 5-3 5.2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 5-7 5.3 Radioactive Gaseous Effluent Monitoring Instrumentation (Deleted) 5-12 5.4 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements (Deleted) 5-14 Revision 17 I

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LIST OF FIGURES Table Title Page 1-1 Yankee Atomic Electric Company, SITE BOUNDARY LINES 1-20 1-2 Yankee Atomic Electric Company, Effluent Discharge Points, Site Plot Plan 1-21 4-1 Yankee Plant Radiological Environmental Monitoring Locations Within 1 Mile (Airbome, Waterbome, and Ingestion Pathways) 4-16 4-2 Yankee Plant Radiological Environmental Monitoring Locations Within 12 Miles (Airbome, Waterbome, and Ingestion Pathways) 4-17 4-3 Yankee Plant Radiological Environmental Monitoring Locations Outside 12 Miles (Airborne, Waterbome, and Ingestion Pathways) 4-18 4-4 Yankee Plant Radiological Environmental Monitoring Locations at the Restricted Area Fence (Direct Radiation Pathway) 4-19

' 4-5 Yankee Plant Radiological Environmental Monitoring Locations Within 1 Mile (Direct Radiation Pathway) 4-20 4-6 Yankee Plant Radiological Environmental Monitoring Locations Within 12 Miles (Direct Radiation Pathway) 4-21 4-7 Yankee Plant Radiological Environmental Monitoring Locations Outside 12 Miles (Direct Radiation Pathway) 4-22 6-1 Liquid Effluent Streams, Radiation Monitors, and Radioactive Waste Treatment System at the Yankee Plant 6-7 6-2 Gaseous Effluent Streams, Radiation Monitors, and Radioactive Waste Treatment System at the Yankee Plant (Deleted) 6-8 Revision 17

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1.0 INTRODUCTION

According to Definition of Terms (Table 1.6), the OFF-SITE DOSE CALCULATION MANUAL (ODCM) contains the methodology and parameters used in the calculation of off-site doses resulting from radioactive gaseous and liquid effluents, in the calculation of liquid effluent monitoring alarm/trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM also contains: (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Program required by Appendix D, Section B.4, of the Yankee Decommissioning Quality Assurance Program (YDQAP) and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Controls 7.1 and 7.2, respectively. The ODCM forms the basis for plant procedures which document the off-site doses due to plant decommissioning activities which are used to show compliance with the numerical guides for design controls of Section II,Appendix I, 10CFR Part 50.

The methods contained herein follow accepted NRC guidance, unless otherwise noted In the text. The basis for each method Is sufficiently documented to allow regeneration of the methods by an experienced health physicist.

All changes to the ODCM shall be reviewed by an Independent Safety Reviewer and approved by the Site Manager in accordance with YDQAP, Appendix D, Section B.5.d prior to implementation. Changes made to the ODCM shall be submitted to the Commission for their information Inthe Annual Radioactive Effluent Release Report for the period in which the change(s) was made effective.

1.1 Summary of Methods, Dose Factors, Umits, Constants, Variables, and Definitions This section summarizes the methods for the user. In addition, the applicability of controls and surveillance requirements are listed in this section. The first time user should read Chapters 2 through 5. The concentration and setpoint methods are documented in Table 1.1, as well as the Method I dose equations. Where more accurate dose calculations are needed, use the Method II for the appropriate dose as described in Sections 3.7 through 3.14 and 3.16. The dose factors used in the equations are in Tables 1.2,1.7, and 1.8 and the regulatory limits are summarized in Table 1.3. The constants, variables, special definitions, and FREQUENCY NOTATION used in the Revision 17 1-1

ODCM are in Tables 1.4, 1.5, 1.6, and 1.10, respectively. Lastly, Figures 1-1 and 1-2 depict the I Yankee plant site boundary line and liquid effluent discharge points, respectively.

1.2* Applicability of Controls and Surveillance Requirements (SR)

Control 1.1 The controls and ACTION requirements shall be applicable during conditions I

specified for each control.

Control 1.2 Adherence to the requirements of the controls andlor associated ACTION within the specified time interval shall constitute compliance with the control. In the event that the control is restored prior to expiration of the specified time interval, completion of the ACTION statement is not required.

SR 1.1 Surveillance requirements shall be applicable during the conditions specified for individual controls.

SR 1.2 Each surveillance requirement shall be performed within the specified time interval with:

a. A maximum allowable extension not to exceed 25 percent of the surveillance interval, and

b. Deleted SR 1.3 Performance of a surveillance requirement within the specified time interval shall constitute compliance with OPERABILITY requirements for a control and associated ACTION statements unless otherwise required by the control.

Revision 17 1-2

TABLE 1.1 Summary of Concentration and Setpoint Methods, and Method I Dose Equations for Normal Operations at the Yankee Plant Equation No. Maximum Equation(a) 2-1 Unrestricted Area, Total Fraction of MPC in ENG cENG Uiquids F. =1 I~ MPC1 2-2 Deleted I

3-1 Total Body Dose Due to Liquids D,, (mrem)= K QI DFL1 ,

3-2 Maximum Organ Dose Due to Liquids Dom,,, (mrem) = K E Qj DFLm.

3-3 Deleted 3-4 Deleted 3-5 Critical Organ Dose Rate Due to (mremrn Particulates with T' > 8 Days e Dcor yr J- QRI DFG'O 3-6 Deleted 3-6.1 Deleted 3-7 Deleted 3-8 Critical Organ Dose Due to H-3 and D. (mrem) = E Qj DFGj,,

Particulates with 114 > 8 Days l Revision 17 I

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TABLE 1.1 (Continued)

Summary of Concentration and Setnoint Methods, and Method I Dose Equations for Normal Operations at the Yankee Plant Equation No. Maximum Equation(a) 5-1 Liquid Release Rate Reading

.R= f (MPCc) (Se) 5-3 Deleted 5-4 Deleted Note (a):

C} Concentration of radionuclide Si' in a mixture (1ICi/ml).

CI = Concentration of radionuclide wi', at the point of discharge.

C~NG -

Concentration of radionuclide bi", at the point of discharge.

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TABLE 1.1 (Continued)

Summary of Concentration and Setpoint Methods, and Method I Dose Equations for Normal Operations at the Yankee Plant DFG, = Site-specific, critical organ dose factor for a gaseous release of radionuclide ri".

DFGs - Site-specific, critical organ dose rate factor for a gaseous release of radionuclide 'it .

DRl = Site-specific, total body dose factor for a liquid release of radionuclide Ni.

DFLhO = Site-specific, maximum organ dose factor for a liquid release of radionuclide ji'.

f, = Flow rate past the test tank monitor (gpm).

f2 = Flow rate at the point of discharge (gpm).

K = Deerfield River flow rate correction factor.

MPCC = Composite MPG for the mix of radionuclides (gCVml).

EC,

= =& (Eq. 5-2) vC, MPCG 01 = Total release (Curies) for radionuclide wiO.

OI = Release rate (ILCVsec) for radionuclide i.

St = Uquid instrumentation response factor (cprnV(gCVcc)).

Revision 17 1-5

TABLE 1.2 Dose Factors Specific to the Yankee Plant for Noble Gas Releases Table Deleted I

Revision 17 1-6

TABLE 1.3 Summary of Radiological Effluent Controls and Implementing Equations Control Category Method Limit 2.1 Off-Site Total Fraction of MPC Eq. 2-1 *1.0 Concentrations Deleted Deleted Deleted of Liquids__ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _

3.1 Dose Due to Total Body Dose Eq. 3-1

• 1.5 mrem in a qtr.

Liquid Effluents

• 3.0 mrem in a yr.

Organ Dose Eq. 3-2

• 5.0 mrem in a qtr.

• 10.0 mrem in a yr.

3.2 Total Dose Due Total Body Dose Eq. 3-1

• 25.0 mrem in a yr.

to Liquid and Eq. 3-6 Gaseous Eq.3-9 Effluents Organ Dose Eq. 3-2

• 25.0 mrem in a yr.

Eq. 3-8 Eq. 3-9 Thyroid Dose Eq. 3-2

• 75.0 mrem in a yr.

Eq. 3-8 Eq. 3-9 3.3 Dose Rate Due Deleted to Gaseous Effluents Deleted Organ Dose Rate Due to Eq. 3-5 *150 mrem Particulates with TY2 > 8 _ 0.0 Days yr 3.4 Dose Due to Deleted Noble Gases in Gaseous Effluents Deleted 3.5 Dose Due Organ Dose Due to Eq. 3-8

• 7.5 mrem in a qtr.

Particulates In Particulates with TY2 > 8

• 15.0 mrem in a yr.

Gaseous Days Effluents 5.1 Liquid Effluent Alarm/Trip Setpoint Eq. 5-1 Control 2.1 Monitor Setpoint I I Revision 17 I

1-7

TABLE 1.3 (Continued)

Summary of Radiological Effluent Controls and Implementing Equations 1* r Control Category Method Limit 5.2 Gaseous Deleted I Effluent Monitor Setpoint Deleted I

6.1 Liquid Total Body Dose Eq. 3-1. 0.06 mrem in a mo.

Radioactive Organ Dose Eq. 3-2

• 0.2 mrem In a mo.

Waste Treatment 6.2 Gaseous Deleted I

Radioactive Waste Treatment Deleted Deleted Revision 17 I

1-8

TABLE 1.4 Summary of Constants Deleted Revision 17 I

1-9

TABLE 1.5 Summary of Variables Variable Definition Units 1.00 x 1OC = Number of picocuries per microcurie. pci 31.54 = 10X 06 ( Ci '3154x10 7 ( secQ C- Sec 31.54 I -0)r )ic - yr Deleted NG = Concentration of radionuclide NIn, at the point of tic0 discharge. cc C = Concentration of radionuclide i. ttCVm3 or jiCvcc Deleted Deleted Deleted D, -= Dose to the critical organ. mrem D n= Dose to the maximum organ. mrem Deleted I

De, - Dose to the total body. mrem Deleted Deleted Deleted Revision 17 I

1-10

TABLE 1.5 (Continued)

Summary of Variables Variable Definition Units Deleted Deleted DFGiCw= Critical organ gaseous dose factor for radionuclide mrem ia. Ci DFGk = Critical organ gaseous dose rate factor for mrem - sec radionuclide id. ICi - yr DFL" = Maximum organ liquid dose factor for radionuclide mrem Ni.. Ci DFL4 = Total body liquid dose factor for radionuclide "i'. mrem Ci D., = Critical organ dose rate due to particulates (T1/2 > 8 mrem days) in gaseous effluents. yr I Deleted Deleted DIQ = Deposition factor for dry deposition of particulates. sec m2 Deleted Revision 17 I

1-11

TABLE 1.5 (Continued)

Summary of Variables Variable Definition Units Ft= Total fraction of MPC in liquid pathways.

Deleted f= Flow rate past the test tank monitor. gpm f2 Flow rate at the point of discharge. gpm MPCc = Composite MPC for the mix of radionuclides. See toCi Equation 5-2. cc MPCI = Maximum permissible concentration of radionuclide tiCi Wi"(10CFR Part 20, Appendix B, Table 2, Column 2, cc see Appendix B of the ODCM).

Q Release for radionuclide 'iT . Ci Deleted QN= Release rate for radionuclide fig. lO sec X/Q = Average undepleted dispersion factor. sec p[/Q]D = Average depleted dispersion factor. sec m3 Deleted SF = Shielding factor.

Revision 17.

1-12 I

TABLE 1.5

• (Continued)

Summary of Variables Variable Definition Units Deleted St = Uquid monitor response factor. cpm

. Cilcc Revision 17 1-13

TABLE 1.6 Definition of Terms The defined terms of this section appear in capitalized type and are applicable throughout this document ACTION ACTION shall be those additional requirements specified as corollary statements to each principle control and shall be part of the controls.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel, including the alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY, including alarm and/or trip functions.

FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of surveillance requirements shall correspond to the intervals defined InTable 1.9.

MEMBER(S) OF THE PUBLIC MEMBER(S) OFTHE PUBLIC (for purposes of 10CFR50, Appendix I) shall include all persons who are not occupationally associated with the plant. This category does not Include employees of the utility, its contractors, or vendors. Also excluded from this category, are persons who enter the site to Revision 11 1-14

TABLE 1.6 (Continued)

Definition of Terms service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the site operations or decommissioning of the plant.

OFF-SITE DOSE CALCULATION MANUAL (ODCM)

The ODCM contains the methodology and parameters used in the calculation of off-site doses resulting from radioactive gaseous and liquid effluents, in the calculation of liquid effluent monitoring alarm/trip setpoints, and In the conduct of the Environmental Radiological Monitoring Program. The ODCM also contains (1)the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Appendix D, Section B.4 of the YDQAP and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Controls 7.1 and 7.2, respectively.

OPERABLE - OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, electric power, cooling or seal water, lubrication, or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function(s) are also capable of performing their related support function(s).

SITE BOUNDARY The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee. Any area within the SITE BOUNDARY used for residential quarters or recreational purposes shall be considered to be beyond the SITE BOUNDARY for purposes of meeting gaseous effluent dose controls. (Realistic occupancy factors shall be applied at these locations for the purposes of dose calculations.)

SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to radiation.

Revision 17 1-15

TABLE 1.6*

(Continued)

Definition of Terms Page Intentionally Left Blank Revision 17 I

1-16

TABLE 1.7 Dose Factors Specific to the Yankee Plant for Liquid Releases Total Body Dose Maximum Organ Factor Dose Factor Radionuclide DFL (mrm) DFL (mr H-35.99x10' 5.99 x 10 4 C-14 1.64 x 10` 0 8.18 x 10+0 I.

Fe-55 3.46 x 102 2.11 x 10.'

Co-60 2.79 x 10-1 9.04 x 10.1 Sr-90 6.97 x 10+1 2.75 x 10+2 Ag-1 Om 2.32 x 102 2.21 x 10 Cs-134 1.79 x 10+' 2.40 x 10+'

Cs-137 1.07 x 10+1 2.07 x 10+'

Ag-108m/Ag-108 5.70 x 10-1 1.81 x 10+1 Revision 16 I

1-17

TABLE 1.8 Dose and Dose Rate Factors SPecific to the Yankee Plant for Particulate Gaseous Releases I

Critical Organ Critical Organ Dose Factor Dose Rate Factor DFG (mrem) DFG' mrem - sec Radionuclide 10~C ~ d-y H-3 5.60E-03 1.77E-01 C-14 4.77E+00 1.50E+02 Mn-54 3.73E+00 1.46E+02 Fe-55 1.70E+00 5.36E+-1 Co-60 2.01 E+01 8.17E+02 Ni-63 9.54E+01 3.01 E+03 Sr-90 3.41 E+03 1.08E+05 Ru-1 06 1.86E+02 5.87E+03 Ag-11 Om 1.75E+01 6.24E+02 Sb-i125 4.96E+00 1.85E+02 Cs-137 1.75E+02 5.58E+03 Pu-238 9.17E+03 2.89E+05 Pu-239 1.06E+04 3.34E+05 Pu-240 1.05E+04 3.31 E+05 Am-241 1.08E+04 3.41 E+05 Cm-242 4.09E+02 1.29E+04 Cm-243 7.29E+03 2.30E+05 Cm-244 5.63E+03 1.78E+05 A1 Revision 17 I

1-18

• TABLE 1.9 Frequency Notation Notation Frequency S At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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

W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184 days.

R At least once per 18 months.

P Prior to each release.

N.A. Not applicable.

Revision 11 1-19

Figure 1-1 Yankee Nuclear Power Station Site Boundary Lines Revision 17 1-20

Figure 1-2 Yankee Nuclear Power Station Effluent Discharge Points Site Plot Plan Revision 17 l 1-21

2.0 RADIOACTIVE LIQUID EFFLUENTS

.2.1 Off-Site Concentrations Control 2.1 In accordance with Yankee Decommissioning Quality Assurance Program (Appendix D, Section B.4.a), the concentration of radioactive material released to Unrestricted Areas (see Figure 1-2) shall be limited to the concentrations specified in IOCFR Part 20, Appendix B, Table II, Column 2, for all.

ApDlicability At all times.

ACTION With the concentration of radioactive material released from the site to Unrestricted Areas exceeding the above limits, without delay, take actions to restore the concentration to within the above limits.

Surveillance Requirements SR 2.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 2.1.

SR 2.1.2 The results of radioactive analysis shall be used in accordance with the methods of the ODCM to assure that concentrations at the point of release are maintained within the limits of Control 2.1.

Bases Control 2.1 is provided to ensure that the at any time concentration of radioactive materials released in liquid waste effluents from the site above background (unrestricted areas for liquids is at the point of discharge from the plant discharge structure Into Sherman Pond) will be less than the concentration levels specified in 10CFR Part 20, Appendix B, Table 11, Column 2 (Appendix B of the ODCM contains a listing of these values as taken from the regulations). These requirements provide operational flexibility, compatible with considerations Revision 16 2-1

of health and safety, which may temporarily result in releases higher than the absolute value of the concentration numbers in Appendix B, but still within the annual average limitation of the revised (Jahuary 1, 1993 effective date) 10CFR, Part 20, regulation. Compliance with the design objective doses of Section I.A of Appendix I to 10CFR, Part 50, assure that doses are maintained ALARA, and that annual concentration limits of Appendix B to 10CFR20.1001-20.2401 will not be exceeded.

I Revision 16 I

2-2

TABLE 2.1 Radioactive Uquid Waste Samplina and Analvsis Proaram Lower Limit Minimum Type of of Detection Sampling Analysis Activity LLD(a)

Uquid Release Type Frequency Frequency Analysis (AiCVml)

A. Batch Waste P P Principal 5.00 x 10 7 Release TanksF Each Batch Each Batch Emitters(m draindown viaDetdDltd batch effluent DeletedDeleted test tanks in p M Tritium 1.00 x 1 05 batch mode) Each Batch Composite(c) Gross Alpha 1.00 x le 7 P Q Sr-90 5.00x109 I Each Batch Composite(c) Fe-55 1.00 x 104 B. Plant Continuous Releases(e) Deleted Turbine Building (Pathway Abandoned)

Sump C. Plant Continuous Continuous(d) M/2 Principal Gamma 5.00 x 10 Releases(e) Composite" Emitters(f)

Continuous() M Tritium 1.00 x 105 Composite(d) Gross Alpha 1.00 x 10' 7 SFP Drain Down Continuous(d Q Sr-90 5.00 x 108 Skid for Direct Composite(Q Fe-55 1.00 x 104 Release or Addition to TK-39(g)

D. Construction P P Principal Gamma 5.00 x 10 '

Dewatering° Each Batch Each Batch Emitters(0 Tritium 1.00 x 10'5 P M Gross Alpha 1.00 x 10o Each Batch Composite(c)

P Q Sr-90 5.00 x 108 Each Batch Composite(c) Fe-55 1.00 x 10-6 Revision 16 2-3 I

TABLE 2.1-(Continued)

Notation

a. The LLD is defined in Table Notation (a)of Table 4.3 of SR 4.1.

b. A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analysis, each batch shall be isolated and thoroughly mixed to assure representative sampling. For construction dewatering sources, a sample aliquot for every 1000 gallons (or less) of water transferred to temporary holiday tanks or basins not capable of recirculation or internal mixing shall be collected and composited to satisfy representative sampling requirements. Alternately, if three separate samples taken from the dewatering source indicate that the gamma emitter/tritium radioactivity does not vary by more than a factor of two, then the dewatering operation maybe considered as a continuous discharge where a composite sampler on the discharge line will collect a representative sample for analysis following the release.

c. 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 which is representative of the liquids released. If there is no effluent discharge during the period, no composite sample of collected waste is required.

d. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the average effluent release.

e. A continuous release is the discharge of liquid wastes of a nondiscrete volume; e.g.,

from a volume or system that has an input flow during periods when flow exist through the system.

f. The principal gamma emitters for which the LLD requirement applies exclusively are the following radionuclides: Co-60, Cs-134 and Cs-137. This list does not mean that only these radionuclides are to be detected and reported. Other peaks that are measurable and identifiable, together with the above radionuclides, also shall be identified and reported. Radionuclides that are below the LLD for the analyses should not be reported as being present at the LLD level.

9. If SFP drain down process flow is halted to allow for SFP hydrohazing, a liquid grab sample from the SFP shall be taken arid analyzed for gamma isotopic activitiy in accordance with notation (f.) above before discharge flow is re-established when the release path is in the continuous mode to either the Auxiliary Service Water (ASW) for immediate release to the environment, or to TK-39 for future release.

Revision 16 2-4

2.2 Method to Calculate Off-Site Uquid Concentrations The basis for plant procedures that the plant operator requires to meet Control 2.1, I

which limits the total fraction of MPC (Fl) in liquid pathways at the point of discharge (see ODCM Figure 6-1) is discussed. (F.) is limited to less than or equal to one, i.e.,

1IZ C,

, MPCi Evaluation of (F.) is required concurrent with the sampling and analysis program I

in Table 2.1 of Control 2.1.

Determine the total fraction of MPG in liquid pathways as follows:

(F1) = E C, (Eq.2-1)

IMPC1 Where:

MPCI = Maximum permissible concentration of radionuclide "i" (IOCFR Part 20, Appendix B, Table 2, Column 2. See Appendix B of ODCM for listing).

I Revision 16 2-5 I

C 1=CTr IC 0 ite I

IoC = Concentration at the point of discharge of radionuclide i' from the test tank.

Csws = Concentration at the point of discharge of radionuclide "i from the Auxiliary Service Water System.

C0 tier = Concentration at the point of discharge of radionuclide wit from any other I

significant sources which may be created during plant decommissioning activities.

2.3 Method to Determine Radionuclide Concentration for Each Liquid Effluent Pathway 2.3.1 Test Tank Pathway CU'r is determined for each radionuclide above the analytical LLD from the activity in a proportional grab sample of the test tank and the predicted flow at the point of discharge.

Most periodic batch releases are from the two 5000-gallon capacity test tanks. When test tanks are filled with liquid waste, they are isolated for sampling and release. The volume of the tank's contents are determined from the liquid level in each tank. A chemist extracts a sample for radionuclide analysis. Aliquots of the sample proportional to the volume of the tank's contents are composited for appropriate radionuclide analyses. The composites contain suitable acids, alkalis, or carriers to assure the composite is representative of the sample.

Composite samples are analyzed at a minimum for tritium and gross alpha activity. At a Revision 16 I

2-6

minimum, each test tank batch is analyzed for principal gamma emitters.

2.3.2 Auxiliary Service Water System Pathway CP1 is determined for each radionuclide above the analytical LLD from the activity in composite samples from the effluent lines of the Auxiliary Service Water System downstream of any potential inleakage source.

2.3.3 Remaining Pathways ClVW is determined for each of the remaining pathways as follows:

a. Miscellaneous batch releases of potentially contaminated water, i.e., rain water collected in the containment liner of the inservice radioactive waste 20K tank, are analyzed to environmental detection levels and treated like a Test Tank according to Section 2.3.1 if plant related radioactivity is detected.

b. Construction Dewatering: The dismantling of buildings and related structures, including foundation excavations, may fill with either ground water or storm water.

The water-filled excavations, in many cases, must be dewatered to complete the dismantling activity. Construction dewatering may also include water generated during the process of digging new ground water monitoring wells, or other dismantlementldemolition related water and waste water sources. This discharge will be directed to either the existing storm drain systems or process treatment flow path with final release to Sherman Pond or the Deerfield River just below the Sherman Dam. With respect to effluent control, dewatering will be sampled and analyzed to determine the radionuclide content as detailed in Table 2.1. Releases to the environment without treatment will occur only if the projected impact is less than ODCM Control 6.1 dose limits. If higher activity water is found, it will be treated as appropriate (see Figure 6.1) prior to release to reduce the radionuclide inventory (other than tritium).

c. Spent Fuel Pool Draining: After all spent fuel and other contaminated materials are transferred to the Independent Spent Fuel Storage Installation (ISFSI), the Spent Fuel Pool (SFP) will be drained before dismantlement of this facility. Discharge of the SFP water in either batch or continuous mode will be treated by demineralization and filtration before release to Outfall 001. Figure 6.1 indicates the flow paths and in-line radiation monitoring prior to release.

Revision 16 2-7

3.0 DOSE/DOSE RATE CONTROLS AND CALCULATIONS 3.1 Dose Due to Radioactive Liquid Effluents Control 3.1 In accordance with Yankee Decommissioning Quality Assurance Program (Appendix D, Sectin B.4.a), the dose or dose commitment to a MEMBER OF THE PUBLIC from

• radioactive materials in liquid effluents released from the site (see Figure 1-2) to available uptake pathways shall be limited:

a. During any calendar quarter: less than or equal to 1.5 mrem to the total body and less than or equal to 5 mrem to any organ, and

b. During any calendar year: less than or equal to 3 mrem to the total body and less than or equal to 10 mrem to any organ.

Applicability At all times.

ACTION

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, and if not applicable to 10CFR Part 50.73, prepare and submit to the Commission within 30 days, pursuant to Control 7.4, a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be within the above limits.

Surveillance Reguirement SR 3.1 Dose Calculations - Cumulative dose contributions from liquid effluents shall be determined in accordance with the ODCM at least once per 31 days.

Bases Control 3.1 is provided to implement the requirements of Sections Il.A, lll.A, and IV.A of Appendix I, 10CFR Part 50. The control Implements the guides set forth in Section II.A. The Revision 15 3-1

-ACTION statements provide the required operating 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 materials in liquid effluents will be kept as low as Is reasonably achievable. The surveillance requirement implements the requirements in Section Il.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. Existing pathways of liquid exposure to MEMBER(S) OF THE PUBLIC which form the basis for calculating liquid doses In the ODCM are described in detail in Yankee Atomic Electric Company's design report, uSupplemental Information for the Purpose of Evaluation of 10CFR Part 50, Appendix I",dated June 2, 1976 (with amendments). The point of exposure from existing pathways for dose calculational purposes is taken downstream of Sherman Dam in the Deerfield River. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents were developed from 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 10CFR Part 50, Appendix I,' Revision 1, October 1977, and Regulatory Guide 1.1 13, Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I,' April 1977. Also, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in finished drinking water that are in excess of the requirements of 40CFR1 41.

No drinking water supplies from the Deerfield River below the plant have been identified.

Revision 15 3-2.

3.2 Total Dose Due to Radioactive Liquid and Gaseous Effluents Control 3.2 In accordance with Yankee Decommissioning Quality Assurance Program (Appendix D, Section B.4.a), the dose or dose commitment to any real MEMBER OF THE PUBLIC from all station sources is limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which is limited to less than or equal to 75 mrem) over a calendar year.

ADplicabilitv At all times.

ACTION

a. With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Controls 3.1.a, 3.1.b, 3.4.a, 3.4.b, 3.5.a, or 3.5.b, calculations should be made including direct radiation contributions from the reactor and from outside storage tanks to determine whether the above limits of Control 3.2 have been exceeded. Ifsuch is the case, and if not applicable to 10CFR Part 50.73, prepare and submit to the Commission within 30 days, pursuant to Control 7.4, 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. The Special Report shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from station sources, Including all effluent pathways and direct radiation, for the calendar year that Includes the release(s) covered by the report. It also shall 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 ifthe release condition resulting in violation of 400FR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40CFR190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

Revision 15 3-3

Surveillance Reauirement SR 3.2 Dose Calculations - Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with SR 3.1, 3.4, and 3.5 and in accordance with the ODCM.

Bases Control 3.2 is provided to meet the dose limitations of 40CFR Part 190 that have been incorporated into 10CFR Part 20 by 46FR18525. The control requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. 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 40CFR Part 190 if the individual reactors remain within the reporting requirement level.

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 40CFR Part 190 limits. For the purposes of the Special Report, It may be assumed that the dose commitment to a MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40CFR Part 190 have not already been corrected), in accordance with the provisions of 40CFR Part 190.1 1, is considered to be a timely request and fulfills the I requirements of 40CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40CFR Part 190 and does not apply in any way to the other requirements for dose limitation of 10CFR Part 20, as addressed in liquid and gaseous effluent controls. I Revision 11 34

3.3 Dose Rate Due to Radioactive Gaseous Effluents Control 3.3 In accordance with Yankee Decommissioning Quality Assurance Program (Appendix D, Section B.4.a), the dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Figure 1-1) shall be limited to the following:

a. Deleted

b. For radionuclides in particulate form with half-lives greater than 8 days: less than or equal to 1,500 mremlyr to any organ.

Applicability At all times.

ACTION With the dose rate(s) exceeding the above limits, without delay, take actions to decrease the release rate to within the above limit(s).

Surveillance Requirements SR 3.3.1 Deleted SR 3.3.2 The dose rate due to radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 3.1.

Bases With the removal of nuclear fuel from the Spent Fuel Pool (SFP) and its placement in sealed dry storage casks on the Independent Spent Fuel Storage Installation (ISFSI), along with the disassembly of all components of the waste gas treatment and holdup system, the potential for noble gas releases from the plant have been removed. Consequently, the need for control measures to limit noble gas releases and resulting dose rates to less than or equal to 500 mrem/yr whole body and less than or equal to 3,000 mrem/yr to the skin have been eliminated at this stage of the decommissioning process.

Revision 17 3-5

I Control 3.3 restricts, at all times, comparable with the length of the sampling periods of Table 3.1 the corresponding maximum organ dose rate above background to 1500 mrem/year for the highest impacted member of the public off-site from radionuclides In particulate form.

I This dose rate limit allows for operational flexibility when off normal occurrences may temporarily Increase gaseous effluent release rates (particulates) from the plant, while still providing controls to ensure that licensee meets the dose objectives of Appendix I to 10CFR50.

Revision 17 3-6 I

TABLE 3.1 Radioactive Gaseous Waste Samolina and Analvsis Proaram Sampling Minimum Analysis LLD Gaseous Release Type Frequency Frequency Type of Activity Analysis laCi/ml~a)

A. Plant Vent (Primary Deleted Vent Stack)

Deleted Deleted .

Deleted B. Building Demolition Notation b. Notation b. Notation b. Notation b.

Revision 17 3-7 I

I

• TABLE 3.1 (Continued)

Table Notation

a. The LLD is defined in Table Notation (a)of Table 4.1 of Control 4.1.

b. Prior to release of structures for demolition, surveys shall be conducted to confirm that residual radioactivity within the structure does not cause the average dose rates at 1 meter from the structure to exceed 50 mR/hr. Work Zone loose surface contamination levels on all structural surfaces released for demolition shall exhibit average removable residual alpha contamination <1000 dpm/1 00 cm2, or the minimum ,3,y to a ratio of Z 1000: 1 from samples determined to have residual alpha contamination > 1000 dpm/100 cm2 . (Alpha measurements are only required in plant areas of known or suspected alpha contamination). The average values shall be established by taking the mean of all the samples in a given building, or any subdivision thereof, with the MDA value used for all samples that are less than MDA.

If the above criteria are exceeded, the area shall be remediated to below the criteria -

prior to demolition, or engineering controls (e.g. temporary tenting of the work area with HEPA filtration of exhaust air and effluent monitoring) to mitigate and / or monitor potential airborne effluent releases, shall be utilized along with site boundary dose assessments to demonstrate that the dose criteria of ODCM Control 3.5. are met.

[Conversion of building residual radioactivity to estimated airborne release fractions of particulate radioactivity, if required, should be in accordance with Reference I (Also see Table 4-4, note a)]

Revision 17 3-8

3.4 Dose Due to Noble Gases Released in Radioactive Gaseous Effluents

• Control 3.4 Deleted Bases I

With the removal of nuclear fuel from the Spent Fuel Pool (SFP) and its placement in sealed dry storage casks on the Independent Spent Fuel Storage Installation (ISFSI), along with the shutdown and dismantlement of all components of the waste gas treatment and holdup system and drain down of all water from the SFP, the potential for noble gas releases from the plant have been removed. Consequently, the need for control measures to limit noble gas air doses to less than or equal to 5 mrad per calendar quarter (10 mrad /year) for gamma radiation and less than or equal to 10 mrad per calendar quarter (20 mrad/year) for beta radiation are no longer required to demonstrate the requirements of Section 11.8, lll.A, and IV.A of Appendix I, 10CFR Part 50.

Revision 17 I

3-9

Page Intentionally Left Blank Revision 17 l 3-10

3.5 Dose Due to Radionuclides in Particulate Form With Half-Lives Greater than Eight Days Control 3.5 In accordance with Yankee Decommissioning Quality Assurance Program (Appendix D, Section B.4.a), the dose to a MEMBER OF THE PUBLIC from radionuclides in I particulate form with half-lives greater than 8 days in gaseous effluents released from the site to areas at and beyond the SITE BOUNDARY (see Figure 1-1) shall be limited to the following:

a. During any calendar quarter less than or equal to 7.5 mrem to any organ, and

b. During any calendar year: less than or equal to 15 mrem to any organ.

Anlplicabilitv At all times.

ACTION

a. With the calculated dose from the release of radioactive materials in particulate form, in gaseous effluents exceeding any of the above limits, and if not applicable to 10CFR Part 50.73, prepare and submit to the Commission within 30 days, pursuant to Control 7.4, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be within the above limits.

Surveillance Requirement SR 3.5 Dose Calculations - Cumulative dose contributions for the current calendar quarter and current calendar year for radionuclides in particulate form with half-lives greater than 8 days I shall be determined in accordance with the ODCM at least once every 31 days.

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Bases Control 3.5 is provided to implement the requirements of Sections 11.C, 1llA, and IV.A of Appendix I, 10CFR Part 50. The control is the guide set forth in Section II.C. 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 to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as Is reasonably achievable." The surveillance requirement implements the requirements In Section Ill.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 equations specified in the ODCM for calculating the doses due to the actual release rates of the subject materials were developed using 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 10CFR Part 50, Appendix l," Revision 1, October 1977, and Regulatory Guide 1.1 11, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Ught-Water Cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for radionuclides in particulate form with half-lives greater than eight days are dependent on the existing radionuclide pathways to man in areas at and beyond the SITE BOUNDARY. The pathways which were examined in the development of these specifications were: (1) individual inhalation of airborne radionuclides, (2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition onto grassy areas where milk and meat animals graze with consumption of the milk and meat by man, and (4)deposition on the ground with subsequent exposure of man.

Revision 17 3-12

3.6 Dose Calculation Concepts The term 'dose' for Ingested or inhaled radioactivity means the dose commitment, measured in mrem, 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 refer to the fifty-year dose commitment from one year's worth of releases. 'Dose In a quarter similarly means a fifty-year dose commitment from one quarter's releases. The term "dose,"

with respect to external exposures, such as to ground phase deposition, 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.

The dose calculated by 'Method Ioequations is not necessarily the actual dose received by a real individual, but usually provides an upper bound for a given release because of the conservative margin built into the dose factors and the selection and definition of the critical receptors. The radioisotope specific dose factors in each "Method 1s dose equation represent the greatest dose to any organ of any age group accounting for existing or potential pathways of exposure. The critical receptor assumed by 'Method I" equations Is typically a hypothetical individual whose behavior in terms of location and intake results in a dose which is expected to be higher than any real individual. Method II allows for a more exact dose calculation for real Individuals if necessary by considering only existing pathways of exposure with the recorded release.

Revision 17 3-13

3.7 Method to Calculate the Total Body Dose from Liquid Releases Control 3.1 limits the total body dose commitment to a MEMBER OF THE PUBLIC from radioactive material in liquid effluents to 1.5 mrem per quarter and 3 mrem per year. Control 6.1 requires liquid radioactive waste treatment when the total body dose estimate exceeds 0.06 mrem in any 31-day period. Control 3.2 limits the total body dose commitment to any real MEMBER OF THE PUBLIC from all station sources (including liquids) to 25 mrem In a year.

Dose evaluation is required at least once per 31 days. Ifthe liquid radioactive waste treatment system is not being used, dose evaluation is required before each release.

To evaluate total body dose for Control 6.1 add the total body dose from today's expected releases to the total body dose accumulated for the time period of interest.

3.7.1 Method I The total body dose from a liquid release is:

Dtb = KEQ1 DFL1 ,t (Eq. 3-1)

(mrem) I Where:

DFLitA = Site-specific total body dose factor (mrem/Ci) for liquid release. See Table 1.7.

Q1 = Total activity (Curies) released to liquids of radionuclide 'it during the period of interest. For i = Fe-55, Sr-90, or H-3, use the best estimates (such as the most recent measurements).

K = 366/Fd; where Fd is the average (typically monthly average) dilution flow of the Deerfield River below Sherman Dam (in ft3 /sec). If Fd cannot be obtained or Fd is greater than 366, K can be assumed to equal 1.0. The value, 366, is the ten-year minimum monthly average Deerfield River flow rate below Sherman Dam (in ft3 /sec).

Revision 16 3-14

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

a. Uquid releases to the auxiliary service water pathway to Sherman Pond or Deerfield River just below the Sherman Power House Dam (Outfalls 001, 003 and 004 as specified in Reference K).

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

3.7.2 Method II If Method I cannot be applied or if the Method I dose exceeds the limit or if a more exact calculation is required, then Method I!should be applied. Method II consists of the models, Input data, and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific models, data, or assumptions are more applicable. The base case analysis is a good example of the use of Method II. It Is an acceptable starting point for a Method IIanalysis.

3.7.3 Basis for Method I Method I may be used to show that the controls which limit off-site total body dose from liquids (Controls 3.1, 3.2, and 6.1) have been met for releases over the appropriate periods.

These requirements are based on design objectives and standards in 10CFR Part 50 and 40CFR Part 190. Control 3.1 is based on the ALARA design controls in 10CFR Part 50, Appendix I, Subsection IIA. Control 6.1 is an appropriate fraction', determined by the NRC, of the ALARA design control. Control 3.2 is based on Environmental Standards for the Uranium Fuel Cycle in 40CFR Part 190 which applies to direct radiation as well as liquid and gaseous effluents. Method I applies only to the liquid contribution.

Method I was developed such that "the actual exposure of an individual ... is unlikely to be substantially underestimated (10CFR Part 50, Appendix I). The definition of a single 'critical receptors (a hypothetical Individual whose behavior results in an unrealistically high dose) provides part of the conservative margin to the calculation of total body dose in Method I.

Method IIallows that actual Individuals with real behaviors be taken into account for any given release. In fact, Method I was based -on a Method IIanalysis for the critical receptor and annual average conditions instead of any real individual. The analysis was called the Obase case"; it Revision 16 3-15

was then reduced to form Method I. The base case, the method of reduction, and thb assumptions and data used are presented.

The steps performed in the Method I derivation follow. First, in the base case, the dose impact to the critical receptor (in the form of dose factors in mremfCi) for a one Curie release of each radionuclide in liquid effluents was derived. The base case analysis uses the methods, data, and assumptions in Regulatory Guide 1.109 (Equations A-3, A-7, A-13, and A-1 6, Reference A). Tables 3.2 and 3.3 outline human consumption and environmental parameters used in the analysis. It is assumed that the critical receptor fishes below Sherman Dam and eats the fish caught from this location and consumes leafy vegetables and produce from a farm which is irrigated with water from the Deerfield River below Sherman Dam. It also is assumed that the critical receptor drinks milk and eats meat from cows who drink water from the Deerfield River below Sherman Dam and eat silage from the Irrigated farm above.

For any liquid release during any period, the increment in annual average total body dose from radionuclide WI'is:

ADftb = (Q.) (DFL*)

where DFL- Is the total body dose factor for radionuclide "i", and Qj is the activity of radionuclide "ig released in Curies.

Method I is more conservative than Method II because it is based on the following reduction of the base case. The dose factors, DFLib, used in Method I were chosen from the base case to be the highest of the four age groups for that radionuclide. In effect, each radionuclide is conservatively represented by its own critical age group.

Revision 11 3-16

TABLE 3.2 Environmental Parameters for Liquid Effluents at Yankee Rowe (Derived from Reference A)

Food Grown with Contaminated Water Aquatic Shoreline Leafy Variable Food Activity Vegetables Veg. Meat Cow Milk MP Mixing Ratio( 0.84 0.84 0.84 0.84 0.84 0.84 TP Transit Time (hrs) 24.00 0.00 0.00 0.00 480.00 48.00 YV Agricultural Productivity (kg/m 2) 2.00 2.00 2.00 2.00 P Soil Surface Density (kg/m2) - 240.00 240.00 240.00 240.00 IRR Irrigation Rate (l/m2/hr) . 0.15 0.15 0.15. 0.15 TE Crop Exposure Time (hrs) . 1440.00 1440.00 1440.00 1440.00 TH Holdup Time (hrs) . 1440.00 24.00 2160.00 2160.00 QAW Water Uptake Rate for (Vd) . - 50.00 60.00 Animal QF Feed Uptake Rate (kg/d) . . . 50.00 50.00 Location of Critical Individual Below Below Below Below Below Below Sherman Sherman Sherman Sherman Sherman Sherman Dam Dam Dam Dam Dam Dam (1) Listed mixing ratios apply to Method I dose factors. Method II analyses can apply calculated mixing ratios based on river flow and plant discharge dilution flow which exist over the period of actual release.

Revision 11 3-17

TABLE 3.3 Age-Snecific Usage Factors for Various lUquid Pathways at Yankee Rowe (From Reference A, Table E-5. Zero where no pathway exists)

Leafy Potable Age Veg. Veg. Milk Meat Fish Invert. Water Shoreline Group (kg/yr) (kfyr) (Vyr) (kg/yr) (kglhr) (kg/yr) (l/yr) (hr/yr)

Adult 520.00 64.00 310.00 110.00 21.00 0.00 0.00 12.00 Teen 630.00 42.00 400.00 65.00 16.00 0.00 0.00 67.00 Child 520.00 26.00 330.00 41.00 6.90 0.00 0.00 14.00 Infant 0.00 0.00 330.00 0.00 0.00 0.00 0.00 0.00 Revision 11 3-180

3.8 Method to Calculate Maximum Organ Dose from Liquid Releases Control 3.1 limits the maximum organ dose commitment to a MEMBER OF THE PUBLIC from radioactive material in liquid effluents to 5 mrem per quarter and 10 mrem per year.

Control 6.1 requires liquid radioactive waste treatment when the maximum organ dose estimate exceeds 0.2 mrem in any 31-day period. Control 3.2 limits the maximum organ dose.

commitment to any real MEMBER OF THE PUBLIC from all station sources (including liquids) to 25 mrem in a year except for the thyroid, which is limited to 75 mrem in a year. Dose evaluation Is required at least once per 31 days. If the Liquid Radioactive Waste Treatment System Is not being used, dose evaluation is required before each release.

To evaluate the maximum organ dose for Control 6.1, add the organ dose from the expected releases to the organ dose accumulated for the time period of interest.

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

Doman = KEQODFL 1,r,0 (Eq. 3-2)

(mrnm) I Where:

DFLimo = Site-specific maximum organ dose factor (mrem/Ci) for a liquid release. See Table 1.7.

Qi = Total activity (Curies) released to liquids of radionuclide 'i" during the period of interest. For i = Fe-55, Sr-90, or H-3, use the best estimates (such as the most recent measurements).

K = 366IFd; where Ed is the average (typically monthly average) dilution flow of the Deerfield River below Sherman Dam (in ft3/sec). If Fd cannot be obtained or Fd is greater than 366, K can be assumed to equal 1.0. The value, 366, Is Revision 16 3-19

the ten-year minimum rrionthly average Deerfield River flow rate below Sherman Dam (in ft3 /sec).

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

a. Liquid releases to Sherman Pond or Deerfield River just below the Sherman Power House Dam, as permitted by Reference K.

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

3.8.2 Method II If Method I cannot be applied, or if the Method I dose exceeds the limit, or if a more exact calculation is required, then Method II should be applied. Method II consists of the models, input data, and assumptions In Regulatory Guide 1.109, Revision 1 (Reference a),

except where site-specific models, data, or assumptions are more applicable. The base case analysis is a good example of the use of Method II. It is an acceptable starting point for a Method II analysis.

3.8.3 Basis for Method I The methods to calculate the maximum organ dose parallel the total body dose methods (see Section 3.7.3). Only the differences are presented here.

For any liquid release during any period, the increment in annual average dose from radionuclide i to the maximum organ is:

ADi. = (Qi) (DFL&")

where DFLr,1, is the maximum organ dose factor for radionuclide wi", and Qi is the activity of radionuclide i released in Curies.

The dose factors, DFLI., used in Method I were chosen from the base case to be the highest of the set of seven organs and four age groups for each radionuclide. This means that the maximum effect of each radionuclide is conservatively represented by its own critical age group and critical organ.

Revision 16 3-20

3.9 Method to Calculate the Total Body Dose Rate from Noble Gases Section Deleted Revision 17 I

3-21

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3.10 Method to Calculate the Skin Dose Rate from Noble Gases Section Deleted I

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3-26

3.11 Method to Calculate the Critical Organ Dose Rate from Particulates with Half--Uves I

Greater Than Eight Days Control 3.3 limits the dose rate at any time at location at or beyond the Site Boundary from radionuclides in particulate form with half-lives greater than eight days to 1,500 mrem/year to any organ. The peak release rate averaging time in the case of particulates is commensurate with the time the particulate samplers are in service between changeouts.

3.11.1 Method I The critical organ dose rate can be determined as follows:

DM =Z Qf, DFG',

(mrem (Eq. 3-5)

C (r yr fh(iCi

) ksec

- sec pCi- yr Where:

QR = Estimated airborne particulate radionuclide "i"released to the atmosphere during building (or structure) demolition averaged over the time duration of actual field demolition work, in uCisec DFG% = Site-specific critical organ dose rate factor (mrm -sec ) for a gaseous release. See Table 1.8.

I Revision 17 I

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3.11.2 Method II If Method I cannot be applied, or if the Method I dose exceeds the control limit, or if a more exact calculation is required, then Method II may be applied. Method II consists of the models, input data, and assumptions in Regulatory Guide 1.109, Revision 1 (Reference A),

except where site-specific models, data, or assumptions are more applicable l 3.11.3 Basis for Method I I

The equation for Dco is derived by modifying Equation 3-8 from Section 3.14 as follows:

A\ V" ,^

uCO =L iVUI tii Cmrem )~ (Ci '(mrem )

)~r)I(

yr Ci)

(Eq. 3-8)

Applying the conversion factor, 31.54 (Ci-secljiCi-yr), and converting Q to QR in piCVsec yields:

DCO = 31.54 E Qx, DFGic Cmrem A(Ci-sec (pCi )mrem) tyr pCi-yr )7sec Ci )

Revision 17 3-28 I

Equation 3-5 is rewritten in the form:

Dor= Q1DFGk o DFG Where:

DFG' = (DFGKO.) (31.54)

(mrem -sec = (mrem ) Cl-sec tLCI- r Ci - )iLCi-yr)

The dose conversion factor, DFGEO, has been developed assuming a unit activity (by radionuclide) ground level release with historical site meteorological information and assuming that all exposure pathways exist at the site boundary with the most limiting atmospheric dispersion factors. These dose factors are used to determine accumulated doses over extended periods and have been calculated with a Shielding Factor (SF) for the ground plane exposure set equal to 0.7, as referenced in Regulatory Guide 1.109 (Revision 1). In the case of the dose rate conversion factor, DFG's,, the same development process was followed, but the SF was set equal to 1.0 since no credit for residential shielding is taken when the exposure rate is not averaged over an extended period of time. In addition, the soil exposure time (th) to any effluent release plume is changed from the Regulatory Guide default value of 131,400 h r (15 years) to a conservative 26,280 hours0.00324 days <br />0.0778 hours <br />4.62963e-4 weeks <br />1.0654e-4 months <br /> (3 years) estimate of the time remaining to complete the site remediation process. The 15-year duration was intended to estimate the midpoint of plant operating life In order to approximate buildup in the environment. The time to complete the dismantlement of potentially contaminated plant structures is far short of this default assumption.

Should Method II be needed, the analysis for critical receptor critical pathway(s) may be performed with latest land use census data to identify the location of those pathways which are most impacted by these types of releases.

Revision 17 3-29

3.12 Method to Calculate the Gamma Air Dose from Noble Gases (Kr-85)

Section Deleted I

Revision 17 3-30 I

3.13 Method to Calculate the Befa Air Dose from Noble Gases Section Deleted I

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3-31

3.14 Method to Calculate the Critical Organ Dose from Particulates Control 3.5 limits the critical organ dose to a MEMBER OF THE PUBLIC from I

radioactive particulates with half-lives greater than eight days In gaseous effluents to 7.5 mrem per quarter and 15 mrem per year. Control 3.2 limits the total body and organ dose to any real MEMBER OF THE PUBLIC from all station sources (including gaseous effluents) to 25 mrem in a year except for the thyroid, which is limited to 75 mrem in a year.

3.14.1 Method I The critical organ dose from a gaseous release is:

DC0 0j1DFGWc (Eq. 3-8)

(Mrmrn I Where:

0j = Estimated airborne particulate radionuclide "i" released to the atmosphere during building (or structure) demolition (Ci).

DFGi O= Site-specific critical organ dose factor (mrem/Ci) for a gaseous release. See Table 1.8.

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3.14.2 Method II Method II consists of the models, input data, and assumptions in Regulatory Guide 1.109, Revision 1 (Reference A), except where site-specific models, data, or assumptions are more applicable. The base case analysis, documented below, is a good example of the use of Method II. It is an acceptable starting point for a Method II analysis.

3.14.3 Basis for Method I Method I may be used to show that Control 3.5, which limit off-site organ dose from gases, have been met for releases over the appropriate periods. Control 3.5 is based on the ALARA requirements in 10CFR Part 50, Appendix I, Subsection II C. The methods in this section also provide one component of the inputs required by Control 3.2. Control 3.2 is based on Environmental Standards for Uranium Fuel Cycle In40CFR1 90 which applies to direct radiation as well as to liquid and gaseous effluents. The methods apply only to the particulates in gaseous effluents component of effluents.

Method I was developed such that '... the actual exposure of an individual ... is unlikely to be substantially underestimated' (10CFR Part 50, Appendix I). The use 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 with real behaviors and exposure pathways be taken into account for any given release. In fact, Method I was based on a Method Il analysis of the critical receptor for the annual average conditions with all realistically potential pathways (per Regulatory Guide 1.109, Revision 1) assumed.

The steps performed in the Method I derivation follow. First, the dose impact to the critical receptor in the form of dose factors, DFGj;o (mremlCi), for a one Curie release of each particulate radionuclide released to gaseous effluents was derived using simplifying and conservative assumptions. The base case analysis uses the methods, data, and assumptions in Regulatory Guide 1.109 (Equations C-2, 0-4, and C-13 in Reference a). Tables 3.4 and 3.5 outline human consumption and Revision 17 3-33

environmental parameters used in the analysis. It Is conservatively assumed that the critical receptor lives at the 'maximum SITE BOUNDARY dilution factor location' as defined in Section 3.1 5.

For airborne releases during any period, the dose from radionuclide 'i' is:

I Djw = (DFGLco) (Qu) where DFGjC is the critical dose factor for radionuclide "i",and Qj is the activity of radionuclide

'i' released in Curies.

Revision 17 3-340 I

TABLE 3.4 Aae-Soecific Usaae Factors (from Regulatory Guide 1.109, Table E-5)

Leafy Age Vegetables Vegetables Milk Meat Inhalation Group (kglyr) (kgfyr) (Vyr) (kg/yr) (m3/yr)

Adult 520.00 64.00 310.00 110.00 8,000.00 Teen 630.00 42.00 400.00 65.00 8,000.00 Child 520.00 26.00 330.00 41.00 3,700.00 Infant 0.00 0.00 330.00 0.00 1,400.00 Revision 17 I

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TABLE 3.5 Environmental Parameters for Gaseous Effluents at the Yankee Plant (Derived from Reference a)

Vegetables Cow Milk Goat Milk* Meat Variable Stored Leafy Pasture Stored Pasture Stored Pasture Stored YV Agricultural (kg/m 2) 2.00 2.00 0.70 2.00 0.70 2.00 0.70 2.00 Productivity P Soil Surface (kg/m 2) 240.00 240.00 240.00 240.00 240.00 240.00 240.00 240.00 Density T Transport Time to (hrs) - - 48.00 48.00 48.00 48.00 480.00 480.00 User TB Soil Exposure (hrs) 26280 26280 26280 26280 26280 26280 26280 26280 720.00 1440.00 I

TF Crop Exposure (hrs) 1440.00 .1440.00 720.00 1440.00 720.00 1440.00 Time to Plume TH Holdup After (hrs) 1440.00 24.00 0.00 2160.00 0.00 2160.00 0.00 2160.00 Harvest QF Animals Daily (kg/day) . . 50.00 50.00 6.00 6.00 50.00 50.00 Feed FP Fraction of Year . . 0.50 . 0.50 .0.50 -

on Pasture(2)

FS Fraction Pasture 1.00 1.00 - 1.00 When on Pasture(3)

FG Fraction of Stored 0.76 . .

Veg. Grown in Garden Revision 17 I

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TABLE 3.5 (Continued)

Environmental Parameters for Gaseous Effluents at the Yankee Plant (Derived from Reference a)

Vegetables Cow Milk Goat Milk* Meat Variable Stored Leafy Pasture Stored Pasture Stored Pasture Stored FL Fraction of Leafy 1.00 Veg. Grown in Garden Fl Fraction - . - . .

Elemental Iodine

=0.5 H Absolute(4) (gm/mr3) . . . . .

Humidity = 5.6 Pathway Is not included in Method I. It Is listed for informational purposes and the possible use in a Method It analysis.

Notes:

(1) For Method II dose/dose rate analyses of identified radioactivity releases of less than one year, the soil exposure time for that release may be set at 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> (1 year) for all pathways.

(2) For Method II dose/dose rate analyses performed for releases occurring during the first or fourth calendar quarters, the fraction of time animals are assumed to be on pasture is zero (nongrowing season). For the second and third calendar quarters, the fraction of time on pasture (FP) will be set at 1.0. FP may also be adjusted for specific farm locations if this Information Is so Identified and reported as part of the land use census.

(3) For Method II analyses, the fraction of pasture feed while on pasture may be set to less than 1.0 for specific farm locations if this information is so identified and reported as part of the land use census.

(4) For all Method II analyses, an absolute humidity value equal to 5.6 (gm/m3) shall be used to reflect conditions in the Northeast (

Reference:

Health Physics Journal, Vol. 39 (August), 1980; Page 318-320, Pergammon Press).

Revision 17 3-37.

3.15 Critical Receptors and Long-Term Average Atmospheric Dispersion Factors for Important Exposure Pathways The gaseous effluent dose equations (Method I) have been simplified by assuming an individual whose behavior and living habits inevitably lead to a higher dose than anyone else.

The following pathways of exposure to gaseous effluents as listed in Regulatory Guide 1.109 (Reference a) have been considered. They are:

a. Direct exposure to contaminated air,

b. Direct exposure to contaminated ground,

c. Inhalation,

d. Ingestion of vegetables, I

e. Ingestion of milk, and

f. Ingestion of meat.

I 3.15.1 Critical Receptors The most limiting SITE BOUNDARY location in which individuals are or could be located 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 SSE sector, 800 meters.

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3.15.2 Yankee Atmospheric Dispersion Model The annual average dispersion factors are computed for routine (long-term) releases using the Yankee Atomic Electric Company's (YAEC) AEOLUS (Reference b) computer code.

AEOLUS produces the following annual average dispersion factors for each location:

a. X/Q, nondepleted dispersion factors for evaluating ground level concentrations;

b. [XtQ]c, depleted dispersion factors for evaluating ground level concentrations of particulates;

c. DIQ, deposition factors for dry deposition of particulates. I The AEOLUS diffusion model Is described In the AEOLUS manual (Reference b).

AEOLUS is based, in part, on the straight-line airflow model as discussed in Regulatory Guide 1.1 11 (Reference C).

An additional consideration in the dispersion modeling relates to the location of the plant In a relatively narrow valley. Wind channelling Is assumed to occur in the seven sectors which make up the valley. The seven sectors are SSE, S. SSW, SW, WSW, W, and WNW. If a receptor location is in one of the valley sectors, the contributions from the other six valley sectors are averaged into the particular valley receptor. This is done for distances greater than I

500 meters from the center of the plant complex where the valley effects are assumed to cause channelling.

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3.15.3 Long-Term Average Dispersion Factors for Critical Receptors Actual measured meteorological data for the five-year period, January 1981 through December 1985, was analyzed to determine the locations of the maximum off-site atmospheric dispersion factors. Each dose and dose rate calculation incorporates the maximum applicable off-site, long-term average atmospheric dispersion factor. The values used and their locations are summarized InTable 3.6.

Revision 17 3

TABLE 3.6 Yankee Nuclear Power Station Five-Year Average Atmospheric Dispersion FactorsM)

DoselDose Rate to Individual Total Body Skin Critical Organ X/Q Depleted ( .s )2.39x10 5 Deleted D/Q (12 ) 5.24 x 104 Deleted

(')SSE SITE BOUNDARY, 800 meters from the location of the former primary vent stack.

Revision 17 3-41

3.16 Method to Calculate Direct Dose from Plant Operation Control 3.2 restricts the dose to the whole body and any organ of any real MEMBER OF THE PUBLIC at and beyond the Site Boundary from all station sources (including direct radiation) to the limit of 25 mrem in a year, except for the thyroid which is limited to 75 mrem in a year.

Estimates of direct exposure above background in areas at and beyond the site boundary (or in residential areas inside the site boundary) can be determined from measurements made by environmental TLDs that are part of the Environmental Monitoring Program (see Table 4.4). Alternatively, direct dose calculations from identified fixed sources on site can be used to estimate the off-site direct dose contribution where TLD Information may not be applicable.

Revision 17 3-42.

4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING 4.1 Monitoring Program Control 4.1 In accordance with Yankee Decommissioning Quality Assurance Program (Appendix D, Section B.4.b), the Radiological Environmental Monitoring Program shall be conducted as specified in Table 4.1.

Arplicabilitv At all times.

AClION

a. With the Radiological Environmental Monitoring Program not being conducted as specified in Table 4.1, prepare and submit to the Commission in the Annual Radiological Environmental Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, or to malfunction of automatic sampling equipment. If the latter, every effort shall be made to complete corrective action prior to the end of the next sampling period.

b. With the level of radioactivity as the result of plant effluents in an environmental sampling media at one or more of the locations specified in Table 4.1 exceeding the reporting levels of Table 4.2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days from the receipt of the laboratory analyses, pursuant to Control 7.4, a Special Report which includes an evaluation of any release conditions, environmental factors, or other aspects which caused the limits of Table 4.2 to be exceeded. When more than one of the radionuclides in Table 42 are detected in the sampling medium, this report shall be submitted H:

concentration (1) concentration (2) t1.0 reportinglevel (1) reportinglevel (2)

Revision 15 4-1

When radionuclides other than those InTable 4.2 are detected and are the result of plant effluents, this report shall be submitted If the potential annual dose to a MEMBER OF THE PUBLIC is equal or greater than the calendar year limits of Controls 3.1, 3.3, and 3.4. 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.

Surveillance Requirement SR 4.1 The radiological environmental monitoring samples shall be collected pursuant to Table 4.1 from the locations given in the ODCM and shall be analyzed pursuant to the requirements of Table 4.1 and the detection capabilities required by Table 4.3.

Bases The Radiological Environmental Monitoring Program required by Control 4.1 provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides, which lead to the highest potential radiation exposures of MEMBER(S) OF THE PUBLIC resulting from the station operation. The monitoring program implementsSection IV.B.2 of Appendix I, 10CFR Part 50, and thereby, supplements the Radiological Effluent Monitoring Program 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 modeling of the environmental exposure pathways. Guidance for the monitoring program is Revision 16 I

4-2

provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1, November 1979. Program changes may be initiated based on operational experience.

The detection capabilities required by Table 4.3 are considered optimum for routine environmental measurements in Industrial laboratories. It should be recognized that the LLD is defined as an a prior (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.

Revision 15 4-3.

TABLE 4.1 Radiological Environmental Monitorinq Program*

Exposure Pathway Number of Sample and/or Sample Locations Sampling and Collection Frequency Type and Frequency of Analysis

1. AIRBORNE

a. Particulates 5 Continuous operation of sampler with sample Gross beta radioactivity following- I collections as required by dust loading, but at filter change. Composite (by least once per two weeks. location) for gamma isotopic at I least once per quarter.

2. DIRECT RADIATION 24** Quarterly Gamma dose, at least once per I quarter.

3. WATERBORNE

a. Surface 2 Composite sample* collected over a period of Gross beta and gamma Isotopic one month at downstream location; monthly grab analysis of each sample. Tritium sample at upstream control location. analysis of composite sample at least once per quarter.

b. Ground 2 At least once per quarter. Gamma isotopic and tritium analyses of each sample.

c. Sediment from 1**** At least once per six months. Gamma Isotopic analysis of each Shoreline sample.

I Revision I1 4-4

TABLE 4.1 (Continued)

Radiological Environmental Monitoring Program*

Exposure Pathway Number of Sample and/or Sample Locations Sampling and Collection Frequency Type and Frequency of Analysis

4. INGESTION

a. Fish 2 Commercially and recreationally important Gamma isotopic analysis on species. Seasonal or semiannually, if not edible portions.

seasonal.

b. Food Products 3 At time of harvest. One sample of any of the Gamma isotopic analysis on I following classes of food products: edible portions.

1. Tuberous vegetable

2. Above ground vegetable

3. Fruit

• Specific sample locations for all media are specified in the ODCM and reported in the Annual Radiological Environmental OperatIng Report.

• • Composite samples shall be obtained by collecting an aliquot at Intervals not exceeding two hours.

• • • Does not include Restricted Area Fence locations, or those TLD's associated with the ISFSI pad monitoring.

• • • • One sample from downstream area with existing or potential recreational value.

Revision 16 4-5

.I

TABLE 4.2 Reporting Levels for Radioactivity Concentrations in Environmental Samples Reporting levels for nondrinking water pathways.

Revision 16 4-6 I.

TABLE 4.3 Detection Capabilities for Environmental Sample Analysis() (C)

Airborne Water Particulates Fish Food Products Sediment Analysis(d) (pCIA) (pCVm3) (pCI/kg. wet) (pCI/kg. wet) (pCI/kg. dry)

Gross beta 4 x 10+° I x 2 H-3 2 x 103 .

Line Deleted I Co-58, -60 1.5x 10+1 1.3x10+ 2 ._-

Line Deleted Line Deleted Cs-134 1.5 x lo+ 5x 10'2 1.3 x10+2 6 x 1lo+ 1.5 x 10+2 Cs-137 1.8x10+ 6x 10 2 1.5x10+ 2 8x1O+ 1 1.8x10 2 I.-, I?

95 -4i, f . 3~

II

' 0 e

,D §

'T

'411'Vu 10)

Revision 16 4-7 I

TABLE 4.3 (Continued)

Table Notation

a. The LLD is the smallest concentration of radioactive material in a sample that 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.

For a particular measurement system (which may include radiochemical separation):

LLD (4.66) (Sb)

(E)(V)(2.22)(Y)[Exp(-At)j Where:

LLD = A priori lower limit of detection as defined above (microcuries or picocuries per unit mass or volume).

Sb = Standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute).

E = Counting efficiency (counts per disintegration).

V = Sample size (units of mass or volume).

2.22 = Number of disintegrations per minute per picocurie.

Y = Fractional radiochemical yield (when applicable).

A = Radioactive decay constant for the particular radionuclide.

At = Elapsed time between sample collection and analysis.

Typical values of E, V, Y, and At can be used in the calculation. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include Revision 15 4-8.

the typical contributions of other radionuclides normally present in the samples (e.g., potassium-40 in milk samples).

Analysis shall be performed in such a manner that the stated IIDs will be achieved under routine conditions. Occasionally, background fluctuations, unavoidably small sample sizes, the presence of interfering radionuclides, or other uncontrollable circumstances may render these LLDs unavailable. In such cases,-the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.

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.

b. Parent only.

c. If the measured concentration minus the 5 sigma counting statistics is found to exceed the specified LLD, the sample does not have to be analyzed to meet the specified LLD.

d. This list does not mean that only these radionuclides are to be considered. Other peaks that are identifiable, together with those of the listed radionuclides, also shall be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to Control 7.1.

Revision 15 4-9.

4.2 Land Use Census Control 4.2 In accordance with Yankee Decommissioning Quality Assurance Program (Appendix D, Section B.4.b), a land use census shall be conducted to identify the location of the nearest milk animal, the nearest residence, and the nearest garden* of greater than 500 square feet producing fresh leafy vegetables in each of the 16 meteorological sectors within a distance of five miles.

Apnlicabilitv At all times.

ACTION

a. With a land use census identifying a location(s) which yields at least a 20 percent greater dose or dose commitment than the values currently being calculated in SR 3.5, identify the new location(s) in the next Annual Radioactive Effluent Release Report.

b. With a land use census identifying a location(s) which yields a calculated dose or dose commitment (via the same exposure pathway) at least 20 percent greater than at a location from which samples are currently being obtained in accordance with Control 4.1, add the new location(s) to the Radiological Environmental Monitoring Program within 30 days 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 (via the same exposure pathway) may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted. Identify the new location(s) in the next Annual Radioactive Effluent Release Report.

Surveillance Requirement SR 4.2 The land use census shall be conducted at least once per 12 months between the dates of June 1 and October 1 by either a door-to-door survey, aerial survey, or by consulting local agriculture authorities.

• Inlieu of the garden census, broad leaf vegetation sampling may be performed at the site boundary in the direction sector with the highest DIQ.

Revision 15 4-10

• The results of the land use census shall be included in the Annual Radiological Environmental Operating Report pursuant to Control 7.1.

Bases Control 4.2 is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARIES are identified and that modifications to the monitoring program are made if required by the results of the land use census. The census satisfies the requirements of Section IV.B.3 of Appendix I, 10CFR Part 50. Restricting the census to gardens of greater than 500 square feet provides 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 used: (1) 20 percent of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage) and (2) a vegetation yield of 2 kg/square meter. In lieu of the garden census, broad leaf vegetation samples from the SITE BOUNDARY in the direction sector with the highest DIQ may be substituted. The use of the maximum off-site D/Q value predicted for gaseous effluents from the plant stack sitewill generate the maximum possible calculated dose, and thus, no real garden located at any other point could have a greater calculated dose or dose commitment:

The addition of new sampling locations to Control 4.1, based on the land use census, is limited to those locations which yield a calculated dose or dose commitment 20 percent greater than the calculated dose or dose commitment at any location currently being sampled. This eliminates the unnecessary changing of the Environmental Radiation Monitoring Program for new locations which, within the accuracy of the calculation, contribute essentially the same to the dose or dose commitment as the location already sampled. The substitution of a new sampling point for one already sampled when the calculated difference in dose is less than 20 percent, would not be expected to result in a significant increase in the ability to detect plant effluent-related radionuclides.

Revision 17 4

4.3 Intercomparison Program Control 4.3 In accordance with Yankee Decommissioning Quality Assurance Program (Appendix D, Section B.4.b), analyses shall be performed on referenced radioactive materials supplied as part of the quality assurance Laboratory Intercomparison Program.

Applicabilitv At all times.

ACTION 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.

Surveillance Requirement SR 4.3 A summary of the results of analyses performed as part of the above required Intercomparison Program shall be included in the Annual Radiological Environmental Operating Report.

Bases The control for participation in the Intercomparison 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. The independent checks are completed as part of a quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid for the purposes of Section IV.B.2 of Appendix I, 10CFR Part 50.

4.4 Environmental Monitoring Locations The radiological environmental monitoring stations are listed in Table 4.4. The locations of these stations with respect to the Yankee plant facility are shown on the maps in Figures 4-1 through 4-7.

Revision 15 4-12

TABLE 4.4 Radioloaical Environmental Monitorina Stations*

Exposure Pathway Sample Location and Distance From the Direction From and/or Sample Designated Code** Plant (km) Plant

1. AIRBORNE (Particulates)(a) AP-1 1 Observation Stand 0.5 NW I AP-12 Monroe Bridge 1.1 SW AP-13 Rowe School 4.2 SE AP-14 Harriman Power Station 3.2 N AP-21 Williamstown, MA 22.2 W

2. WATERBORNE

a. Surface WR-11 Bear Swamp Lower 6.3 Downriver Reservoir WR-21 Harriman Reservoir 10.1 Upriver

b. Ground WG-1 1 Plant Potable On-Site Well WG-12 Sherman Spring 0.2 NW

c. Sediment SE-1I Number 4 Station 36.2 Downriver From SE-21 Harriman Reservoir 10.1 Upriver Shoreline

3. INGESTION Line Deleted

a. Fish FH-1 1 Sherman Pond 1.5 At Discharge and Point Invertebrates FH-21 Harriman Reservoir 10.1 Upriver

b. Food TF-1 1 Monroe Bridge 1.9 WSW Products TF-13 Monroe, MA 1.9 WNW TF-21 Williamstown, MA 21.0 WSW Revision 17 I

4-13

TABLE 4.4

• (Continued)

Radiological Environmental Monitoring Stations*

Exposure Pathway Sample Location and Designated Distance From Direction and/or Sample Code* the Plant (km) I From Plant

4. DIRECT GM-1 Furlon House 0.80 SW RADIATION GM-2 Observation Stand 0.50 NW GM-3 Rowe School 4.20 SE I GM-4 Harriman Station 3.20 N GM-5 Monroe Bridge 1.10 SW GM-6 Readsboro Road Barrier 1.30 N GM-7 Whitingham Line 3.50 NE GM-8 Monroe Hill Barrier 1.80 S GM-9 Dunbar Brook 3.20 SW GM-10 Cross Road 3.50 E GM-li Adams High Line 2.10 WNW GM-12 Readsboro, VT 5.50 NNW GM-13 Restricted Area Fence 0.08 WSW GM-14 Restricted Area Fence 0.11 WNW GM-15 Restricted Area Fence 0.08 NNW GM-16 Restricted Area Fence 0.13 NNE GM-17 Restricted Area Fence 0.14 ENE GM-18 Restricted Area Fence 0.14 ESE GM-19 Restricted Area Fence 0.16 SE GM-20 Restricted Area Fence 0.16 SSE GM-21 Restricted Area Fence 0.11 SSW GM-22 Heartwellville 12.60 NNW GM-23 Williamstown Substation 22.20 W GM-25 Whitingham, VT 7.70 NNE GM-27 Number 9 Road 7.60 ENE GM-29 Route 8A 8.20 ESE GM-31 Legate Hill Road 7.60 SSE GM-32 Rowe Road 7.90 S GM-33 Zoar Road 6.90 SSW GM-35 Whitcomb Summit 8.60 WSW GM-36 Tilda Road 6.60 W GM-38 West Hill Road 6.60 NW GM-40 Readsboro Road 0.50 W Revision 16 4-14

TABLE 4.4 (Continued)

Radiological Environmental Monitoring Stations*

Exposure Pathway Sample Location and Designated Distance From Direction and/or Sample Code- the Plant (km) From Plant

5. DIRECT IF-1 ISFSI Security Fence 20 WNW RADIATION IF-2 Observation Stand 560 NW' (Plant) IF-3 ISFSI Security Fence 20 N IF-4 ISFSI Security Fence 34 NE IF-5 ISFSI Security Fence 28 E IF-6 ISFSI Security Fence 15 SE IF-7 ISFSI Security Fence 23 S IF-8 ISFSI Security Fence 38 SW IF-9 Restricted Area Fence (plant) 50 SE IF-10 Restricted Area Fence (plant) 55 SSE IF-11 Restricted Area Fence (plant) 135 SW IF-12 Restricted Area Fence (plant) 225 N IF-18 C. W. Intake 240 NNW

.IF-19 Restricted Area Fence (Admin Bldg)*** 170 W IF-20 Restricted Area Fence (Gate House)*** 235 WNW IF-40 Readsboro Road *** 700 N

• Sample locations are shown on Figures 4-1 through 4-7.

Station iX's are indicator stations, and Station 2X's are control stations (excluding the direct radiation stations)

Not included as part of the Radiological Environmental Monitoring Program.

a. As compliment to the routine air particulate monitoring, but not part of REMP, air particulate samplers (AP) will be used during demolition activities to provide supporting data to confirm that predicted levels of airborne radioactivity potentially released during building demolition is not significantly underestimated. Four fixed-location samplers will operate continuously during all radiation area associated structure demolition activities. A portable air particulate sampler will also be used periodically during demolition activities when the work zone dose rate exceeds 5 mR/hr (10% or greater of the demolition control limits given in Table 3.1, note b). The portable air particulate sampler will be placed as close as practicable to the demolition activity in an area likely to be impacted by any dust releases to the atmosphere. The LLD requirements of Table 4.3 for air particulate analysis do not apply for air samples collected with a portable air sampler when run for periods of less than 1 week duration.

Revision 17 I

4-15

Figure 4-1 Yankee Plant Radiological Environmental Monitoring Locations Within 1 Mile (Airborne, Waterborne and Ingestion Pathways)

Revision 17 4-16.

Figure 4-2 Yankee Plant Radiological Environmental Monitoring Locations Within 12 Miles (Airborne, Waterborne and Ingestion Pathways)

Revision 17 4-17*

Figure 4-3 Yankee Plant Radiological Environmental Monitoring Locations Outside 12 Miles (Airborne, Waterborne and Ingestion Pathways)

Revision 17 4-18.

Figure 4-4 Yankee Plant Radiological Environmental Monitoring Locations at the Restricted Area Fence (Direct Radiation Pathway)

Revision 17 4

Figure 4-5 Yankee Plant Radiological Environmental Monitoring Locations Within 1 Mile (Direct Radiation Pathway)

Revision 17 4-20.

Figure 4-6 Yankee Plant Radiological Environmental Monitoring Locations Within 12 Miles (Direct Radiation Pathway)

Revision 17 4-21.

Figure 4-7 Yankee Plant Radiological Environmental Monitoring Locations Outside 12 Miles (Direct Radiation Pathway)

Revision 16 l 4-22.

5.0 INSTRUMENTATION 5.1 Radioactive liquid Effluents Control 5.1 In accordance with Yankee Decommissioning Quality Assurance Program (Appendix D, Section B.4.a), the radioactive liquid effluent monitoring instrumentation channels shown in Table 5.1 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Control 2.1 are not exceeded. The alarm/trip setpoints of these channels shall be determined in accordance with the ODCM.

Applicability As shown in Table 5.1.

ACTION

a. With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than a value which will ensure that the limits of Control 2.1 are met, without delay, take actions to suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint, so it is acceptably conservative.

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

Exert reasonable efforts to return the instrument(s) to OPERABLE status within 30 days and if unsuccessful, explain in the next Annual Radioactive Effluent Release Report the reason for the delay in correcting the inoperability.

Surveillance Requirement SR 5.1 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations during the MODES and at the frequencies shown in Table 5.2.

Revision 15 5-1

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 ensure that the alarm/trip will occur prior to exceeding the limits of 10CFR 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, 10CFR Part 50.

A gross radioactivity monitor which provides for automatic isolation of liquid discharges on detection of radioactivity concentrations in excess of the values of 10CFR Part 20 (see Appendix B of ODCM), is included on the liquid radioactive waste effluent discharge line from the plant's test tanks. The automatic alarm/trip function provided by this monitor gives assurance as a final check that all conditions assumed, measured, or calculated that were used to determine effluent discharge rates have been appropriately made. This provides a degree of protection against calculational errors on discharge rate, operator errors in setting discharge flow, nonrepresentative samples used for Isotopic content of discharge volume, or crud releases during discharge which could lead to the discharge concentration limits of Control 2.1 being exceeded.

Composite samples are provided for continuous potential radioactive effluent pathways (i.e., ASW discharge) to give assurance that potential radioactive liquid releases to the environment are accounted for (See Figure 6-1).

Revision 16 5-2

TABLE 5.1 Radioactive Liquid Effluent Monitoring Instrumentation Minimum Channels Instrument OPERABLE Applicability ACTION

1. Gross Radioactivity Monitor Providing Automatic Isolation

a. Liquid Radwaste Effluent Line' (1) At All Times 4 1

2. Deleted I

3. Continuous Composite Samplers

a. Auxiliary Service Water Effluent Line2 (1) 5

b. SFP Dewatering Skid Effluent Line3 (1) 5

.I

4. Flow Rate Measurement Devices

a. Liquid Radwaste Effluent Line (1) ** 3

b. Auxiliary Service Water System Effluent (1) 3

c. SFP Dewatering Skid Effluent (1) ** 3

' A common radioactivity monitor (ASW-RM-001) provides indication for both the Liquid Radwaste Effluent Line and the Auxiliary Service Water Effluent Line. Actuation of the radioactivity monitor alarm will isolate test tank releases (if in progress).

2 The ASW effluent line composite sampler is intended to provide water samples during the discharge of water from SFP drain down only. ASW composite sampler not required during discharge of evaporator processed test tanks.

Revision 16 I

5-3

- I TABLE 5.1 (Continued)

• Table Notation 3 The SFP dewatering composite sampler is required only during SFP drain down processing operations.

'4 Monitor preventive maintenance, testing and calibration are permitted under the ACTION requirements.

• Via this pathway during SFP drain down only.

• • Via this pathway during releases.

I Revision 16 I

5-4

TABLE 5.1 (Continued)

ACTION Statements ACTION 1 - With the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, all radioactive waste effluent releases from the

- Test Tanks, TK-39 or direct SFP drain down shall be curtailed. Effluent releases from the Test Tanks or TK-39 may continue, provided that prior to initiating the release:

a. At least two independent samples of the Test Tankes contents, or two separate aliquots from the SFP dewatering skid composite sampler servicing TK-39, are analyzed in accordance with SR 2.1.1,

b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving, ACTION 2 - Deleted ACTION 3 - With the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, effluent releases via this pathway may continue, provided that the flow rate is estimated at least once per four hours during actual releases.

ACTION 4 - Deleted I

Revision 16 I

5-5

TABLE 5.1 (Continued)

ACTION Statements ACTION 5 - With the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, operation of the potential effluent release pathway may continue, provided that the associated effluent monitor is verified to be operating and grab samples are taken at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and analyzed for gross radioactivity (beta or gamma) at a limit of detection of at least 1.0 x 1O7microcuries/ml l at least weekly on a composite sample.

ACTION 6 - Deleted I

Revision 16 5-6 I

TABLE 5.2 Radioactive Liquid Effluent Monitorina Instrumentation Surveillance Reaulrements MODES in CHANNEL Which CHANNEL SOURCE CHANNEL FUNCTIONAL Surveillance Instrument CHECK CHECK CALIBRATION TEST is Required

1. Gross Beta or Gamma Radioactivity Monitor Providing Alarm and Automatic Isolation

a. Liquid Radwaste Effluent Llne(a) D P R(2) Ql) At All Times*

2. Deleted

3. Continuous Composite Samplers

a. Auxiliary Service Water Effluent Line D NA NA Q

b. SFP Dewatering Skid Effluent Line D NA NA Q

4. Flow Rate Measurement Devices

a. Liquid Radwaste Effluent Line D(3) NA R NA

b. Auxiliary Service Water Discharge D(3) NA R NA

c. SFP Dewatering Skid Effluent Line D(3) NA R NA **

Revision 16 I

5-7

TABLE 5.2 (Continued)

Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements

• • Via this pathway during waste effluent releases.

Via this pathway during SFP drain down only.

(a) A common radioactivity monitor (ASW-RM-001) provides indication for both the Liquid Radwaste Effluent Line and Auxiliary Service Water Effluent Line. Actuation of the radioactivity monitor alarm will Isolate test tank releases (if in progress.

I Revision 16 I

5-8

TABLE 5.2 (Continued)

Table Notation (1)- The CHANNEL FUNCTIONAL TEST also shall demonstrate that automatic isolation of the liquid radwaste effluent line and Control Room alarm annunciation occurs if any of the I following conditions, except as noted, exist:

a. Instrument indicates measured levels above the alarm/trip setpoint,

b. Circuit failure,

c. Instrument indicates a downscale failure (automatic pathway isolation, and Control I Room alarm indication.

(2)- The CHANNEL CALIBRATION shall include the use of a known radioactive source(s)

- positioned in a reproducible geometry with respect to the sensor whose effect on the system was established at the time of the primary calibration. Primary calibration is the determination of the electronic system accuracy when the detector is exposed in a known geometry to radiation from sources emitting beta and gamma radiation with fluences and energies in the ranges anticipated to be measured by the channel during normal operation. Sources should be traceable to the National Institute of Standards and Technology (NIST).

(3) - The CHANNEL CHECK shall consist of verifying indication of flow during periods of release. The CHANNEL CHECK shall be made at least once daily on any day on which continuous, periodic, or batch releases are made.

Revision 16 I

5-9

5.2 Radioactive Gaseous Effluents

• Control 5.2 Section Deleted I

Revision 17 I

5-10

Bases With the removal of nuclear fuel from the Spent Fuel Pool (SFP) and its placement in seared dry storage casks on the Independent Spent Fuel Storage Installation (ISFSI), along with the disassembly of all components of the waste gas treatment and holdup system, the potential for routine gaseous releases from the plant systems that were designed to contain, process, treat and direct exhaust of effluents to the atmosphere have been removed. Consequently, the need for instrumentation to control and monitor routine operating releases has been eliminated.

Temporary local air monitoring/sampling and portable HEPA filtration units will be used during remediation activities of contaminated areas on a case-by-case basis per guidance in Reference 1.

Revision 17 I

5-11

TABLE 5.3 Radioactive Gaseous Effluent Monitoring Instrumentation Table Deleted Revision 17 I

5-12

TABLE 5.3 (Continued)

ACTION Statements Deleted Revision 17 I

5-13

TABLE 5.4 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements Table Deleted Revision 17 I

5-14

TABLE 5.4 (Continued)

Table Notation Deleted Revision 17 I

5-15

5.3 Liquid Effluent Instrumentation Setpoints Control 5.1 requires that the radioactive liquid effluent instrumentation in Table 5.1 have alarm/trip setpoints in order to ensure that Control 2.1 is not exceeded. Control 2.1 limits the activity concentration in liquid effluents to the appropriate MPCs in 10CFR Part 20, as listed in Appendix B of the ODCM, and a total noble gas MPC.

Use the method below to determine the setpoints for the required instrumentation.

5.3.1 Method The Uquid Radwaste Effluent monitor response (cpm) for the limiting concentration at the point of discharge is the setpoint, denoted R, and is determined as follows:

R= (*2 ) (MPCC) (St) (Eq. 5-1) l Where:

f, = Flow rate past the test tank monitor (gpm).

f2 = Flow rate at the point of discharge (gpm).

St = Instrument response factor (cpm/(pCVml)).

MPCc = Composite MPC for the mix of radionuclides (pCi/ml).

MPCc = C C= E fl f 1/MPC 1 = 1{l fi/MPCi (Eq. 5-2)

I i I iI Where:

MPCj = MPC for radionuclide i" from 10CFR Part 20, Appendix B, Table II, Column 2 (pCVml). See ODCM Appendix B.

Revision 11 5-16

C, = Concentration of radionuclide i in mixture (pCi/ml).

f = Fraction of radionuclide i" in mixture.

Other setpoint methodologies also can be applied which are more restrictive than the approach used here.

The setpoint, R, may be administratively set lower to accommodate pathways which normally are nonradioactive (Auxiliary Service Water). The auxiliary Service Water is a normally clean system. The same radiation effluent monitor provides detection of the presence of an off normal condition that may have unexpectedly introduced radioactive contamination to this clean system. The alarm setpoint when only ASW cooling flow is in operation is set at two to three times background to give as early an alarm as practicable. SFP cooling flow is secured from the SFP heat exchanger when test tank discharges are made. This requirement allows the common radiation monitor to see either the expected clean ASW pathway flow when SFP cooling operations are on going (potential source), or the expected radioactivity in the test tank effluent flow when this discharge pathway is in operation.

5.3.2 Liquid Effluent Setpoint Example The effluent monitor for the test tank release pathway is gamma sensitive monitor. It has a typical sensitivity, S, of 2.8E+8 cpm per pCiml of gamma emitters which emit one photon per disintegration and a typical background of about 330 cpm.

The composite MPC and setpoint can be calculated based on the following example I data:

i _. fMPCG (fiCiml)

Cs-1 34 0.02 9 x 1Oe Cs-1 37 0.18 2 x 1O0 5 Co-60 0.80 3 x 10-5 Revision 11 5-17

MPCC-E 1,/MPCG 1 (Eq. 5-2)

(0.02/9 x 10-6 + 0.18/2 x 10-5 + 0.80/3 x 10-5)

MPCC = 2.6 x10-5 (gCi/ml)

For this example, normal liquid effluent flow rate, (f,), is assumed to be 2.8 gpm. Dilution water flow, f2, is assumed to be 90 gpm (equivalent to total flow of both contaminated and clean water). The setpoint for the monitor when the test tank effluent pathway is operating is then calculated for these example conditions to be:

R = (i.) (MPCC) (SE)

=(90 gPm)(2.6x1i0 5 fCi/m1) (2.8E+8cpmI(1iCilm1)) (Eq. 5-1)

=234,000 cpm This setpoint value may be administratively set lower than the maximum count rate for conservatism.

5.3.3 Basis The liquid effluent monitor setpoint must ensure that Control 2.1 is not exceeded for the appropriate in-plant pathways. The monitor responds to the concentration of radioactivity as follows:

(R) =(S)( ff Si JMON) (Eq. 5-5)

Revision 16 l 5-18

Where variables are the same as those in Section 5.3.1 except CMON = Total concentration (pCi/ml) seen by the monitor.

Si = Ratio of response from equal activities of radionuclide Nji to a reference radionuclide.

Calibration of the radiation monitors have established that the gross gamma detector response, SE f~s, was fairly independent of gamma energy as expected. Thus, the response is a function of radioactivity concentration and the gamma yield of the mixture. Since E fis1 is approximately one:

R = (S e) (CMoN) (Eq. 5-6)

For simplicity, assume that the monitor looks at a flow for fI. We know that:

• 1

(

= TI-J (C.O) (Eq. 5-7)

Where:

C = Total concentration at the point of discharge.

Solve Equation 5-5 for CMON and substitute into Equation 5-4 to get:

R ( fJ () (Se) (Eq. 5-8)

Revision 11 5-19

We defined C = I C, and define the composite MPCC such that:

I C =y CM (Eq. 5-9)

MPC' , MPCj The right side of the equation is the sum of the ratios of the MPC limits in 10CFR Part 20 (Appendix B of the ODCM). Solving for MPC,, the composite MPC for the mixture, we get the definition of MPC,:

EC; MPCC = (Eq. 5-2) v C;

, MPG, Substituting MPCC into Equation 5-6, we get the response of the monitor as MPCC is reached at the point of discharge, which is the setpoint:

R = (Lf I) (MPCC) (Se) (Eq. 5-1)

Revision 11 5-20

5.4 Gaseous Effluent Instrumentation Setpoints Section Deleted I

Revision 17 1 5-21

6.0 RADIOACTIVE WASTE TREATMENT SYSTEMS, EFFLUENT PATHWAYS, AND RADIATION MONITORS 6.1 Liquid Radioactive Waste Treatment Control 6.1 In accordance with Yankee Decommissioning Quality Assurance Program (Appendix D, Section B.4.a), the Liquid Radioactive Waste Treatment System shall be used to reduce the radioactive materials in the liquid waste prior to its discharge when the estimated doses due to the liquid effluent from the site (see Figure 1-2) when averaged over 31 days, would exceed 0.06 mrem to the total body or 0.20 mrem to any organ.

Applicabilitv At all times.

ACTION

a. With liquid waste being discharged without processing through appropriate treatment systems as defined in the ODCM and estimated doses in excess of the above limits, and if not applicable to IOCFR Part 50.73, prepare and submit to the Commission within 30 days pursuant to Control 7.4, a Special Report which includes the following information:

1. Explanation of why liquid radioactive waste was being discharged without treatment, Identification of any Inoperable equipment or subsystems, and the reasons for the inoperability;

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

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

Surveillance Requirement SR 6.1 Doses due to liquid releases shall be estimated at least once per 31 days in accordance with the ODCM. No dose estimates are required if the Uquid Radioactive Waste Treatment Revision 15 6-1

System has been continually used to reduce the radioactive materials in liquid waste prior to its discharge or if no liquid discharges have taken place over the appropriate 31-day period.

Bases The control that the appropriate portions of the Uquid Radioactive Waste Treatment 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." Control 6.1 implements the requirements of 10CFR Part 50.36a, General Design Criterion 60 of Appendix A, 10CFR Part 50, and the design objective of Section ll.D of Appendix I, 10CFR Part 50. The specified limits governing the use of appropriate portions of the Uquid Radioactive Waste Treatment System were specified as a suitable fraction of the dose design controls set forth In Section llA of Appendix I, 10CFR Part 50, for liquid effluents.

Revision 15 6-2.

6.2 Gaseous Radioactive Waste Treatriient Control 6.2 Deleted Revision 17 6-3 I

Bases Deleted

.. I Revision 17 l 6-4

6.3 liquid Effluent Streams, Radiation Monitors, and Radioactive Waste Treatment Systems Figure 6-1 shows the liquid effluent streams, radiation monitors, and the appropriate Liquid Radioactive Waste Treatment System.

6.4 In-Plant Liquid Effluent Pathways Batch effluent tanks called test tanks' collect the distillate from the atmospheric liquid radioactive waste evaporator. Normally, liquid waste accumulates at about 0.3 gpm and is processed at about 1 gpm. When the test tanks are full, they are sampled, analyzed, and released at a nominal 5 gpm:

Auxiliary service water provides dilution water flow. Typically, flow rates range up to 90 gpm for normal auxiliary service water use.

Calibrations of the radiation monitor has established that the gross gamma detector response was fairly independent of the gamma energy, as expected. Thus, the response is a function of the radioactivity concentration and the gamma yield of the mixture, but not the gamma energies of the mixture. The electronics of the monitor channel has an adjustable alarm setpoint.

Spent Fuel Pool drain down occurs after all spent fuel and other contaminated materials are transferred to the Independent Spent Fuel Storage Installation (ISFSI). Discharge of the SFP water In either batch or continuous mode will be treated by demineralization and filtration before release to Outfall 001. Figure 6.1 indicates the flow paths and in-line radiation monitoring prior to release. The primary flow path for SFP dewatering is to direct the flow (continuous mode) after waste treatment through a composite sampler (for representative sampling of the discharge stream) up-stream of ASW dilution flow and finally past the ASW in-line radiation monitor before release to Sherman Pond via Outfall 001.

Revision 17 6-5

6.5 In-Plant Gaseous Effluent Pathways Deleted I

Revision 17 I

6-6

FIGURE 6-1 Liquid Effluent Streams. Radiation Monitors, and Radioactive Waste Treatment System at the Yankee Plant r* ct-cs North~east ...... ... " ' " 1"',,,

" "u'a

§tkerMan I'OI4` I Oeer~eld Rve Sherman Da~m

-. ... ...... roxringency Path. SFF Derwatern

..... .-.... Primary PatN 6FP Dewateting I For construcdicddewaterinig. pirocess treatment. If needed. VA be provided before telease td the evroNtment. Treatment I&oraltdioatd~ty redticlori c~dId U015~de any combination d~ mpratoiNon, aut rw; ation bWedtwe installed or temporr prcess equimeil as approptinte for Ithe mclimmlie ponterg anwater i quality of th4-souroc

'"Ni ur wst~ir ndes bt sMImirited to. sourties ~Rdic at; cornstruton deang that are riot derived from enciosed plant sst~sems.

CMMRNGk6.

Revision 16 I

6-7

FIGURE 6-2 Deleted I

Revision17 6-8 I

7.0 REPORTING REQUIREMENTS 7.1 Annual Radiological Environmental Operating Report Control 7.1

a. An Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted to the NRC prior to May 1 of each year.

b. The Annual Radiological Environmental Operating Report shall include summaries, Interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment. The report also shall include the results of the land use census required by Control 4.2.

The Annual Radiological Environmental Operating Report shall include summarized and tabulated results of all radiological environmental samples taken during the report period pursuant to the table and figures in the ODCM. In the event that some results are not available to include in 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 report also shall include the following: a summary description of the Radiological Environmental Monitoring Program with a map of all sampling locations keyed to a table giving distances and directions from the reactor, the results of licensee participation in the Intercomparison Program required by Control 4.3, and a discussion of all analyses in which the LLD required by Table 4.3 was not achievable.

Revision 17 7-1

7.2 Annual Radioactive Effluent Release Report Control 7.2

a. Before May 1 of each year, a report shail be submitted to the NRC covering the radioactive content of effluents released to unrestricted areas during the previous calendar year.

b. The Annual Radioactive Effluent Release Report shall include a summary of the quantities of radioactive liquid and gaseous effluents released from the unit as outlined in Regulatory Guide 1.21, Revision 1, June 1974, "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," with data summarized on a quarterly basis following the format of Appendix B thereof.

In addition, the Annual Radioactive Effluent Release Report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit during the previous calendar year. This report also shall include an assessment of the radiation doses from radioactive effluents to MEMBER(S) OF THE PUBLIC due to the allowed recreational activities inside the SITE BOUNDARY (Figures 1-1 and 1-2) during the previous calendar year.

All assumptions used in making these assessments (e.g., specific activity, exposure time, and location) shall be included in the report. Historical average meteorological conditions shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the ODCM.

The Annual Radioactive Effluent Release Report also shall include an assessment of radiation doses to the likely most exposed real MEMBER(S) OF THE PUBLIC from reactor releases (including doses from primary effluent pathways and direct radiation) for the previous calendar year to show conformance with 40CFR1 90, Environmental Radiation Protection Standards for Nuclear Power Operation,' if Control 3.2 has been exceeded during the calendar year.

Revision 17 7-2

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

The Annual Radioactive Effluent Release Report shall include any changes made during the reporting period to the ODCM, as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Control 4.2.

Revision 13 7-3

7.3 Major Changes to Liquid Radioactive Waste Treatment Systems Control 7.3 Licensee initiated major changes to the liquid radioactive waste systems:

a. Shall be reported to the Commission in the Annual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by ISR and approved by the Site Manager. The discussion of each change shall contain:

1. A summary of the evaluation that led to the determination that the change could be made in accordance with IOCFR Part 50.59,

2. Sufficient detailed information to support the reason for the change without benefit of additional or supplemental information,

3. A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems,

4. An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents that differ from those previously predicted in the license application and amendments thereto,

5. An evaluation of the change, which shows the expected maximum exposures to MEMBER(S) OF THE PUBLIC at the SITE BOUNDARY and to the general population that differ from those previously estimated in the license application and amendments thereto,

6. A comparison of the predicted releases of radioactive materials in liquid and gaseous effluents to the actual releases for the period prior to when the changes are to be made,

7. An estimate of the exposure to plant operating personnel as a result of the change, and Revision 17 7-4

8. Documentation of the fact that the change was reviewed by ISR and approved by the Site Manager.

b. Shall become effective upon review by.!SR and approved by the Site Manager.

7.4 Special Reports Control 7.4 Special Reports shall be submitted pursuant to 10CFR50.4 within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference controls:

a. Uquid Effluents, Controls 3.1 and 6.1.

b. Gaseous Effluents, Control 3.5.l

c. Total Dose, Control 3.2.

d. Radiological Environmental Monitoring, Control 4.1.

Revision 17 I

7-5

8.0- REFERENCES

a. Regulatory Guide 1.109, "Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I," U.S. Nuclear Regulatory Commission, Revision 1, October 1977.

b. HamaWi, J. N., 'AEOLUS - A Computer Code for Determining Hourly and Long-Term Atmospheric Dispersion of Power Plant Effluents and for Computing Statistical Distributions of Dose Intensity.From Accidental Releases," Yankee Atomic Electric Company, Technical Report, YAEC-1 120, January 1977.

c. Regulatory Guide 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. NEP I and 2 Preliminary Safety Analysis Report, New England Power Company, Docket Nos. STN 50-568 and STN 59-569.

e. Yankee Atomic Technical Specifications.

f. Yankee Atomic Electric Company Supplemental Information for the Purposes of Evaluation of 10CFR Part 50, Appendix I, Amendment 2, October 1976 (Transmitted by J. L. French - YAEC to USNRC In letters, dated June 2, 1976; August 31, 1976; and October 8, 1976).

9. 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.

h. Slade, D. H., "Meteorology and Atomic Energy - 1968," USAEC, July 1968.

i. TDR-122374, Isotopic Standardization of Yankee Rowe Vent Stack Monitor."

j. Yankee Decommissioning Quality Assurance Program (YDQAP), Yankee Atomic Electric Company.

k. Issuance of NPDES Permit No. MA0004367; Letter to J. A. kay from R. Janson, US EPA, dated July 29, 2003 I. YNPS, RP Memo #03-024, "Airborne Effluent Dose Consequence of Building Demolition," Rev. 0, dated 09/03/03.

Revision 17 8-1

APPENDIX A DISPOSAL OF SEPTAGE Revision 7 - Date: M 21 9 A-1 Approved By:

UNITED STATES

o. NUCLEAR REGULATORY COMMISSION WASHI-INGTOtN. D.C. 20555 MAY 1 7 1990 Docket -No.50-029 Mr. George Papanic, Jr.

Senior Project Engineer - Licensing Yankee Atomic Electric Company 580 Main Street Bolton, Massachusetts 01740-1398

SUBJECT:

DISPOSAL OF SEPTAGE - YANKEE NUCLEAR POWER STATION By letter dated April 11, 1990, you requested NRC approval for a proposed disposal of sewage sludge containing very low concentrations of radionuclide according to 10 CFR 20.302. We have completed our review of your request and our evaluation is enclosed. We have found that your proposed transfer of the sludge by a contracted vendor to a public owned treatment works is acceptable.

Sincerely, KQe4 Patrick Sears, Project Manager

' Project Directorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Enclosed:

As stated cc w/encl:

See next page Revision 7 - Date: MAY 21 1990 A-2 Approved By:

Mr. George Papanic, Jr. Yankee Rowe cc:

Dr. Andrew C. Kadak, President and Chief Operating Officer Yankee Atomic Electric Company 580 Main Street Bolton, Massachusetts 01740-1398 Thomas Dignan, Esquire Ropes and Gray 225 Franklin Street Boston, Massachusetts 02110 Mr. T. K. Henderson Acting Plant Superintendent Yankee Atomic Electric Company Star Route Rowe, Massachusetts 01367 Resident Inspector Yankee Nuclear Power Station U.S. Nuclear Regulatory Commission Post Office Box 28 Monroe Bridge,.Massachusetts 01350 Regional Administrator,' Region I U.S. Nuclear Regulatory Commission 475 Allendale Road -

King of Prussia, Pennsylvania 19406 Robert M. Hallisey,.Director Radiation Control Program Massachusetts Department of Public Health 150 Tremont Street, 7th Floor Boston, Massachusetts 02111 Mr. George Sterzinger Commissioner Vermont Department of Public Service 120 State Street, 3rd Floor Montpelier, Vermont 05602 Ms. Jane M. Grant Senior Engineer - PLEX Licensing Yankee Atomic Electric Company 580 Main Street Bolton, Massachusetts 01740-1398 Revision 7 - Date: HAY 21 19S0 AW A A-3 Approved By: _<

SAFETY ASSESSMENT BY THE OFFICE OF NUCLEAR REACTOR REGULATION YANKEE ATOMIC ELECTRIC COMPANY YANKEE NUCLEAR POWER STATION DOCKET NO.50-029

1.0 INTRODUCTION

By letter of April 11, 1990, the Yankee Atomic Electric Company (Yankee) submitted, pursuent to 10 CFR 20.302(a)., a method for the routine disposal of septic tank waste containing very low levels of licensed material. Yankee proposed to periodically dispose of accumulated septic waste solids from the plant's sanitary system septic tank by transferring them to a public Sanitary Waste-Water Treatment Facility (SWTF) where they will be mixed with, processed with, and disposed as part of the sanitary waste generated from many sources.

Yankee proposed to make such disposals every one to two years over a period of 30 years.

In the submittal, the licensee addressed specific information requested in accordance with 10 CFP 20.302(a), provided-a detailed description of the licensed material, thoroughly analyzed and evaluated the information pertinent to the effects on the environment of the proposed disposal of the licensed material, and committed to follow specific procedures to minimize the risk of unexpected or hazardous exposures.

MAY 21 1990 Revision 7 - Date: A-4 Approved By: __ _ _ _ _

2.0 WASTE WATER-STREAM DESCRIPTION 2.1 Physical and-Chemical Properties The waste involved consists of residual septage (the accumulated settled and suspended solids and scum) produced by the sanitary sewerage collection and treatment system at the Yankee plant. To safely dispose of the plant's sanitary waste stream, the Yankee plant supplements the onsite septic system supplemented with offsite treatment at a SWTF.

The onsite septic system consists of a 7,000-gallon buried septic tank and a subsurface soil-absorption leach field. In the overall system design, the septic tank collects sludge and scum and partially separates liquids from the incoming sanitary waste.

The septage is retained in the septic tank, and the remaining conditioned waste-water liquid-flows into the underground leaching field for treatment.

The leach field is the terminal point of the onsite portion of the plant sanitary waste treatment process.

In the offsite portion of this process, the septage is removed from the septic tank and transported to a SWTF.

Revision 7 - Date: MAY 21 1990 Ak5 ApOproved By:

3-2.2 Radiological Properties The plant's sanitary system septic tank collects waste from the lavatories, showers, and janitorial facilities outside the Radiological Control Area (RCA).

No radioactivity is intentionally discharged to the septic system. However, plant investigations into the source of low levels of licensed material found in septic tank waste have identified very small quantities of radioactive materials, which are below detection limits for radioactivity releases from the RCA. It is suspected that these materials are carried out of the control area on individuals and spread to floor areas outside the RCA. Floor wash water from these areas is poured through a filter bag to remove suspended solids and dirt before the water is released into a janitorial sink. Although the wash water is returned to the RCA for disposal, if it is known to contain radioactivity, very small quantities can be released to, and accumulate in the sceptic tank.

The following values are estimates of the maximum total activity presently in the septic tank based on measurements of radionuclide concentrations in the liquid and solid phases:

Total Activity Nuclide (uCi)

Co-60 *1.94 Mn-54 0.057 Cs-134 0.082 Cs-137 0.248 TOTAL 2.33 Revision 7 - Date: MAY 21 1990 A-6 Approved By:,. I

3.0 PROPOSED DISPOSAL METHOD Yankee proposes to periodically dispose of accumulated septage from its septic tank by contracting with a septic tank pumper that is approved by the Board of Health, Rowe, Massachusetts and transfer the septage to a Massachusetts SWTF for treatment. This septic tank pumper will transfer the septage to an SWTF, where it is mixed and diluted with other raw sewage and introduced either into an anaerobic digester or an aeration pond for biological treatment. The resulting processed sludge from the SWTF is then mixed with sand and disposed of in a sanitary landfill, where it will be covered by clean soil daily. An alternate disposal means could result in the processed sludge being spread as a fertilizer, though generally for vegetation, such as sod, which is not consumed by humans. None of the region's SWTFs that receive sewage from local septic tank pumpers incinerate their sludge as a means of treatment.

This method of pumping the tank and transferring the septage to an SWTF is the same method normally applied to septic tank systems,..regardless of the presence of licensed material.

3.1 Septic Tank Waste-Procedural Requirements andLimits The licensee will perform a gamma isotopic analysis on a representative sample of waste from the septic tank no more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> before a contracted septic tank pumper begins to pump the waste from the tank to transfer to-a SWTF.

The licensee will collect at least two septage samples from the plant's septic tank by taking a volumetric column sample that will allow the licensee to Revision 7 - Date: MAY 21 199D A-7 Approved By:

determine the ratio of the solid content to the total content of the tank.. By determining the weight of the percentage of solid content of the collected sample and applying this value to the gamma isotopic analysis, the licensee will be able to estimate the total radioactivity of the contents of the tank.

To document the estimation of radiological effect of septage disposal, the licensee will perform these gamma isotopic analyses of the representative samples at the Technical Specification Environmental Lower Limit of Detection (LLD) requirements for liquids, as required in Technical Specification Table 4.12-1,

'Detection Capabilities for Environmental Sample Analysis,"

The radionuclide concentrations and total radioactivity identified in the septage will be compared to the concentration and total curie limits established herein before disposal. The following limits apply to these analyses:

1. The concentration of radionuclides detected in the volume of septage to be pumped to a disposal truck shall be limited to a combined sum of fractional Maximum Permissible Concentrations in Water (MPC) (as listed in 10CFR Part 20, Appendix B, Table II, Column 2), summed over all nuclides present, of less than or equal to 1.0.

2. The total gamma activity that can be released during septage transfer to any SWTF or combination of such facilities in one year (12 consecutive months) is limited to not more than 20 microcuries (equivalent to a maximum whole-body dose of 1 mrem to any individual in the public).

Revision 7 - Date: MAY 21 lqqn A-8 Approved By:

3.2 Administrative Procedures The licensee will maintain complete records of each disposal. In addition to copies of invoices with approved septic tank pumpers, these records will include the concentration of radionuclides in the septage, the total volume of septic waste disposed, the total activity in each batch, and the total accumulated activity of the septage pumped in any 12 consecutive months.

For periods in which disposal of septage occurs under this application, the licensee shall report, to the Nuclear Regulatory Commission (NRC) in the plant's Semiannual Effluent Release Report, the volume, liquid, and solid mass fractions, radionuclide concentrations in the liquid and solid fractions, and the total activity disposed.

4.0 EVALUATION OF ENVIRONMENTAL IMPACT The proposed method for disposal of septage is the same as currently used by all facilities designed with septic tanks for the collection of septic waste.

No new structures or facilities need be built or modified, nor any existing land uses changed. Septage from Yankee will be transported to an existing SWTF, where it will make up a small fraction of the total volume of sanitary waste treated each year. The normal method of septage handling and treatment Revision 7 - Date: MAY 21 1990 A-9 Approved By:

would involve dilution of Yankee's septage with other waste-water at a public SWTF. The processed sludge from the SWTF is usually buried in a sanitary landfill, thus limiting the potential exposure pathways to man. Otherwise, the sludge is widely dispersed in fertilizer, thereby preventing any buildup of activity from successive annual pumpouts from the plant's septic tank. This method of disposal will not affect topography, geology, meteorology, hydrology, or nearby facilities.

5.0 RADIOLOGICAL IMPACTS The licensee has evaluated the following potential exposure pathways to members of the general public: (1) inhalation of resuspended radionuclides, (2) ingestion of food grown on the disposal site, (e) external exposure to a truck driver or SWTF worker, and (4)-external exposure -caused by long-term buildup and external exposure from standing on the ground above the disposal site. The staff has reviewed the licensee's calculational methods and assumptions, and finds that they are consistent with regulatory Guide 1.109.1 The staff finds the assessment methodology acceptable.

Revision 7 - Date: MA 21 1990 A-10 Approved By:

IRegulatory (uide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance Vith 10 CFR Part 50, Appenedy !," Revision 1, October 1977.

Doses calculated in this manner by the licensee for the maximum exposed member of the public were as follows (based on a total activity awaiting disposal of 2.3 pCi, more than 80% of which is Co-60):

Maximally Exposed Individual/Whole Body (Child)

Pathway (mrem/year)

Ground Irradiation 0.099 Inhalation 0.0001 Stored Vegetables 0.0214 Leafy Vegetables 0.O0o0 Milk Ingestion* (0.0036)

TOTAL 0.12 The licensee then performed a similar calculation using a concervative upper bound activity of 20 pCi to be discharged in any one year. Based on this upper bound analysis, the dose to the maximally exposed individual member of the general public was estimated to be 1.1 mrem/year, as shown-in the following table:

Revision 7 - Date: MAY 2I. 199o A-11 Approved By:

Maximally Exposed Individual/Whole Body Pathway (mrem/year)

Ground Irradiation 0.980 inhalation 0.0004 Stored Vegetables 0.13 Leafy Vegetables 0.007 TOTAL 1.1 Based on this same total activity, the dose to truck drivers and SWTF workers was estimated to be 0.01 mrem/yr. These doses are within the design objectives of 10 CFR 50, Appendix I and well within the environmental standards for uranium fuel cycle activities as stated in 40 CFR 190.10(a) and are therefore acceptable.

6.0.

SUMMARY

AND CONCLUSIONS.

The-disposal of septage by transferring it to a public SWTF is in accordance with standard practices for treatment of the type of waste material generated by a septic tank and leach field sanitary waste system. Periodic pumping of the septic tank is necessary for the maintenance and continued operation of Yankee's sanitary waste system. Yankee requested approval for disposal of septic waste from the Yankee sanitary system to prevent failure of the sanitary system to adequately handle plant domestic waste.

Revision 7 - Date: MAY 21 1990 A-12 Approved By:

An alternate means of disposal would involve the treatment of the septage as radwaste. Such a disposal would require that the licensee stabilize, solidify, and dispose of the material at a licensed burial ground, requiring excessive cost and valuable disposal ground.

The results of the radiological analysis indicate that the public health effects of the biological activity and pathogenic constituents of such sanitary waste far outweigh the concerns related to any radioactivity that'is present.

By setting release limits that restrict the exposure for an individual to a maximum value of 1 millirem per year, Yankee ensured that radiological risks from the proposed disposal method are insignificant.

The proposed release limits represent a small fraction of NRC limits permitted for disposal of similar waste by licensed facilities who have their sanitary systems connected directly to a public sanitary sewerage system. These proposed limits are also well within the plant's allowable release limits for the discharge of normal liquid waste to the environment. Any resulting dose to any individual in the public is less than exposures caused by natural background radiation.

Based on our review of the proposed disposal of septage, the staff makes the following conclusions: (1) the radionuclide concentrations in-disposed septage will be a small percentage of permissible standards set forth in 10 CFR Part 20; (2) the radiation risk to workers involved in the disposal would be small compared to the routine occupational exposures at the Yankee Nuclear Power Station; (3) because the proposed action involves such very low levels of radioactivity, it will require no change in the decommissioning aspects of the Revision 7 - Date: MAY 21 1990 A-13 Approved By: ____________

facility and will require only insignificant changes in the handling or transport of radioactive material (septage); and (4) the licensee's procedures with commitments .aidocumented in the submittal are acceptable, provided that the submittal is permanently incorporated into the licensee's Offsite Dose Calculation Manual (ODCM) as an Appendix, and future modifications will be reported to NRC in accordance with licensee commitments regarding ODCM changes.

Contributors: J. Minns P. Sears Revision 7 - Date: MAY 21 1990 A-1'4 APpr oved By: I 1:&

Telephone (508) 779-6711 YANKEE A TOMIC ELECTRIC COMPANY TWX 710-380-7619 580 Main Street, Bolton, Massachusetts01740-1396 YANKE April II, 1990 BYR 490-42 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555

References:

. (A) License No. DPR-3 (Docker No. 50-29)

Subject:

10 CFR 20.302 Application

Dear Sirs:

Pursuant to 10 CFR 20.302, Yankee has prepared the attached application for the routine disposal of septage from Yankee Nuclear Power Station. This application utilizes guidance contained in NRC regulation 10 CFR 20.303 for the disposal of licensed material into a sanitary sewerage system.

We trust that you will find this submittal satisfactory, however, if you have any questions please contact us.

Very truly yours,,

YAMKEE ATOMIC ELECTRIC COMPANY George Papanic, Jr.

Senior Project Engineer Licensing Enclosure GP/emd Revision 7 Date M.AY 21 I3 Approved By A-15

YANKEE NUCLEAR POWER STATION APPLICATION FOR APPROVAL TO ROUTINELY DISPOSE OF SEPTAGE UNDER 10CFR20.302 Revision 7 - Date: MAY 21 199 A-16 Approved By: i uJ

TABLE OF CONTENTS TABLE OF CONTENTS . . .................... ii LIST OF TABLES .. iii LIST OF FIGURES .................................................... iv

1.0 INTRODUCTION

... . 1 2.0 WASTE STREAM DESCRIPTION.............................. . 2 2.1 Plhysical/Chemical Properties ....... 2 2.2 Radiological Properties. . . 3 3.0 PROPOSED DISPOSAL METHOD.................................. 5 3.1 Septic Tank Waste Procedural Requirements and Limits. 6 3.2 Administrative Procedures.... .. . 7 4.0 EVALUATION OF ENVIRONMENTAL IMPACT .. 8 5.0 EVALUATION OF RADIOLOGICAL IMPACT ............... .. 9 5.1 Septic Tank Sample Analysis Data... . 9 5.2 Pathway Exposure Scenarios . ... 10 5.3 Dose Assessments . . . ........... 11 5.3.1 External Exposure to a Truck Driver/SWTF Worker ..... 11 5.3.2 External Exposure Due to Long-Term Buildup .......... 12 5.3.3 Garden Pathway Scenario .. 14 5.3.4 Incineration Pathway Scenario . . . 20 5.4 Maximum Releasable Activity ..... 21 6.0

SUMMARY

AND CONCLUSIONS.................................. 23

7.0 REFERENCES

... .. 24

-ii-

,.Y 21131 Revision 7 -Date: MY2 2OA-17 Approved By: X -z

• LIST OF TABLES Number Title 1 Landspreading Ingestion Pathways (Adult) 25 2 Landspreading Ingestion Pathways (Teen) 26 3 Landspreading Ingestion Pathways (Child) 27 4 Landspreading Ingestion Pathways (Infant) 28

-iii-Revision 7 - Date:

MAY 21 I°SO

_ A-18 Approved By: /,-

LIST OF FIGURES 1umber i.. Page Yankee Plant Sanitary Waste.Disposal Process 29

-iv-Revision 7 - Date: MAY 21 19S A-19 Approved By:

YANKEE NUCLEAR POWER STATION Application for Approval to Routinely Dispose of Septage Under 10CFR20.302

1.0 INTRODUCTION

Yankee Atomic Electric Company (YANKEE) requests approval, pursuant to IOCFR20.302(a), of a method proposed herein for the routine disposal (typically, once every one to two years) of septic tank waste containing very low levels of licensed material over an extended period of time of 30 years.

Yankee proposes to periodically dispose of accumulated septic waste solids from the plant's sanitary system septic tank by transferring it to a public Sanitary Waste-Water Treatment Facility (SWTF) where it will be mixed with, processed, and disposed of, as part of sanitary waste generated from many sources. This is analogous to other Nuclear Regulatory Commission (NRC) licensed facilities who have their sanitary waste systems connected directly to a municipal sewer line. Part 20.303 of Title 10 to the Code of Federal Regulations already permits these NRC licensees to discharge licensed material into a sanitary sewerage system.

Routine maintenance of Yankee's septic system is necessary to ensure proper operation of the system. Periodic pumping of the septic tank to remove accumulated solids is necessary to prevent the carryover of solids into the subsurface leach field which would inhibit the soil absorption capabilities of the field.

This application addresses specific information requested in 10CFR20.302(a), and demonstrates that the periodic disposal of septage from Yankee's Sanitary Waste System over an extended periods of time (30 years),

under both normal and unexpected conditions, will not result in significant impacts either to the environment or to individuals in the general public.

Revision 7 - Date: IAAY 2 1 Ine A-20 Approved By: !L/4

2.0 WASTE WATER STREAM DESCRIPTION 2.1 PhysicallChemical Properties The waste involved in this application consists of residual septage (accumulated settled and suspended solids, and scum) associated with the sanitary sewerage collection and treatment system at the Yankee plant. The Yankee plant utilizes an on-site septic system supplemented with off-site treatment at a SWTF for the safe disposal of the plant's sanitary waste stream. Figure 1 is a schematic of the overall sanitary waste disposal process.

The on-site septic system consists of a 7,000 gallon buried septic tank and a subsurface soil absorption leach field. Sanitary sewage from the plant flows (estimated 2,600 gallons/day) into the septic tank. The septic tank function in the overall system design is for the collection of sludge and scum and partial separation of liquids from the incoming sanitary waste; Some of the solid particles settle to the bottom and form a layer of sludge, where greases and oils float to the surface creating a scum layer.

The septage is retained in the septic tank and the remaining conditioned waste-water liquid is permitted to flow into the underground leaching field for treatment. The leach field is the terminal point of the on-site portion of the plant sanitary waste treatment process. Some of the septage stored in the septic tank is reduced to liquid by bacterial action in the septic tank, but the rest-of the septage remains essentially untreated.

This material must be pumped out at regular intervals to prevent it from overflowing the tank and entering the leaching field (References 1, 2, 3, 4, 5, 6, 7, 8, 9, and 10) where it will clog the soil and eventually lead to septic system failure.

In general, septage pumped from septic tanks is discharged to a SWTF for treatment as part of the overall system design (Reference 10). The septage is then co-treated with other sanitary wastes at the SWTF. The septage pumped periodically from the Yankee plant has, in the past, been treated and disposed of in this fashion when no licensed material was determined to be present.

Revision 7 - Date: 1190A-21 Approved By: ________,___

The removal of the septage from the septic tank and subsequent transportation to a SWTF constitutes the off-site portion of the Yankee plant overall sanitary waste disposal process.

2.2 Radiological Properties The plant's sanitary system septic tank collects waste from the lavatories, showers, and janitorial facilities outside the Radiological Control Area (RCA). No radioactivity is intentionally discharged to the septic system. However, plant investigations into the source of low levels of licensed material found in septic tank waste have identified that very small quantities of radioactive materials, which are below detection limits for radioactivity releases from the RCA, appear to be carried out of the control area on individuals and accumulate in the septic tank. The suspected primary source of the radioactivity (i.e., floor wash water) is now either poured through a filter bag to remove suspended solids and dirt before the water is released into a janitorial sink, or the wash water is returned to the RCA for disposal.

An isotopic analysis, at environmental detection limits, of two composite volumetric sample columns of septage taken from the plant's septic tank identified the following plant-related radionuclides:

Activity Concentration iNuclid (p0i/kg wet +/- 1 sirma)

West Manhole East Manhole Sample Location Sample Location Co-60 92.4 +/- 3.9 13.2 +/- 2.2 Cs-134 5.9 + 1.3 Cs-137 9.2 + 1.5 3.2 i 1.0 After the initial analysis of the composite samples noted above, the samples were subsequently centrifuged into liquid and wet solids portions and reanalyzed. There was no activation or fission products identified in any of the liquid fraction samples indicating that the detected activity was in a form that had been carried out of solution with the solid fraction of the samples.

Revision 7 - Date: MAY_2___1 ____A-22 Approved By: ___________ __

Analysis of the resulting solid fraction of the septage indicated the following radionuclide concentrations:

Activity Concentration Nuclide (pCi/kg wet + 1 sigma)

I-West Manhole East Manhole.

Saxmple Location Sample Location Co-60 1,588 + 42 528 +/- 26 Mn-54 47 +/- 13 Cs-134 67 +/- 11

.Cs-137 203 17 100 +13 The original septic tank samples were volumetric samples representative of the distribution of solids and liquid from bottom to top of the tank. The ratio of the weight of the solid fraction sample to the weight of the.solid fraction plus liquid fraction sample allows a determination of the percentage of total solids content of the septic tank. For the waste sample from the west manhole, the solid fraction of the composite sample was found to be 0.024, or 2.4 wt. %. For the east manhole, the solid fraction of the total sample was 0.046, or 4.6 wt.Z. The principle radionuclide is Cobalt-60, which accounts for approximately 82% of all plant-related activity detected in the septage.

The total radioactivity content of the septic tank can be estimated by calculating the mass of solids present in the 7,000 gallon tank by taking the higher (conservative) solids fraction determined from the sample data. This is multiplied by the mass of septage in the tank and by.the highest activity concentration determined in the solids. As a result, the estimated maximum total activity is:

Total Activity Nuclide (PCi)

Co-60 1.94 Mn-54 0.057 Cs-134 0.082 Cs-137 0.248 TOTAL 2.33 Revision 7- Date: ~21I9 A-23 Approved By: _____________

3.0 PROPOSED DISPOSAL METHOD Upon approval from the U.S. Nuclear Regulatory Commission (NRC), Yankee proposes to periodically dispose of accumulated septage from its septic tank by contracting with a town-approved (Board of Health, Rowe, Massachusetts) septic tank pumper for the removal and transfer by truck of the septage to a Massachusetts SWTF for treatment. At the SWTF, the septage would typically be mixed and diluted with other raw sewage and introduced either into an anaerobic digester or aeration pond for biological treatment. The resulting processed sludge from the SWTF is typically then mixed with sand in a ratio of 50/50 and disposed of in a sanitary landfill, where it would be covered by clean soil daily. An alternate disposal means could potentially result in the processed sludge being landspread as a fertilizer, though generally for nonhuman-consumed vegetation, such as sod. None of the regions SWTFs which would be used by local septic tank pumpers were identified as incinerating their sludge as a means of treatment.

This method of tank pumping and transfer to an SWTF is identical to that normally applied to septic tank systems, irrespective of the presence of licensed material. Once the septage is pumped into the contract vendor's transporting vehicle, the situation is analogous to the handling of licensed material under IOCFR20.303. Part 20.303 of Title 10 to the Code of Federal Regulations already permits these NRC licensees to discharge licensed material into a sanitary sewerage system if certain conditions are met. Due to the remoteness of the Yankee plant's location, it is impractical to directly connect sewer lines to a facility to handle sanitary waste. In this case, a tank truck acts as a sewer line in transferring septage to a SWTF. The quantity and form (soluble or dispersable) of any licensed material contained in our septage is not affected by the means employed to transfer it to a SWTF for processing. 'Therefore, it would be the same whether the plant was directly connected to a municipal sewerage system or trucked its septage on a periodic basis to a SWTF.

?leA 21 I9M5 Revision 7 - Date: __A-24 Approved By:

3.1 Septic Tank Waste Procedural Requirements and Limits Gamma isotopic analysis of septic tank waste shall be made prior to transfer of the waste by a contracted septic tank pumper to a SWTF by obtaining a representative sample from the tank no more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to initiating.pump-out. At least two septage samples shall be collected from the plant's septic tank by taking a volumetric column sample which will allow for analysis of the solid's content of the tank. The weight percent of solid content of the collected sample will be determined and applied to the gamma isotopic analysis in order to estimate the total radioactivity content of the tank.

These gamma isotopic analyses of the representative samples will be performed at the Technical Specification Environmental Lower Limit of Detection (LLD) requirements for liquids (see Technical Specification Table 4.12-1, "Detection Capabilities for Environmental Sample Analysis") in order to document the estimation of radiological impact from septage disposal.

The radionuclide-concentrations and total radioactivity identified in

'the septage will be compared to the concentration and total curie limits established herein prior to disposal. The limits to be applied are as follows:

1. The concentration of radionuclides detected in the volume of septage to be pumped to a disposal truck shall be limited to a combined Maximum Permissible Concentration of Water (MPC) (as listed in loFR, Part 20, Appendix B, Table II, Column 2) ratio of less than or equal to 1.0.

2. The total gamma activity which can be released via septage transfer to any SWTF in one year (12 consecutive months) is limited to not more than 20 microcuries (equivalent to a maximum whole body dose of 1 mrem to any individual in the public).

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If the total activity limit is met, compliance with the self-imposed dose criteria will have been demonstrated since the radiological impact (Section 5) is based on evaluating the exposure to a maximally exposed hypothetical individual such that his annual whole body dose would be limited to approximately I mrem.

Both the concentration and total activity limits represent a small fraction of the allowable limits permitted under 10CFR20.303 to other NRC licensees who have their sanitary waste systems directly connected to a public sewerage system. If not for the biological nature of sanitary waste, the above release limits would also allow for the direct discharge of the waste under the plant's existing Technical Specification requirements for release of liquids to the environment.

3.2 Administrative Procedures Complete records of each disposal will be maintained. In addition to copies of invoices-with approved septic tank pumpers, these records will include the concentration of radionuclides in the septage, the total volume of septic waste disposed, the total activity in each batch, as well as total accumulated activity pumped in any 12 consecutive months.

For periods in which disposal of septage occurs under this application, the volume, total activity, and relative nuclide distribution, shall be reported to the NRC in the plant's Semiannual Effluent Release Report.

Revision 7 - Date: A-26 Approved By: f. e 2vJ

4.0 EVALUATION OF ENVIRONMENTAL IMPACT The proposed method for disposal of septage is the same as currently used by all facilities designed with septic tanks for the collection of septic waste. No new structures or facilities need be built or modified, nor any existing land uses changed. Septage from Yankee will be trucked to an existing SWTF, where it will make up a small fraction of the total volume of sanitary waste treated each year. As a result, there will be no impact on topography, geology, meteorology, hydrology, or nearby facilities by the proposed method of disposal.

Revision 7 -Date: HA 190A-27 Approved By: ~ 1 2~ e4.

5.0 EVALUATION OF RADIOLOGICAL IMPACT Radiological evaluations have been performed for the purpose of bounding the dose impact associated with the disposal of septage. The normal method of septage handling and treatment would provide for dilution of Yankee's septage with other waste-water at a public SWTF. The processed sludge would typically be buried in a sanitary landfill, thus limiting the potential exposure pathways to man, or widely dispersed if used as a fertilizer, thereby preventing any build-up of activity from successive annual pumpouts from the plant's septic tank. The dose assessments, however, did consider the maximum potential impact of long-term buildup of activity resulting from 30 years of placing septage waste in the same SWTF, with all the processed sludge assumed to be buried in one landfill disposal cell.

5.1 Septic Tank Sample Analysis Data The analysis of the septic tank's measured radioactivity, and its distribution between liquid and solid fractions, provides the bases upon which a dose assessment of disposal of septage can be made. The composition.of the septic tank waste determined from the sample analysis is:

Composite Sample East End Composite Sample West End Manhole Location Manhole Location Wt. Liquid 3.502 kg 3.460 kg Wt. Solid 0.087 kg 0.167 kg Solid fraction of the composite sample as collected is equal to:

Solid fraction = Wt. solid/(Wt. solid + Wt. liquid)

The solid fraction for the East End sample was 0.0242, and 0.0460 for the West End. The activity in the solid fraction was basically found to contain all the detected radioactivity as noted below:

East End Solids Sample West End Solids Sample (pCi/kg) Wet (pCilkg) Wet Mn-54 47 Cs-134 - 67 Cs-137 100 203 Co-60 528 1,588 MAY 21 1990 Revision 7 - Date: ________A-28 Approved By:

With the septic tank volume taken as approximately 7,000 gallons (26,500 liters), and assuming the maximum solid fraction (0.046) and maximum radionuclide concentration applies to the total tank's content, the total maximum radioactivity content is estimated to be:

Isotope lfLif Qe (Ci)

Mn-54 312.2 day 5.73 E-08 Co-60 5.272 yr 1.94 E-06 Cs-134 2.065 yr - 8.17 E-08 Cs-137 30.17 yr 2.48 E-07 5.2 Pathway Exposure Scenarios Radiological evaluations were performed for both the expected activities associated with handling, processing, and disposal of septage waste at a SWTF, and a hypothetical event causing undiluted septage release. The bounding case was determined to be associated with a hypothetical event which lead to the spreading of undiluted septage from Yankee's septic tank directly on a garden area where food crops are grown. The contracts with town approved septic tank pumpers will direct that septage be disposed of at a SWTF in Massachusetts. It is not expected that any disposal will occur other than at an SWTF. It is, therefore, not considered credible that successive bounding case activities could occur which lead to a long-term buildup of activity on a single minimum size garden plot.

In addition, since incineration of septic waste is not a current practice in the local area, the potential exposures associated with incineration are not of current concern. However, the establishment of a conservative-total whole body dose criteria for release of sanitary waste, via the above-noted garden scenario, assures that the potential resulting whole body dose due to incineration would not be expected to result in significant doses to any individual. This assessment is further detailed in Section 5.3.4.

The contributing pathways of exposure for the normal SWTF disposal process include:

1. External exposure to a truck driver.

2. External exposure to a SWTF worker.

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3. External exposure to an individual standing on the SWTF landfill after 30 years of buildup and decay.

The following garden exposure pathways were addressed for the maximally exposed hypothetical individual:

1. Standing on the ground plane.

2. Inhalation of resuspended material.

3. Ingestion of leafy vegetables.

4. Ingestion of stored vegetables.

5. 'Ingestion of milk.

6. Liquid pathways.

It should be noted that the milk pathway is mutually exclusive, to the other food production pathways since it would be impossible to support the grass-cow-milk-man exposure chain if the limited land area is utilized for the growing of food crops for direct human consumption. The two sets of ingestion pathways have been calculated so that the potential maximum impact can be assessed. Similarly, radionuclide movement into the ground water pathway would tend to reduce the impact of surface-related exposure paths and is, therefore, considered independently.

5.3 Dose Assessments 5.3.1 External Exposure to a Truck Driver/SWTF Worker The external dose rate from a 3,500-gallon tank truck filled with septage containing the total measured activity in the septic tank (2.33 pCi) was calculated for the purpose of estimating exposures associated with shipping the waste to a SWTF. A three-dimensional point-kernel shielding code for the determination of direct radiation from gamma radiation emanating from a self-attenuating cylindrical source (DIDOS-IV, Reference 14) was utilized to calculate the external dose rate from the, tank truck. The truck was modeled as a cylindrical radiation source with a radius equal to 1.22 meters and a length of 2;84 meters. A dose rate of 1.2E-04 mrem per hour for a point one meter from the end of the cylinder along the axis was calculated. No credit for shielding provided by the tank truck or cab was assumed. The dose to a Revision 7 - Date: IIAY 21 1990 A-30 Approved By: ,

truck driver making a 100-mile trip to a treatment facility at an average of 20 miles per hour plus a three-hour waiting period at the SWTF, is estimated to be 9.5E-04 mrem. It is concluded, based on the total activity limits proposed, that this pathway will not lead to significant exposure of any individual. It is also concluded that due to the sanitary properties of septage handling, a SWTF employee's direct exposure time is kept to a minimum. Using the dose rate estimated for the truck driver above, and conservatively assuming that it requires an employee at the SWTF a full eight-hour day to process each truckload of waste, and not taking any credit for dilution or increased distance from the waste, a waste processing facility employee's dose is also estimated to be 9.5E-04 mrem.

If the maximum activity content proposed to be disposed of each year were assumed as the source term (20 pCi), the dose to the truck. driver/SWTF worker is estimated to be less than l.OE-02 mrem using the same assumptions as noted above.

5.3.2 External Exposure Due to Long-Term Buildu, In order to assess the potential impact from the postulated buildup of activity resulting from 30 years of septage disposed at the maximum annual allowed activity content, it was conservatively assumed that the entire quantity of accumulated activity at the end of 30 years was buried in a common landfill disposal cell which was then available to the general public for uncontrolled access (8,760 hours0.0088 days <br />0.211 hours <br />0.00126 weeks <br />2.8918e-4 months <br /> per year).

For regional SWTFs, waste sludge is typically mixed with sand and placed in landfill disposal cells on a daily basis and covered by a layer of at least six inches of composited material before the end of each working day, as required by Massachusetts Department of Environmental Protection regulations (Reference 16). The landfill disposal cells range in size from about one acre up to about five acres. Aftei a cell is full, a final layer of compacted material is required to be placed over the entire surface of the cell to a minimum depth of two feet (Reference 16).

Revision 7 - Date: MAY 21 1993 A-31 Approved By: __

Analytically, if QO is the amount of radioactivity per tank full of.

septage for a give nuclide, then the total accumulated radioactivity Qe(max) disposed of after 30 pumpouts is given by:

Qe(max) = QO (1 + E + E2 + E3 + E4 +....

+ E2 9 )

Q0 (1 - E2 9 )/(l - E) (A) where:

E = exp(-XAt)

X = is the decay constant for the selected nuclide (1/year), and At = time interval between applications, assumed to be 1 year.

If the maximum total activity of 20 microcuries (with the same relative distribution as determined in the current septic tank analysis) were assumed to be released each year, then the accumulated activity at the end of 30 years is found in the following table:

X Q0 Qe(max)

Nuclide Half Life (l/year) (uCi/batch) uCi Co-60 5.27 y 0.1315 16.65 132.14 Mn-54 312. d 0.8109 0.49 0.88 Co-134 2.07 y . 0.3357 0.70 2.45 Co-137 30.2 y 0.023 _2.15 46.04 Total 20 182 If the 20 microcuries per year limit is assumed to be all Co-60, then the resulting accumulated total after 30 years would be 159 microcuries, and result in a higher calculated dose than that from the above mix.

Assuming a minimum landfill disposal cell to be one acre in area, and that the 30-year accumulated activity (159 uCi; Co-60) was disposed-of in one year along with SWTF sludge that formed a minimum one foot layer which was placed immediately below the two-foot disposal cap of the cell, the resulting Revision 7 - Date:MY21S A-32 Approved By: f./

dose rate one meter above the ground surface was calculated to be 6.4E-07 mrem/hour. If it is also assumed that an individual remained on the landfill for a full year (8,760 hours0.0088 days <br />0.211 hours <br />0.00126 weeks <br />2.8918e-4 months <br />) without taking any credit for shielding by a residential structure, the total whole body dose would be 5.6E-03 mrem, or about 56% of the truck driver's/SWTF workers calculated exposure.

Since the landfill cap (2' minimum) effectively isolates the vegetation zone of the top 15 cm plow layer, no garden pathways of exposure are included. However, it is noted that the'30-year accumulated activity concentration spread over a one acre landfill disposal cell would result in an area density of only 3.7E-03 microcuries per square foot. This is approximately a factor of 11 below the surface area density of the garden pathway scenario in Section 5.3.3 for the bounding case of placing 20 microcuries directly on a 500 ft2 garden. Therefore, even if it is postulated that an individual were to dig a cellar hole for a new home on the landfill site after closure, the resulting dose impact would still be bounded by the garden scenario as described below.

It is, therefore, concluded that for normal handling, processing, and disposal of septage at a SWTF, the maximum annual dose is received by the truck driver or SWTF worker handling the annual batches of septage pumped for disposal, and not the result of accumulated activity buildup over extended time periods.

5.3.3 Garden Pathway Scenario The radiological impact associated with an event which place undiluted septage directly on a garden was carried out using the dose assessment models in Regulatory Guide 1.109 (Reference 13), and in a manner consistent with the methodology employed by the plant's ODCM. Special consideration was given to the following:

1. The computation of an effective self-shielding factor to account for the effect provided by the soil after the waste is plowed or mixed in the top 15 cm surface layer.

Revision 7 - Date: MAY 21 12~l A-33 Approved By:

2. The definition of an annual activity release rate, which following a year's time of continuous release, would yield the ground deposition expected to prevail after a tank pump-out and spreading on the 500 ft2 garden.

3. The definition of an effective atmospheric dispersion factor to represent the resuspended radioactivity.

4. The proper representation of partial occupancy factors and usage data.

Landspreading. Resuspension. and Occupancy Factors If it is assumed that the garden plot is limited to a surface area of 2

500 ft , then the land-deposited radioactive material Se (Ci/m 2) following landspreading will be equal to:

2 Se Qe (Ci)/(500 ft

• 0.0929 m 2 /ft2 ) (B)

The denominator of this equation is equivalent to the (D/Q) deposition factor normally employed in the airborne impact assessment of deposited radionuclides; that is:

(D/Q) = 1/(500 ft2

• 0.0929 m2 /ft2 )

=.2.15E-02 (m 2 ) (C)

Following the application of undiluted septage on the garden, some of the radioactivity may become airborne as a result of resuspension effects.

The model used to estimate the radionuclide concentration in air above the disposal plot was taken from WASH-1400, Appendix VI. According to that model, the relationship between the airborne concentration Ae (Ci/m 3 ) and the surface.

deposition is:

Ae = Se (Ci/m 2 ) x K (1/m) (D)

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where: K is the resuspension factor and is taken to be equal to 1.OE-06 -(l/m)

(Reference 11) which is believed conservative due to the limited surface area involved and the irrigation provided to a garden which minimizes airborne dust.

The 500 ft garden size was selected based on the minimum surface area necessary to include a garden as part of the land-use census as required by Yankee's Technical Specification 3/4.12.2. This is the minimum area which could be expected to produce sufficient food to support the uptake assumption on food consumption noted below.

In addition, by limiting the garden surface area to 500 ft2 (a circle with a 3.85 m radius) the concentration of radioactivity in the garden is maximized since the concentration for any given surface area is physically limited by the total activity available in the septage. For direct radiation estimates from standing on the ground plane, a commonly used assumption of an infinite plane source (which can be approximated by a circle with a radius of 15 m) would in fact undercalculate the surface dose rate from that of a 2

500 ft garden by a factor of about 8 due to the dispersal of the fixed quantity of activity available to be spread. For use with the garden pathways of exposure, it is assumed that the septage is mixed in the top cultivated 15 cm of soil with no additional clean soil cover placed over it.

As for the occupancy factors for direct exposure to the ground deposition and for immersion in the resuspended radioactivity, 360 hours0.00417 days <br />0.1 hours <br />5.952381e-4 weeks <br />1.3698e-4 months <br /> was used for the radiological impact analysis. The 360-hour interval is believed to be a reasonably conservative time frame a gardener would spend each year on a plot of land or garden during the growing season in the northeast (average two hours a day for six months).

Garden pathway data and usage factors as applicable to the area in the v-.cinity of the plant are shown below. These are the same factors as used in the plant's ODCM assessment of the off-site radiological impacts due to routine releases from the plant, with the following exceptions.:

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1. The soil exposure time was changed from 15 years to 1 year to account for the discrete application of septage on a garden plot.

2. The fraction of stored vegetables grown in the garden was.

conservatively increased from 0.76 to 1.0.

3. The crop exposure time was changed from 2,160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> to 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> to reflect the condition that no radioactive material would be dispersed directly on crops for human or animal consumption, the deposition on crops of resuspended radioactivity being insignificantly small; that is, crop contamination is only through root uptake.

USAGE FACTORS Vegetables Leafy Veg. Milk Inhalation*

Individual (kg/yr) (kglyrxL (liters/vr) (.m/yr)

Adult 520 64 310 329 Teen 630 42 400 329 Child 520 26 330 152 Infant 330 58

• Inhalation rates have been modified to reflect an annual occupancy factor of 360 hours0.00417 days <br />0.1 hours <br />5.952381e-4 weeks <br />1.3698e-4 months <br />.

VEGETABLE PATHWAY Stored Leafy Vegetables Vegetables Agricultural productivity 2kg/m 2 ) . 2.0 2.0 Soil surface density (kg/m ) 240.0 240.0 Transport time to user (hours), 0.0 0.0 Soil exposure time (hours) 8,766.0 8,766.0 Crop exposure time to plume (hours) .0 .0 Holdup after harvest (hours) 1,440.0 24.0 Fraction of stored vegetables

.grown in garden 1.0 Fraction of leafy vegetables grown in garden 1.0 Revision 7 - Date: A-36 Approved By:

COW-MILK PATHWAY Pasture Feed Stored Feed Agricultural productivity 2kg/m 2 ) .7 2.0 Soil surface density (kg/m ) 240.0 240.0 Transport time to user (hours) 48.0 48.0 Soil exposure time (hours) 8,766.0 8,766.0 Crop exposure time to plume (hours) .0 .0 Holdup after harvest (hours) .0 2,160.0 Animals daily feed (kg/day) 50.0 50.0 Fraction of.year on pasture .5 Fraction pasture when on pasture 1.0 As noted above, liquid exposure pathways are considered independent from those associated with garden exposures. Since .the laboratory analysis data of septic tank waste shows that all the activity is associated with the suspended or settled solids fraction, and not dissolved in the liquid portion, transport of activity through groundwater would not be expected to lead to drinking water supplies being impacted by septage placed an farm lands. It is, therefore, not anticipated that the groundwater pathway could result in doses comparable to the direct surface exposure pathways. As confirmation of this, however, a methodology for groundwater analysis, as developed by Kennedy, et al. (1990) (Reference 12), was used as a check. This model assumes that the radionuclides on the ground are leached into the water table with a leach rate based on continuously saturated soil. Once into the water table, the radionuclides are immediately available for consumption. The volume of water used for dilution is limited to the quantity used by one person in one year (91,250 liters). No credit is taken by soil retardation of the nuclides, either during the leaching process or during groundwater movement. Consumption of water is assumed to be 2 liters/day. The resulting dose factors, by radionuclide, are listed in Table 3.4 of Reference 12.

Of the radionuclides detected in the septage, Co-60 is the dominant nuclide, and has the highest dose factors. The total effective dose equivalent from drinking water is 4.4E-6 mrem/yr for 1 pCi of disposed Co-60.

The maximum organ dose is 1.9E-5 mrem/year per pCi, with the organ being the LLI wall. These results are several orders of magnitude below the direct surface exposure doses as detailed below. The groundwater pathway is, therefore, not significant.

MAR 2 1- DA91 Revision 7 - Date: ________A-37 Approved By:

Direct Ground Plane Exposure To account for the gamma attenuation provided by the soil, it was necessary to carry out an appropriate shielding calculation. This was accomplished through use of the DIDOS computer code which computed the radiation levels from a cylindrical volume source with a radius of 3.85 m and a height of 0.15 m, with the receptor located along the axis, 1 m above the source.

The source density was set equal to 1.6 g/cc, which is equivalent to the Regulatory Guide 1.109 value of 240 kg/m 2 for the effective surface density of soil within a 15 cm plow layer. If the total activity content of the septic tank, as listed earlier, were assumed to be uniformly distributed in the source disk, the volume source dose rate is equivalent to a dose rate of 2.8E-04 mrem/hr. The total dose from standing on the garden area for 360 hours0.00417 days <br />0.1 hours <br />5.952381e-4 weeks <br />1.3698e-4 months <br /> each year is seen to be 0.099 mrem from the total activity content measure in the septic tank (2.33 pCi) being placed on the garden.

Garden Pathway Total Dose The maximum individual ingestion/inhalation exposure assessments resulting from garden crops or pasture grass grown on a septage disposal plot were added to the direct ground plane doses discussed above. This results in a bounding estimate of dose to a hypothetical maximum exposed individual. The whole body and critical-organ radiation exposures after a tank pump-out and spreading on a garden at a concentration level equivalent to the measured concentrations in septic waste are as follows:

Radiation Exposure Individual/Organ Maximum Exposed Individual 0.122 mrem/yr Child/Whole Body 0.157 mrem/yr Child/Liver Revision 7 - Date: MAY 21 19SO A-38 Approved By: ,f < y

The individual pathway contributions to the total dose are as follows:

Pathway-Dependent Critical Organ Doses Maximally Exposed Maximally Exposed Individual/Organ Individual/Whole Body (Child/Liiver) (Child)

Pathway (mrem/year) (mremlvear)

Ground Irradiation 0.099 0.099 Inhalation 0.0003 0.0001 Stored Vegetables 0.055 0.0214 Leafy Vegetables 0.0028 0.0011 Milk Ingestion* (0.019) (0.0036)

TOTAL 0.157 0.122 Tables 1 through 4 detail the internal dose breakdown by radionuclide and pathway of exposure. As can be seen in the results, the whole-body and maximum exposed organ dose are appropriately equivalent. This is6 due to the dominance of the external ground plane exposure pathway controlling the dose to both the organs and whole body.

5.3.4 Incineration Pathway Scenario At the present time, there are no known facilities for the incineration of septage in the vicinity of the Yankee plant. For completeness, however, we have addressed the radiological impact expected from incineration. This will preclude the necessity of revising this application request if such a facility becomes available in the future.

The basis for the. radiological assessment of incineration is a report by Murphy, et al. (1989) (Reference 15), in which they calculated individual and population dose impacts .from low level waste disposal scenarios. This report used a radionuclide distribution that was based on extensive studies of

• As described above, the milk pathway is mutually exclusive to the vegetable ingestion pathway; and, therefore, not added into the total.

Ay Revision 7 - Date: 21 A-3 Aprvd By: 4~

power reactor low level wastes. This distribution was similar to the measured distribution in the Yankee septage in that Co-60 and Cs-137 were the predominant gamma emitters.

The results of their analyses show that the transport worker receives the highest dose from the incineration scenario. The transport worker dose is approximately a factor of 5 higher than either the maximum incinerator worker or the maximum disposal site operator, and is several orders of magnitude higher than the maximum individual doses to the general public.

The dose to the transport worker has been discussed above (Section 5.3.1) for the off-site disposal of septage from Yankee. This transport worker dose will not change if the septage is incinerated, since it was conservatively assumed that the worker spends 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> traveling to the disposal site. Therefore, the dose to the individual landowner, from the garden scenario, will still be controlling for all disposal options, including incineration.

5.4 Maximum Releasable Activity The above analysis for landspreading on a garden the measured activity levels detected in the septic tank indicates that over 80 of the total whole body dose received by the hypothetical individual is due to direct external exposure to the ground plane. Of this direct dose component, Co-60 accounts for about 962 of the exposure. In determining a practical means by which any future detectable levels of licensed material can be limited to ensure that the controlling hypothetical individual's annual dose is limited to approximately 1 mrem or less, the sum of all measured gamma emitting nuclides can be assessed as Co-60 to determine the quantity of gross activity that, if released in septage, would limit the dose to 1 mrem.

Repeating the above controlling analysis for the event which placed the septage shipment directly on a garden plot, and assuming that the activity available is all Co-60, the total activity which relates to the annual dose Revision 7 -Date: MAY 21 tIS A-40 Approved By: __ _

limit criteria of 1 mrem is determined to be approximately 20 microcuries.

The breakdown by exposure pathway for this scenario, assuming an activity release of 20 microcuries in the form of Co-60 is as follows:

Maximum Exposed Individual/Whole Body Pathway (mrem/year)

Ground Irradiation 0.980 Inhalation 0.0004 Stored Vegetables 0.13 Leafy Vegetables 0.0068 TOTAL 1.1 All other scenarios for the normal treatment and disposal of septage, including postulated accumulation and build-up of activity at a single SWTF for a 30-year period (at 20 microcuries/year), result in radiological impacts to individuals which are approximately a factor of 100 or more below the whole body dose for the garden pathway.

The following summary compares the calculated whole body doses associated with normal handling of septage with the I mrem bounding event garden scenario. This demonstrates that by limiting the annual quantity of activity in septage to 20 microcuries, the expected dose impact for disposing of septage at a SWTF will in fact be well below a dose criterion of 1 mrem/year:

'Maximum Whole Body Annual Dose Scenario (mrem)

(a) Septic truck driver/SWTF worker. l.OE-02 (20 uCi Co-60 per year)

(b) SWTF landfill after closure. 5.6E-03 (30-year accumulation; 159 uCi Co-60)

Revision 7 - Date: A-41 Approved By:

6.0

SUMMARY

AND CONCLUSIONS The disposal of septage by transferring it to a public SWTF is in accordance with standard practices for treatment of the type of waste material generated by a septic tank/leach field sanitary waste system. Periodic pumping of the septic tank is necessary for the maintenance.and continued operation of Yankee's sanitary waste system. Approval for disposal of .septic waste from the Yankee sanitary system is requested to prevent failure of the sanitary system to adequately handle plant domestic waste.

Alternate means of disposal of the septage would involve the treatment of it as radwaste, with the subsequent need to stabilize, solidify, and dispose of the material at a licensed burial ground at excessive cost and a loss in valuable disposal ground volume.

The radiological analysis results indicate that the public health effects due to the biological activity and infectious constituents of such sanitary waste far outweigh the concerns due to any radioactivity which is present. By setting release limits which restrict the exposure to a maximum hypothetical individual of .1 mrem per year, it is ensured that radiological risks from the proposed disposal method are of no significance.

The proposed release limits represent a small fraction of NRC limits permitted for disposal'of similar waste by licensed facilities who have their sanitary systems connected directly to a public sanitary sewerage system.

These proposed limits are also within the plant~s current allowable release limits for discharge of normal liquid waste to the environment, with any resulting dose to any Individual in the public being far less than committed exposures due to natural background radiation.

Revision 7 - Date: MAY 21 1i A-42 Approved By:

7.0 REFERENCES

1. "Design Manual - On-Site Waste-Water Treatment and Disposal Systems,"

U.S. Environmental Protection Agency, EPA-625/1-80-012, October 1980.

2. "Septage Management," U.S. Environmental Protection Agency, EPA-600/8-80-032, August 1980.

3. "Handbook - Septage Treatment and Disposal," U.S. Environmental Protection Agency, EPA-625/6-84-009, October 1984.

4. "Septic Tank Care, U.S. Department of Health," Education, and Welfare, U.S. Public Health Service, 1975.

5. "Manual of Septic Tank Practice," U.S. Public Health Service, Publication No. 526, 1957.

6. "Your Septic System," Prepared for the Massachusetts Department of Environmental Quality Engineering, Publication No. 10043-32-625-12-77-CR, January 1978.

7. "Septic System", Massachusetts Metropolitan Area Planning Council, 1981.

8. "Septic Systems," Massachusetts Division of Water Pollution Control, Publication No. 12551-24-300-9-81-CR, 1981.

9. Clark, J. W., W. Viessman, and M. J. Hammer, "Water Supply and Pbllution Control," International Textbook Company, 1971.

10. Metcalf & Eddy, Inc., "Waste-Water Engineering: Treatment, Disposal, and Reuse," McGraw-Hill, 1979.

11. Cember, B., "Introduction to Health Physics," Page. 321, Pergamon Press, 1969.

12. Kennedy, W. E., Peloquin, R. A., "Residual Radioactivity Contamination From Decommissioning," NUREG/CR-5512, January 1990 (Draft Report for Comment).

13. Regulatory Guide 1.109, "Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR, Part 50, Appendix I, USNRC, Revision 1, 1977.

14. J. N. Hamawi, "DIDOS-III - A Three-Dimensional Point-Kernel Shielding Code for Cylindrical Sources, ENTECH Engineering, Inc., Technical Report P100-R2, December 1982 (updated to Version DIDOS-IV, October 1989, Yankee Atomic Electric Company).

15. Murphy, E. S., Rogers, V. C., "Below Regulatory Concern Owners Group:

Individual and Population Impacts From BRC Waste Treatment and Disposal,"

EPRI NP-5680, Interim Report, August 1989.

16. Massachusetts Department of Environmental Protection Regulations 310 CMR 19.15 (Disposal of Solid Waste in Sanitary Landfills).

MAY 21 I' A A o y Revision 7 - Date: A-43 Approved By:v 4_-7Zfi---

TABLE 1

• LA)DSPREADING INGESTION PATRWAYS (ADULT)

(2.33 UCI TOTAL ACTIVITY)

CHREM)

PATNYAY BOWE LIVER MONUEY LUNG Cl-LtI THYROID UHOLE BODY INHALATION 54 1m 0.00E+00 2.'93E-06 7.28E-07 1.04E-04 5.72E-06 0.00E400 4.66E2-07 60 co 0.00E+00 2.11E-05 0.0O0Ee00 1.09E-02 5.21E-04 0.004E00 2.71E-05 134 CS 3.17E-05 7.22E-05 2.44E-05 J.31E-06 J.85E-07 O.00E+00 6.19E-05 137 CS 1.07E-04 1.39E-04 4.98E-05 1.68E-05 1.88E-06 0.00E200 9.58E-05 TOTAL FOR PAThwAY 1.39E-04 2.35E-04 7.49E-05 1.1IE-02 5.30E-04 0.00E'00 1.SSE-04 STORED VEGETABLES 54 mm 3.10E-04 9.21E-05 0.0E+00 9.48E-04

• 0.00240 O.DWE400 5.91E-05 60 C0 0.00E+00 1 .78E-03 0.W0E+00 0.DOE+00 3.34E-02 0.00E400 3.92E-03 134. CS 2.24E-03 5.33E-03 1.72E-03 5.72E-04 9.32E-05 0.00E400 4.35E-03 137 CS 9.25E-03 1.27E-02 4.29E-03 1.43E-03 2.45E-04 O.O0E+0O 8.29E-03 TOTAL FOR PATHWAY 1.15E-02 2.01E-02 6.11E-03 2.00E-03 3.47E-02 O.0E0400 1.66E-02 LEAFY VEGETABLES 54 MR 0.00E+00 4.34E-05 1.29E-05 0.00E+00 1.33E-04 0.00E+00 8.29E-06

• 60 CO 0.00E+00 2.24E-04 0.00E+00 0.00E+00 4.20E-03 0.00E+00 4.93E-04 134 CS 2.91E-04 6.92E-04 2.24E-04 7.44E-05 1.21E-05 0.00E+00 5.66E-04 137 CS 1.142-03 1.56£-03 5.'31E-04 1.76£-04 3.03E-05 0.00E+00 1.02E-03 TOTAL FOR PATHWAY 1.43E-03 2.52£-03 7.68E-04 2.51E-04 4.38E-03 O.00E+00 2.09E-03 COU'HILX 54 mm 0.00E+00 2.39E-06 7.10E-07 0.002+00 7.31E-06 O.004E00 4.552-cT 60 co 0.OE+000 5.33E-05 0.00E+00 0.00E400 1.00E-03 0.00E+00 1.18E-04 134 CS .1 E-04 1.93E-03 6.25E-04 2.07E-04 3.38E-OS 0.0oE+0O 1.58E-03 137 CS 3.31E-03 4.53£-03 1.54E-03 S.11E-04 8.77E-05 a.ooE000 2.97E-03 TOTAL FOR PATHUAY 4.12E-03 6.51E-03 2.16E-03 7.18£-04 1.13E-03 0.00E+00 4.66E-03 25 Revision 7 Date MAY 21 1 A-44 Approved By: 4

TABLE 2 LANDSPREADING INGESTION PATIWAYS (TEEN)

(2.33 UCI TOTAL ACTIVITY)

CHREK)

PATHUAY SONE LIVER KIDNEY LUNG Ct-LLI THYROID WHOLE BODY . .

INHALATION 54 mg 0.00E+00 3.7BE-06 9.41E-07 1.47E-04 4.94E-06 0.00E+00 6.21E-07 60 CO 0.00E+00 2.77E-os 0.00E+O0 1.60E-02 4.75E-04 0.00E+00 3.63E-05 134 CS 4.28E-05 9.60E-05 3.19E-05 1.25E-05 E.31E-07 . 0.00E400 4.67E-OS 137 CS I.S0E-04 1.90E-04 6.tlO-05 2.7UE-05 1.90E-06 0.00E400 6.96E-05 TOTAL FOR PATHWAY 1.93E-04 3.17E-04 1.01E-04 1.62E-02 4.82E-04 O.00E400 1.53E-04 STORED VEGETABLES 54 mN O.00E40O 4.84E-04 1.44E-04 0.00E+00 9.93E-04 0.00E400 9.60E-05 60 CO 0.00E+00 2.83E-03 0.O0E+OO 0.00E+00 3.69E-02 0.00E400 6.37E-03 134 CS 3.65E-03 8.59E-03 2.73E-03 1.04E-03 1.07E-04 0.00E+00 3.98E-03 137 CS. 1.57E-02 .2.10E-02 7.13E-03 2.77E-03 2.98E-04 0.00E+00 7.30E-03 TOTAL FOR PATHWAY 1.94E-02 3.29E-OZ 1.00E-02 3.81E-03 3.83E-02 0.00E+00 1.78E-02 LEAFY VEGETABLES

54 Mg 0.00E+00 3.68E-05 1.10E-05 0.00E+00 7.55E-05 0.00E+00 7.30E-06 60 CO 0.00E+00 1.93E-04 0.00E+00 0.00E400 2.51E-03 0.00E+00 4_34E-04 134 CS 2.57E-04 6.0SE-04 1.92E-04 7.34E-05 7.52E-06 0.00E+00 2.81E-04 137 CS 1.05E-03 1.40E-03 4.77E-04 1.85E-04 1.99E-05 0.00E+00 4.88E-04 TOTAL FOR PATHUAY 1.31E-03 2.24E-03 6.80E-04 2.59E-04 2.61E-03 O.O0E+00 1.21E-03 COU MILX 54 mm 0.0OE+00 3.982-06 1.19E-06 0.OOE+0O 8.15E-06 O.00E+00 7.8BE-07 60 cO *0.00E400 9.03E-OS . 0.00E+O0 0.00E400 1.18E-03 0.00E+00 2.032-04 134 CS 1.41E-03 3.31E-03 1.05E-03 4.02E-04 4.12E-05 0.00E+00 1.54E-03 137 CS 6.00E-03 7.99E-03 2.72E-03 1.06E-03 1.14E-04 0.00E+00 2.78E-03 TOTAL FOR PATHWAY 7.41E-03 1.14E-02 3.77E-03 1.46E-03 1.34E-03 0.00E+00 4.52E-03 26 Revision 7 Date MAY 21 1990 A-45 ' Approved By

I

TABLE 3 LAJNOSPREADt1KG INGESTIOU PATUWAYS CCMILO)

(2.33 UCI TOTAL ACTIVITY)~

(MIREK)

PATHUAY BOUE LIVER KIDNEY LUNG CG-LU THYROID WHOLE BODY INHALATION 54 KM O.0OEOO0 3.17E-06 7.41E-07 1.17E-04 1.69E-06 O.00E+00 7.03E*07 60 co O.0OE.OO '2.1.OE.OS O.OOE+00 1.29E-02 1.76E-04 O.OOE+00 4.15E-OS 134 Cs 5.54E-05 8.63E-05 2.81E-05 1.03E-05 3.27E-07 0.00E+00 1.91E-OS 137 CS 2.03E-D4 1.85E-04 6.32E-05 2.33E-05 8.10E-07 O.00E+00 2.87E-05 TOTAL FOR PATHUAT 2.58E-04 2.98E-04 9.20E-05 1.31E-02 1.79E-04 0.00E+00 9.00E-0S STORMD VEGETABLES 54 Mg 0.0bE.OO 7.2SE-04 2.03E*04 0.00E+00 6.08E-04 0.00E200 1.93E-04 sOc 0.0OE+O0 4.40E-03 O.OOE+00 0.OOE+00 2.44E-02 0.00E+00 1.30E-02 134 CS 8.42E-03 1.38E-02 4.28E-03 1.54E-03 7.45E-05 0.00E200 2.91E-03 137 Cs 3.80E-0Z 3.63E-02 1.18E-02 4.26E-03 2.27E-04 0.00E+00 S 36E-03 TOTAL FOR PATHWAY 4.6.4E-0Z 5.S3E-02 1.63E-02 5.80E-03 2.53E-02 O.O0E+00 2.14E-02 LEAFY VEGETABLES 54 FNg O.OOE+00 4.13E-05 1.16E-05 O.00E400 3.47E-05 O.DOE+00 1.10E-OS 60 CO 0.00E+00 2.25E-04 0.00E+O0 0.00E400 1.24E-03 0.00E200 6.62E-04 134 CS 4.45E-04 730E-04 2.26E-04 8.11E-0S 3.93E-06 0.00E+O0 1.54E-04 137 CS 1.90E-03 1.82E-03 5.94E-04 2.14E-04 1.14E-05 0.00E+00 2.69E-04 TOTAL FOR PATHUAY 2.35E-03 2.82E-03 8.32E-04 2.95E-04 1.29E-03 0.00E+00 1.10E-03 cOm MILX 54 PH O.00E+00 5.95E-06 4.67E-06 O.OOE+00 4.99E;06 0.00E+00 1.58E-06 60 co O.00E+00 1.40E-04 0.00E+OO 0.00E+00 7.77E-04 0.00E+00 4.13E-04 134 CS 3.2SE-03 S;3E-03 1.65E-03 5.93E-04 2.87E-05 0.00E+O0 1.12E-03 137 Cs 1.45E-02 1.3&-0Z 4.51E-03 1.62E-03 8.67E-05 0.00E+00 2.04E-03 TOTAL FOR PATKUAY 1.77E-02 1.93E-02 6.17E-03 2.22E-03 8.97E-04 0.00E+00 3.582-03 27 I

Revision 7 Date MAY 21 1990 A-46 Approved By:

TABLE 4 LAUKDSPREADING INGESTION PATHWAYS (INFANT)

(2.33 UCI TOTAL CTIVITY)

(Clu PATHWAY BOfNE LIVER KIDNEY LUNG Cl-LUl THYROID WHOLE BCOY IRHALATION 54 MR O.OOE.OO 1.5?E-06 3.69E-07 7.39E-05 5.22E-07 O.OOE0OO 3.69E-07 60 co 0.00E4V0 1.47E-05 O.0OE.OO 8.25E-03 S.B4E-05 O.OOE+OO 2.16SE-05 134 CS 3.37E-05 5.98E-05S 1.62E-05 6.7B&-06 1.14E-07 O.OOE+0O ,6.34E-06 137 CS 1.23E-04 1.37E-04 3.85E-05 1.59E-05 2.99E-07 0.OOE+O0 1.OZE-05 TOTAL FOR PATHWAY 1.57E-04 2.13E-04 5.S1E-05 S.'35E-03 5.94E-05 0.OOE+O0 3.84E-05 STORED VEGETABLES 54 MN O.00E4OO O.00E4OO O.OOE+00 O.OOE+OO O.0OE+OO O.cOE*OO O.COE+OO 60 cO O.OOE4-OO O.OOE+OO0.OO.O0EO O.OOEO00 O.OOE+00 O.OCE400 O.OOE+Oc 134 CS 0.00E4'O0 O.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 O.OOE.O0 O.OOE-+OO 137 CS O.OOE+OO O.OOE400 O.OOE+O0 - O.OOE+OO O.OOEE+OO O.OOE400 O.OOE.CO TOTAL FOR PATHWAY O.OOE+OO 0.OOE+00 O.OOE+00 O.OOE400 O.OOE+OO O.OOE+O0 O.OOE.OO LEAFY VEGETABLES 54 MR O.0OE+OO O.OOE+OO 0.OOEe00 O.COE+00 O.OOE+O0 O.OOE.OO O.OOE+OO 60 cc O.OOE4OtJ O.OOEeOO C.OOE+OO O.OOE4O0 0.OOE+00 O.OOE.OO 0.00E4+0O 134 CS

• O0.OE+OO O.00E400 C.OOE.00 O.OOEeOO O.OOE+00 0.OOE+OO O.00E4OO 137 CS O.OOE+00 O.OOE.00 O.00E4OO O.O0E4O0 O.cOE+Oc O.IXIE4OO O.OOE+bO TOTALJfOR PATHWAY 0.OOE+O0 0O.E+OO O.OcE+OO O.OOE4400 O.O0E.OO O.00E4OO O.00E400 COU MILK 54 Kg a.OQE+OO 1.11E-05 2.45E-06 O.00E+00 4.06E-06 O.OOE.OO 2.S1E-06 60 co O.OOE'OO 2.86E-04 0.O0E+OO 0.0+0E4O 6.81E-04 O.OOE4O00 6.76E-04 134 CS 5.23E-03 9.76E-03 2.51E-03 1.03E-03 2.65E-05. 0.OOE.00 9.aSE-04 137 CS 2.31E-02 2.70E-02 7.25E-03 2.94E-03 S.45E-OS 0.OOE4OO I.M2-03

TOTAL FOR PATHWAY 2.83E-02 3.71E-02 9.77-03 3.97E-03 7.96E-04 0.OOE+0O 3.58E-03 28 Revision 7 Date MAY 21 1990 Approved By

c

F0 0

H-.

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................ @e@sb *^*.........................I ^~... 9^**X

. .**^ ovo*96o@

0 ^.@@*^*&^^@..........................................

P

-J SOIL ABSORPTIONi LEACH FIELD I rt 0

SEPTIC TANK LIQUID

- 'RETREATMEN IT) TREATMENT

_SCUM * ...

6 co YANKEE PLANT i LIQUID

• _ _-r---------------------- --------------

WASTEWATER

'10

=0 SLUDGE.-

..11SEPTAGE

... .................  : ET G

..t --------- --! TREATMENT ON-SITE PROCESS (See Note)

(LIQUIDS)

% . _1 _ .- - , . I I -WASTEWATER I aj OFF-SITE PROCESS I TREATMENT I

. (SEPTAGE) uaF iIa I' FACILITY..............................

tz

", Note: - Septago Hauler/Tank Truck Plpe Ltnc YANKEE PLANT SANITARY WASTE DISPOSAL PROCESS FIGURE t1

Appendix B Concentrations in Air and Water Above Natura 1 Background (10CFR20.1-20.602. Appendix B)

Revision '11 RIZ%120 B-1

Appendix B APPENDIX B To ff 20.1 -20.602--CoNCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND (See tootnotes at end of Appencrbx B]

Isotope ITawl I J Tabte It Element (atomic nM-ber o - o.1oi Col. 2-

_C_____ (F Waler (Wm) Water 1

Actinitur (89) AC 227 S 2X10" 6X10 1 2x10'"

8X 100-'

I S9C10' AC 228~ 9X 10-i1 S 2x10-8 9X10'3 Americum (95) ,Arn 241 ~ 4X10"*

1 S 6X10' 1 16XlO-'2 Am 242m -. S lXlO-4 2x10U' S- 3x 10" 3X10'2 Am 242 ........ 4X10-8 4x10'3 gxic,-,: IXIO-'

2X10-'

S 4x10,2 2xl0-'

Am 243 -~ 5x10-to S 4X 101: 3X10'1 Am 244 ~ I 1XIO-'* 1XIO-2 18IO-1 3

5XIb-2xl10' 8X10-4 6X10'1 Antimony Sb 122..$ 3x10'S 2

1x10 ' 5x10-. 2X10'S Sb 124 7xl10' 7xl10' 3 7X10"6 2X10's S~b12 5.. .4S 3X10- k2Xb0-6 1X10-'

3X10'1 3X10,3 SX10"11 Argon (18) A 27~ GX10-3 1X10':

A 41~ Su 2xl10' 4X10' Arsenic (33) As 73~

Su 7X10' 5x10-'

1X10- 6X10'4 S 1X10*' 5X10'5 4Kt0-1 5X10-S I 1X1O-1 As 76 4X10'1 S. 1X1O:? 5X10'2 3xlo-*

As 7 2X10'* 8X10-'

S 7X10-1 2x10-3 ixia-4 Astitine (85) At 211 1 2x10-"

IS _2X10 1XIO-6 Barium (56) Ba 131 I. JIX10-~. 2X10-4 S 7x10'1 1XIO-Ba 140- 4x10'1 ,2x10's IS 9xtO-" IxWO-Berkeliurm (97) 1 6 249- 3X10"1 S

tXlO'- 4xl10' CX 10-'

Bk 250~ 2xlO*: 5xIo-, 2xl0r' S 4K10-4 2xl10' IX10-'1 Beryllium (4 ) =- Be 7 .... L 2K10'1 .2x10'.

S lX 10-2 4X10-4 Bismuth (83) Si 206 ' IX 10'2 6Xb0-'1 4X10'S 2x10': 5X10'1 2X10'2 2X10 6X10'l 4x10O 1

12x10' 5X1O-"

2X10-"

Re~vision II B-2

Nuclear Regulatory Commission Pt. 20 [§§ 20.1-20.602], App.- B APPENDIX B TO §§ 20.1-20.602--CONCENTlIATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUNO-Continued tSec footnotes at end of Appenrdbx E]

Isotope ' Table I _ Table 11 Element (atomic number) j COL 1-Ak Wate (.u-irm) *II -k Col.

Watef 2-I. (pa/crml) (j~ailmr 61x0-' lx10IO' 2x10-' 4X10-'

I 1X10'2 si 212 3x10-9 4X10-4 S 2x10-' IX10-' 7x10* 4X10-4 Bromne (35). Br 82_. __ 1x10-' OXtO-' 4xI0-' 3X10-4 S lx 10-2 2X10-' 6x10' 4X10-S Cadmium (48) Cd 109._ S sx10-. 5x10'1 2x10-' 2X10-4 7x10-' 5 X10-' 3x10'* 2x10-4 Cd 11sm S 4X10-' 7x10' 1X10" 3X-10' I. 4X10-' 7x10' 1 X10-' 3Kx-s Cd 115 S 2x10-' 8x10-' 3x10-$

2X1O-7 I X10-' 6X10' ;xt10o' I

calcium (20) 3X1O-' 3xio-' 1 X10' .9x10-'

S Sx1O-' 4X 10' 2x10'-

Ca47- 2x10'6 6x10' SX105-IS 2x10-' 1 X10-' 6x10' 3X10-'

Catdomium (98) Cf 249 S 2x10-" 1x10-' SX1O-" 4X1O-4 1 xlO-" 4X10-4 7x10 ' 3X10-n 2x10-s I

Of 250 SXtO."2 4x10-' 2xJ0-3' 1XKb-'

5x10-"

S 7x10- 3x10-t 3X10-7 Ca 251 2x10-u GX1l-O4 4)10-'

I tixia-" 8910-4 3x10-" 3x10-'

a 252 S 6x10-" 2x10-4 2x10" 7x10-'

Su 2x10-' 1X10"' 7x10-'

01 253 exia-'. 4X10-8 3x10-"' 1x10-4 S 4x10'$ 3x10"-S lXlO' f 254 AX 1O-'

4x10-4 2x10"-' 1 X10'1 S 1X1O-'

4X1O-4 2X10-"

Catbon (6) C 14~ S 1XlO10 8X10-4 (co) S 3X107'2 1 X10-4 Cedum (ss) Ce 141 4X10-' 3X10-' 2x10" 9K10'$

2x10-' 5X10-9 9x10S S 3IxO-'

1Xb0-'

Ce 143 2x10-' 9x10' 4K10'-

S 2x10-'

t xto.' 3x10-' 7x10'- 410'-

Ce 144 3x10-" 1 X10S-I 3x10-'

6xto-2XIO-' 2x10-Kb 1x10'$

Cesium (55) -- Cs 131 S 7x10-' 4X10- 2x10-1 x10~' 3x10-2 9xba-'

S lX1O-2 Cs 134m_ 4X10-' 2x10-' lx10' 6x10-'

S 6x10-' 3x10-' 2x10'-1 tX10J' Cs 134. 4X10-SxtO-' 3xl&' x10- 9x10-'

Su 1x10-'

1X101 4X10" 4XO-S Cs 135 S 3x10-2 2X10- IX10-4.

9x10-6 7x10- 3X10'* 2X10-4 I 2X10-'

Cs 1368*- 1x 10-' 1 K10- 9XtD0' S 4X10-' 2x10-3 6x10-9 6x10-$

4 Cs 137 S 2x10-I x10-' 2x10-' 2x10-s 5X10-" . 4x10-'

Caom (I17) a 36 2xto- 2X10-' 1XbO-0 8X10-'

-S 2X10-3 6X100 6X10'-

2 2x10-' *1 x1o- 9X10'4 4X10'-

4 t x10~' 1 X10-' 7X 10' 4x 10 comium (24) _- cr51 5x10-' 4X10-3 2x10'-

2x10-'

2ix o-'1 5X10-' 8X10-' 2x10-'

Cobalt (27) Co 57 3X10-' 2X10'- 1 X10-' SXb10-2x10-' IX 10-2 6X 10Z 4X10-4 Co S8M 2X10'4 8X10- 6X10- 3X 10' 6XIO : 3XtO-' 2x10-'

8x10-7 Co 58ss 92XIO-8 x10-' 4x10-' 3x10-' 1X10-4 3X1O-' 3x10-' 2X10-' 9x10"$

6 Co 60_-_ 3x10-' 1 X10.' 1X1-0 5X10-'

I X10-' XtXO-' 3x10-' 3x 10' Copper (29) ___ ...... . Cu 642___, 2xo1-0 IX10-' 7X10 3x10-'

2xio-1 x10- 6x10-' 4x10-' 2x10-'

Curiu(96)m........_ Cmn 242__ 7x 10- 4x10-" 2X10-'

Revision I I B -3

Pt. 20 [§§ 20.1-20.602], App. B 10 CfR Ch. 1 (1-1-93 Edition)

APPENDIX B TO §§ 20.1-2O.6O2-CONCENTRATIONS IN AIn AND WATER A8OVE NATURAL BACKGROUND-Continued (See lootrotes at end of Append B1 isotope ' Table I Table 11 Element (atomic number)

______I___ Col. -A

,~/m) cA 2-l 2-Co. 1 Cae. NIMA Col. 2-(UCcml)_J Water 2xl10" 7x10- 2Xb0r-3 CX 10:'

Cm 243 6X10' 2X10- 5X10'11 1 x10': 2x10-'1 Cm 244 2x10-4

• 9x1O-"1 3X10-43 7X10'6 2x10-'

iXow 3x1O- 1' 3X10-*

Cm 245 5X10 u 2x 10-"3 A4xl0-4 1XI10w 8X10-' Ax 10-t 3K10-'1 Cm 246 5X10 1 12 2x10- 11 Ax 10-'

SI10-'1 ixIa-:. A K10" 3Xb0-5 Cm 247 SX10-u 1 x10-' A4X10-'

1IXO-14 6X10-' 2x10-$

Cm 248 6Kb0-1 1X10-'1 2x 10-"4 4K10-'

1)c1-I! 4X 10-' IX10-'1 Cm 249 1lx 0r- 6x10'2 4xl0-'1 2X10-'

1X10-' 6X10-1 A4Xi0-' 2x10- 2 Dysosium (66) Dy 165 aXlO-'9 1x10-' 9x10-' Ax 10-'1 2X10-'d 'K1b-I 7x10 4X10-'

Dy 16 2x1-'# IX 10-2 exb0-' Ax 10-2 2x10-T 7x 10-'1 Ax10-'

Eisteinium (99) Es 253 7x10-2 8x10-" 3KbO-" 2x10'6

'7X10-'

6XI 10" 2x l0-1 2x10-'s Es 254m sx1o-'1 5x'0-'

2x10-' 2x 10-'

6x1GO1 2x10's Es 254 2x10"1 5x1-lX 10-'1 I xbO-' 4x10"1 lX10-a Es 255 5x10-I. 8X10-' 2x 10-"1 3X10-6 AX10-"6 Ox 10-' 1 XI10- SXlO-5 Erbiun (68) Er 169 6X10-T 3X10-' 2x10-6 9x10'S Ax10'T 3X10-' 1 XbO-4 9K 10-'

Er 171 7x10-1 3X10-' 2X10-4 1ixo-'

3X10-3 '2xb0re 1xib-'

EupiZum (63) Eu 152 2X10-.' 1Xb0-' 6X10-5 (T12-92 hrs)_ AX 10-8 2X10-' 1x10-'1 Cx 10-Eu 152 2X10-' 2X10'3 AXIO10- OX lo-S (T/2=13 yrs)_ Ixia-9 2X10-2 8x10-'1 Eu 154 6X 10-' 1 x10-10 2X1crs 6X10'4 2x10-'1 2x10-s 7X10-'

Eu 155 6x10-2 3x10-a 2x10-'1 7x10-f 7x10-4 4x10'3 3Xb0-9 tfefnium (100) Fm 254 2x10-f I x 10-'

AX10 211 JX10-4 2x10-8 2x10-9 Fm 255 IX10-2 6X10-0" 3xi0-'

I X10*' AX1O0 3xib-a Fm 256 3x1O-* 3X10'6 1 x10- 9X10-'

6XI 10" 9x10-'

Fluoie (9) Fi1 2X10-4 2X10-2 SX10-'

1X1O-2 9X10-6 SX10-'

Gadojm (64) Gd 153 6XI10' 8x10-* 2x10-4 9X10-'1 6KbO-3 3x16-' 2XI0-4 Gd 159 9x10-.

2x10': 2x10-1 8X10-'

AXl0-1 2x10' lX10-' 6XbO'1 Galrum ip1) Ga 72 *2X10'1 lX10-'1 eX10'9 AX 10-2 2x 107'T IX1a-.' Cx 10-' 4x 10-2 Gemunium (32) _ -- Go 71 SXl0-I 4X 10-' 2X10-3 5x10-' 2x10-'J 2X 10-3 lX 10-'

Gold (79) Au 196.___ 5X10- 3 4 x10-a 2x10-' A4xl 0- 2x 10'4 2x10-Au 198.... 2x 10-3 lX 10-' SX10-5 lX 10-' 8x 10' 5XI10' Au199._. 4xl10$ 2x 10-'

7x10-8 4X 10-1 2 x1 ID

_-.- H181 2X 10-' 7 xi0-%

2xl0-' IX10-'

2xlcr2 7x10's Holmium (--. .__1Ho 166-- 5 9X10-' 7K 10-' 3x10-s Revision 1 1 B -4

Nuclear Regulatory Commission Pt. 20 [§§ 20.1-20.602], App. B APPENDIX 1 TO §§ 20.1-20.602-CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND-Continued fSee lootnotes at end of Appendix 81 Isotope Table I Table 11 Element (atomic number) WCalt Col. 2-I ((a I) (1 tCmin-) (jxCVmI I___

2X10- 9x10'1 6x~10- 3x10-'

Hydrogen (1) _ _ _. H3__._ S SX10' 1x10-' 2x10-' 3X10-'

5x10- 2x 101 3x1O-S 2X10-incium (49) 9 ) _ I 113m _ S 8XIO1' 4X10" IX10-'

7x10'1 4x10-2 2x 10' 1X10~'

In 114m S Ax 10-'

1IX 10 2x10-'

5X10-'

SX10-4 47x10"1 2x10'6 2x10'-

In 115m.. S 8x10:1 2x10'1 1 X 10-1 4x10-'

1 X10"' 6x10'4 2x10'4 iX10-'

In 11ttS__ S 2x10'7 3x10-3 9X10-'

lx 10-'

3x 10'4 3x10-' 9x10-'

Iodinoe43). I 125t25 S SX10't Ax 10'1 8x10-1 2x10-'

2x10-7 6x10-' 6x10'9 2X10-4 1126 S OX10'9 5X10'S 3X 10' 310-2 3X10-3 9X10'S 2129 S 2x10'1 9x10'1 1X10' 6X10-7x1O-6 6xl()0" 2x10-4 1131 - S 2x10"1

_ 9X10-9 6X10- 3X10-2x10-3 lX O10' 3X1O-2 6x10-'

1132 - S 2x10'7 2x10'3 3xl0'1 8X10-'

9x10-'j 5XlO-3 3x10-4 2X10-4 1133 S 3X10'8 2x10-' AX 10 '4 1X10-'

2x10-7 1X10-' 7x10'9 4X10O' 1134 S SXlO'2 4X10-2 6x10'1 2X10-'

axle..' 2x10'2 lx10I-' 6X10-4 1 135 5 S IXie-7 7x10-4 1 x10'9 4X10-4 4x1lo. 1 2x10-3 1XI10' 7x10-'

h num (77) j.Ir 190 S lXIO-4 6X10-' 4xlO'Q 2X 10-4 4X10'2 SXIO-3 lx 10-'1 2X 10-4 IC1922 ixie.? 4XIO-' 4X 10-S 3X10'4 1 x10-' 9x10"1 4X10-'

I 194 2x10'# 8X10-'1 3X10-6 4

S 2x10'" 9X10- 5x10-* 3X10-'

Iron (26) -- I Fe 55 2 9XlO-2 2x10- 3X10'0 0X10-4 S ix10-'1 7x10'2 3x10'4 2x10-'

Fe 59 IX10-' 2X10' 5X10-'1 6X10-'

5X10' 2X10-2 2x1(),' SX10-'

Krypton (36) Kr 85m -- ox10' 1x0'2 Kr 85. .I Sub ixia-9 3x 10-a Kr 87- Sub IX10-' 2x10'1 Kr Be _ Su IIxi-'9 2x10'8 Lantanum (57 La 140..... 2x10'2 7x10-4 5X10-' 2X10-S Sb lX10-'s 7X10-' .AX10O 2X10-'

Lead (82) Pb 203_ S 3x10'4 1X10-' 9X10-! 4XIO-4 2xl10' 1 XlO-4 6x10-'1 4XIO-'

Pb 210 S' -1x10-w 4x10"0 1X10-'

2x10"1 SX10' SX10-n 2x10-'

Pb 212_ S. 2x104 6x10' 6x10-" 2X10'-

2x10'4 7x10"1 2X10-'

Lutetium (71) Lu 177 S 6X10'1 3x10-. 2x10'6 1 X10-4 I. 5X10.? 3 X10-' 2x10'4 IX 10-4 Manganese (25) Mn 52 _- S 2x10-' 7X10' 3x10-'

1XIO*4 1 xlO-' Sx10-' 3X10-S Mn 54 --- AXIO'1 1 x1o*' 1.X10-'

4X1.9' 4X10-'

3X10-' I1xlO' 1X10-'

Mn 56_____ S 8X10-' 4X10-' 2X10-6 1 X1O-'

SX10-' 3X10-' 2x10'6 1X10-'

Mercuty (80) __ Hg 197m. _ S 7X10'1 4X10-6X10-3 ax101 2X10-'

Ox,1O-1 9X10-' 3 X 10' SX10-3 2x10-'

IXIa-'1 9X10-3 AX 10'4 3X10-'

3X 10-' 1x10: 9l0 S Xl0-Hg 203 __-._ 7X10'4 SX10~ 1 92XX1 0' 2X10-'

1X10-7 3xio-2 I XIO'9 lxIO1-Rcvision 1 1 B -5

Pt. 20 [§§ 20.1-20.602], App. B 10 CFR Ch. 1 (1-1-93 Ediiion)

APPENDIX B TO §§ 20.1-20.602-CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND-Continued

[See footnotes at end of Appendix P1 IsotopeI Table I Table It Element (atomic naunber = Col. 2- Cd.

Col ColWae I IWater (p0./mi) Wae

________ ____I____ __I__ ("afffmi) ___ (p~a/ml)

Molybdenum (42) 1____ Mo 99___ 7x10-' 3 X10 2 Xi0'4 2x10'7 1

37x10'- dx1S1 NeO&dyilmu(60) -- d14- 8 X10O" 2x10- 7x10 1s 3x10-'* 2X10O 8XIO1 3x10-1 Nd 147 AXIO-1 6x10-1 1 2xI0-7 2x10-- 6X10 3 Nd 149 2x107'1 3xt0-'

IXIC-'1 SX10' sxI0o-.: 3x to-'

NHepturium (93) Np 237. AXIO1 2~ SX 10-' 3x10'6 1X10? 3 X O10 Np 239___ 8x10-3' 4X10'3 2x10-'

IX 10'4 A x 16-5 1

7x10 ' 1XI 0' Nickel (28) - - NI 59 5X10'7 2x10-1 8X10'1 N1 63 - 6x 10: 3x10-1 3X10' 3x10-4 7x 10' 2x10'4 Ni 65 9x10-2j AxZ10- 3 5x10-1 3X10-Nioblun (Cotumblum) (41)_ Nb 93m_ iXiO- 2 2X10'1 Ax 10'4 2X10- 1 4x1lO' NbSS 2x10'7 x 10-2 1XIO-4 lxlo- 1 , 3X10-3 lx 10-'

Nb 97 _- 6X10'1 9x10-'

3x10-' 2x10'7 5X10'4 9x 10-'

Osmkium(76) JOs 1es 5X10-' 7X10's 5x10' 2x10-3 2x10'9 7x10-S 2X10-3 Os t91m. . 2X1-'! 4X10' 3x10'3 7x10'%

9xIO1' 2X10-3 Os 191 lxlo10-' 5X10'2 2x10'4 5X10-3 AX10-1 SX10'3 2x10'4 6x10'6 Os 193 AXiO'7 6x10'-

2x10-23 3X10 21 1X10- Ax 10' Sx10-'1 Palladium (46) 4Pd 103 __ 1X10'9 3 3x10'4 7x10'7 8XIO- 3x10'4 Pd 109__ 6x10'" 3X10 23 Sx10-1 9x10'S AXlO'7 2x10'2 7x10-$

Ptmsphoows (15) 41 32 - 3X10'4 7x10'4 5x10'4 2x10'4 2x10'5 8X10'8 7x10' 2x10's Platinum (78)

• Pt 191 t 8X10-7 4x10': 2xl10' lx10I-'

6XII0- Sx10-2 32X10-' 1xi0-Pt 193m _ 7x10'4 SX10-2 4x10-6 1x10-3 5x10'-, 3X10- 2 2X10'6 1X10'3 Pt 193 ixia-'8 SX10'2 2x.10-7 9X10'4 3X10 17 2x10'2 Pt 197m GxO'I- SX10-1 2x10'1 9x10'3 5X10'4 4X10-2 2x10'1 Pt 197 S BXIO'7 JX 10-'

6x10-7 4X10'2 2 7x10'1 S 2x10-n 3X10 1 2x10'1 Pkftnittm 494)Pu 238___ 5x10'4 3X10"8 Pu 239 S 2x10"1 1x10'4 I x10"1 5x10'9 S 4X10"1 ax 10'1 6x10-'4 3X 10-'

Pu 240---.. 2x10"1 3IxO-"2 AX 10"4 ex10'1 6X 10"1 Pu 241______4 S 9x10"1 7xj0-3 2x 10' 1 x10' 14 S 4 X10'1 Ax 10-21 3Xl -12 Pu 242._.__ 2x 10"1 x 10'4 3x10-'

S 9x1O-'

AX 10"8 2x10-4 S 1x102 8x10', 3x10'1 2x 10'1 Pu 244. S 2x10-' t 3x10"1 3x 10'1 3x105 1 2X10":

Polonum. (84) Po 210 ....... Sx10-" 2 xI 1O 7X10" 2XIO1' ax ia-'

Revision I1 B -6

Nuclear Regulatory Commission Pt. 20 t§§ 20.1-20.6021, App. B APPENDIX B TO §§ 20.1-20.602-CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND-Continued (See footnotes at end of Appendix B3 isotopeI Table I Table tl Element (atomic number) C.I-Ak CWate Cot. 1.-Air CWat.er

-. I___ (p10lmt Cot.w2-(AO/~MO ______

ote 2-(JACifmt)

Potassium 69) 2xIO1' 9XI- 7x Io- 3X10'1 1XlO-1' 6x10' A x 10-' 2x10-1 Praseodym-ium (s9)..-.............-..-. Pt 4.......... 2x 010 9x10" 7X10' 3x10'1 2x10'1 9XI0" 5x10' 3x10'1 3x10'1 1K -10 1x10-6 SX10-1 2 x10'1 6xI10' SX10-1 PromethIum (61) .. __... . Pm 147---..... 2X10'9 2x1O-1 Cx 10' 6X10'3 3K 10-' 2xI101 3 x 0'1 1X10' lx10I' 4X10'1 2xI10' 1X10'2 ex10'9 4XI1-'

Protoacdinium (91).......-IPa 230........ 2K 10'1 7X10'3 6XIO1" 2x10-1 7X10'3 3 Xl 0' 2xI10' 1

IXIO1 3X10- 4xIO1" 9K O10 8Ox 10 4 Xl 0' 2xI105 4x10'2 2x10'* lx 10-'

2xlO'* 3X10'3 6xb0'* 1K O10-Radium (88se)..- .1 Ra 223.___ 2xl10$ 6x10"1 7X.10'1 1Ix 10 8X10-" 4x10'1 2X10"9 1 Ra 224_ 7XI10 2 xl0-go 2x10'1 7xI010 2x10'4 2x10-1" 5K 10'4 Ra 226 ___ 4X10'1 3K 10-1 3 x 10 5X10"1 q9Kb-' 2x10-1', 3x10'3 Ra 2_28 __ 7xl10" Ox 10'1 2xK010 3xI0-8 7X10'4 lK O10' 3kI0'5 Radon (86_ Rn 220.__ 1XIO11 Rn 2222_ 3X10-1 3x`10-'

Rhenium (75U)-- Re 183 __ 3K10'4 2x10'2 9K 10 6X10'4 8x10'2 Sx 10': 3K 10'4 Re 1868_ 3xI1-' 2xI010 9x10-'S 2;(10'. 1xb0.23 8X10'1 SX 10-1 Re 187__ 9K 10'- 7K b0' 3xI10-'1 3xI10' 5K 10-, 4X1bO' 2x10'4 2~Xio-Re 18se --- 2x10'3 1Ix 10 6X O10 9X10'4 6X10-1' 3x10'1 8 10-S RWamn (45)-- Rh'103m_ 4X10'1 3xI010 1X 10-2 8X10'T 3X10'1 2xI010 lX 10-2 Rh 105__ Ax 10' 3xI0'1 I Kb-'

8X10'T 2xI0-1 K10IO' 5K10'7 Rubidium (37) - - Rb 86 2x10'3 1XlO-4 7xI010 2XI0-1 2X10'S 7x10'4 2XIO1' 1KbO-'

Rb 87 5K10'7 7Xl0-6 2XI 10- 2K10'4 lXlO-'1 Rutuefium (44) _ _ Ru 97..__. 2x10'6 8K 10-1 .4K 10' 2xI010 6x10-1 3K10'4 Ru 103..__. 2x10'3 2KIO1' 8X10'.

3xI0'* 8K 10-2 2x10'3 2x10' lX 10-'

Ru 105 7x10-'

3X10'2 2X10-4 1K 10' 5X10'7 Ru 106 8X10'6 3X10'1 1K O10' 6X10'1 3xI0'4 2XKO10" .1Kb-'

.Samartum (62).... --.-. Sftn I47............. S 7x 10"1 2X10-3 2K 10's2 6x10'3 3xK010' 2x10'3 9XIO1' 7K 10'3 I X1Kb-2 2x1K 4XK104 Sm 151_.... OK 10- 2 S I1Kb-I *1X10- 5X10'1 4Kx10'4 Sm153.. 5K10., 2x 10-3 2X 10-' OX 10'6 S 2X10-' lx10I-' 8K 10-'

4X10-'

Scandium (21)_...................... S 2x 10'1 axtO-' 4XK-10 lx 10-2 2xI0'6 IX10-3 8K10-1 4xI10-1 S 2 Xl -1 9X10'1 Sc 47..____.__. 6 x 0-' 3XI0- 2 S 5K O10- 2x10'4 9K 10'1 Sc 47.._........ _... 2x10'1 6xl10' 3X10 13 1X10'1 5XI10' 3x10'1 Selenium 34) ...... ...... .........._. Se 75 ..._.. 1X10'4 ex 10-3 Ax 104 3x 10-'

lx10.2- 4XI10' 3X10'1 Revision 11 B-7

Pt.-20 1§§ 20.1-20.602], App. B 10 CFR Ch. I (1-1-93 Edifion)

APPENDIX B TO §§ 20.1-20.602-CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUNO-Continued J

(See footnotes at end of Appendix BW isotope I Table I Table II Element (atomic numbe) = Co.I-Ai 2-ICl. Water Coli- Col. 2-Cc -k IWater.

Silicon (14). Si 31...

lX 10'1 3x 10, 2xl0 2 9X10O1 3

6Xl0- 3x10-'j 2x10-1 Saver (47) Ag 10OS 2X10-6 1X10*'

3x lo" 1 IxO-'

Ag 1111m........ .S Sx1O-* 9X10-1 7X1'IO 3xI10' 3xl-0" 3xI10' Agll111111... S I X10-3 1X10-' 4X10-1 2X10-' 8X10'1 4x10'1 Sodum (i i) Ha 22_____S 6x' XIOh94X1-3xI10" 3x10'1 Na 24........... 4X10'1 2x10-4 2x10'1 SX10*' 3x1- 11 Stwonlum (38) &SCsm ~S 3x10-'

2x10'1 1X10'4 7XI1-3 lxlO'1 7X10'3

&(85 ~S 3X10-18 3x10'3 8x109 *1x10-'

4X10-' 4x10'1 2x10'4 sr 89 ~S 1X10-'

SX10-a 3x10k" 3xIO1' 1X10'9 3xIO1' 3x 10'1 1 X10'3 3XIO1' 3xl10' 4X10-1 1IxO-' 2x1-0~ 4XIO1' 5c9ll1 S 4X10'1 2 X 107 2x10'4 7x10'*

3 X10-' 9x10-9 5XIf0' St92 ~S 2x10'6 7XIO1' 3x10'7 2x10'3 S4ffu (16) 535 S 2X10:3 9xlO* 6xl0S$

OX10'3 9X10'9 3x10'1 Tantalum (73) Ta 182............S lX 10-3 IIx10 2 7x10"11 4x10's TecnetGum (43) Tc 9Om ~S 3X10'1 1X10-3 2x10-1 3x10- I Tc 96. 3X 10'- 2X10-1 l10I- 41 2x10-' IxlO-' 8x10'9 5x10'1 Tc 97m- - S SX10'4 4x10'"

2X10'4 5X1071 2X10'4 Tc 97 . . 4X10'1 2X10'3 3X10'1 Tc 99m__ _SI 8X10 2- 1X10'4 6x10'3 iXIO-S 2X10'-

Tc 99 S. 2x10'1 7X10- 3X10'3 5x10-1 2x10'9 2x10'4 Teludium (52) Te 125 '~ S 4XIO1' SX10-3 3X10'3 4x1I08 .2X104 Te 127nL........ S 2x10'3 5X10'9 6xl10' 2x10'3 lXlO'9 5x<10' Te 127____ S 2x~10'4 SX10-3 6X1O-8 3Xto04 2XIO'1 SX10'3 3X10-4 2x10'4 Tel12Sm......... S JXJO-3 3x10'1 3x10'3 SXlO-'1 6X10'1 1x10'9 2X10-'

Te1129____ S 5X10'1 2x10-' 2x10'7 8X10'4 4X10-' 2X10-2 1x10'7 8x10'1 Tel131m.......... S 2x10-3 1x10-1 6X10'3 2x10'1 1XJ103 6x10' *'4X10's Te 132

• S 2xIO1' OX10-4 7X10'9 3x10'1 1x 110-1 6x10'4 4XIO1' 2x10'3 Teraium 1 ) Th 6........S I x 10-' 1X10'3 3X10'9 4X10-'

3x10'1 JXJ103 1x10'9 4X10-'

Thallium (81) _ lX 10-' 9X10-' 4x10-1 Tl 2001.......... S 3X10': 7xj1-3 2X 10-'1 4X10-8 2x10-1 9x10'3 7XIl0~ 3x10-1 9X10'1 3x10'1 5X10'3 2X10-1 4X10'3 3x10' JXJ0-4 TI 204 .................. S 2XIO10 8x10'1 7xi10' 6x10'1 3x10-3 2x10'1 ix i0-3X10-' 2x l0-3 9X1O-S 6x lo-,

Revision II B- 8

Nuclear Regulatory CommTssion Pt. 20 [§§ 20.1-20.6021, App. B

• APPENDIX B TO §§ 20.1-20.602-CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKcGROUNo-Continued

[See footnotes at end of Appendix 01 Isotope' TIbleI Table II Element (awffk number) *ICd. 1-Ak l Cc.t22-oC 1- l Water IWate I Wa2-Ttlodurn M - S

• X10-'1 1x1O-" 2x10-'1 9x1O-' 1 LX 10-' 6xI 10-" 2K 10-'

Th 22B. S 2x10-' 3 x 10-e 7x10-'1 6X10'2 2x10-"1 Ixia-'1 S AX 10-'

Th 2301 2x10-"1 8K O10- 2x 10'1 3X10-'1 3K O10-

'Th 231__- S 1x10-'1 7x10-' Lx t0- 2x10-'

AXb0-' 2x10-4 Th 232.___ S 3XIO1' LX 10-'1 1IO -12 2x10-'

IX10-1' Th natural _ S 36X10-"1 2x10-'

6x10-3 2X 10-1 2x10-5 Th 234_- S 2xl0-1 4x1O-O I X10-' 1XbO-0 2X10-S ThurKIm (69) Tm 170___ S lXI0-9 LX 10-'

4X10-1 lXb0-9 5K 10-'

Tm 171. . S 1 X10:' 1X1O-2 4x10-* 5X107' 8KbO-*

Sn 113__ S I 10K-'

Ox 10-'

4X10 50 2X10-' 2x10'*

Sn 125 S LX 10-' 4 Xl 0~ 8XI0-6 5X10-'1 3x10-*

Tungsten (Wolfran) (74) W 181 S 2X10-'9 8X10-'s 2xl0-s I 1X10-2 4X10-'

W185._ S 4x10-' SXl 8X10-2 4x10-*

W 187 S 3x1O-' 2x1O-3

• 2X10-6 2x10-3 lXlO-1 1 X 10-'

Uranium _., U 230_._ S 3x10'11 1 Xb0-1" I XlO-"4 4xI - 1 3X10-'1 U 232 S 8X10-' 3X10-' 1 3K 10-'1 5x10-'.

8X10-' OXlO0-U 233 _- S 9X10-' 2X10-tl 3XI10' IX10-"

4Kx10-" 3X10-'1 U 234 - - S4 6x10": 2x10-"' 3xIO1' lX 10-" 4xio-" 3XIO1' S' 98X10-'

U 235 SX10"41 3xl10' x 10-'

2X 10"1 3x1b0' U 236 _ S 6X10-" 3Kb0-1 I XcO-6" 8X10-'

3K 10-'.

U 238 _._ S. 7X10-"1 3X10-0 4x1O'5 1x10-1 SX10-" 4xIO's U 240 2x101 : 3x 10' 1

S'S' 2X10- IX 10-' 3x10',

6Sx10-0 U-natural- *IX10-"4 3xl10'

.1 5x10-" 3x10-$

Vanadurn(23) V4 S 2x10 2' 6XIO10 3x10'1 S 6X10'1 2x10-$  :~10I-1 Xenon (54) Xe 131m-........ Sub 2xl10s 4x10-!

Xe 133........ Sub 3X10-1 Xe 133m Sub 3 Xl 0' Xe 135............ I X10's Sub 7x10-' 1X10-'

Yttervbm (70) Yb17. S 2XIO11 lx to-4 6x10-' 2X10-1 I x 10-'

Yttriun (39) _ __ Y S IX 10-' 4X10-' 2Kxto-I 3x10-1 2x 10-'

3X10-'

S 6X 10' 8X10-' 3K 10-'

S 6 xl 0' 3XK1--'

S 2 x10-3 1X1O0' 3XK 10-I 3X10-1 2x10-3 lXlO0- 3X 101 Y92.................. S 4X 10' 1XIO-8 6x 10-1 3X O10 2K 10-' I x10-1 6K O10-Y 93 ...... S 2X 10'1 6XI10- 3x 10-*

t 1 x10' LX 10-' 3x10's Revision II B -9

Pt. 20 [§§ 20.1-20.602], App. B 10 CFR Ct. I (1-1-93 Edition)

APPENDIX B TO §§ 20.1-2O.602-CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND-Continued

• See footnotes at end of Appendix 8]

Isotope Table I Table 1I Element (atomnc number Cot. I-,Ak Col. 2- Col. I--Air Cot. 2-Water (C/q Water Zinc(30) Zn 655_ S 1X10-7 3X10-3 4x10'§ tXtO-'

I 6x10'- 5X10-7 2X10- 2XIO-1 Zn 69Sm. _ S 4X10-7 2X10-l 1X10- 7XtO-I 3X10- 2X1O-3 Xt10' 6x10-Zn 69_- - S '7X10- SX1 0- 2X10-' 2X10-3 2

. 9X10-' .5X10- 3X10- 2X10-'

Zirconium (40)_Zr 93- -.. S 1X10-7 2X1O 4X10-' 8X10-'

1 3X10-7 2x10-7 1XtO1 ' 8X10-'

Zr S 1X10'- 2X10'- 4X10-9 6x10-1 3X10- 2X1O-3 1X10'§ 6X10-$

Zr97 S tX10-7 SX1O-' 4X10-' 2X10-5 I 9X10- SX10-' 3X10-* 2x10-Any single radionuclide not lsted above _ Sub 1 X10- _ 3X10- -

with decay mode omther than alpha emission or spontaneous fission and with radioactive half-life less than 2 bours.

Ay single radionuclide notlsted above _ 3X10- 9X10- 1Xt0I-1 3X10-'

with decay mode other than alpha enission or spontaneous fission and wit radioactive halfltie greater than 2 htou.

Any single radlonuclrde not listed 6X10'- 4X10-7 2xt0-1 3X10-above. wich decays by alpha emis-sion or spontaneous fission.

'Soluble (S); Insoluble (1).

Sub- means that values given are o submersion in a serispherical infinite ooud of alrbomne ariatetl.

These raon concentratons are appropriate for protection from radon-222 comabined with its short-lived daughters.

Afteirrytiie the value in Table I may be replaced by one-third (7S) woing loveL- (A -working ler' Is defined as any cornbination of sho~t-I'ed radon-222 daughters. poloxiur-218. Iead-214. b'nvuth-214 and polonrm-214. in one liter of air.

witt regard to the degree of equilbritum. that wit result In the ultimate enission of 1.3x10' MeV of alpha particle energy.)

The Table 11value may be replaced by one-thirtieth (%.) of a wroking.level. The m-it on radon-222 concentrations in restricted areas may be based on an annual average.

'For soluble nixtures of U-238, U-234 and U-235 In air chenical toxicity may be the Simiting factor. If the percent by weight-enrichrient) of U-235 is less than 5. the concentration value for a 40hour workweek, Table I. Is 0.2 miligrams uranium per cbic meter of air average. For any enrichmnent, the product of the average concentration and timne of exposure durinq a 40hour workweek shall not exceed 8x10-3 SA pc-hr/mtL where SA Is the specific acvity of the uranium Inhated. The concentration value for Table It is 0.007 mi rigrams uranium per cubic meter of air. The specific-activity for natural uranium Is.

677x10-7 curies per gram U. The specific actvity fo other mbxures of U-238 U-23S and U-234. If not known, shall be:

SA=3.6x107cries/ gram U U-depleted SA=(OA+038 E-0.0034 E j 10-' EM0.72 where E is the percentage by weight of U-235. expressed as percent.

NoE In any case where there is a mixture In air or water of more than one radionuctide. the limiting values for purposes of tkhi Apponroic should be deterniined as fotlows

1. tt the Iderfty and concentration of each radiotuclade in the mixture are known, the liniting values should be derived as fotlows Oeternine. for each radionucdide in the nixtwre. the-fatio between the quantity present In the mixture and the limit otherwise-established in Appendix B for the specific radionuctide when not in a mixture. The sum of such ratios for all the radionucfides In the mixture may not exceed 1-- (.e. unit)

ExAmPLE: If radionuclides A. B. and C are present in concentrations CA. C.. and Cc. and If the applicable MPCs. are MPQ, and MPCG.and MPCc respectively. then the concentrations shalt be fimited so that the toltowtng relationship exists (CIMPCJ+ (Ch/MPG j+(CCiMPCj) E 1

-2. K either the identity or the concentration of any radionuclide In the mixture Is not known. the limiting values for purposes of Appendix B shalt be:

a. For purposes of Table t. Col. 1-6x10'"

b-For purposes of Table 1.Cot. 2-4 x10"'

c.For-purposesofTablel1.Col. 1-2x10 "

d. For purposes of Table 11.Cot 2-3x 10 3.1f any of the conditions specilied below are met. the corresponding values specified below may be used in lieu ot those specified in paragraph 2 above.

a. If the identity of each radionuclide in the rnixture is known but the concentration of one or more of the radionuctides in the mixture is not known the concentration fimit for the mixture is'the linit specified in Appendix *B- for the radionuclide in the mixture having the lowest concentration limh or b It the identity of each radionuclide in the mixture is not known, but it is known that certain radionuctides specified in Appendix *B- are not present in the mixture. the concentration lirit for the mixture is the lowest concentration limit specified in Appendix 8- for any radionuclide which is not known to be absent from the mixture; or Revision 11 B- 10

Nuclear Regulatory Commission Pt. 20 [§§ 20.1-20.602], App. C Table I Table It

c. Element (atomic number) and isotope COAr J Col. 2-Water Col. *-

Ak (uO/

Co. 2-Water r piC (jaO/ml) mtQ (pUcmq If it is known that Sr90. 1 125. 1 126. t 129. I 131 ( 133. Table It only). Pb 210. Po 210. At 21 1. Ra 223. Pa 224. Ra 226. Ac 227. Ra 228. Th 230. Pa 231.Th 232 Th-nat. Cm 248.Ct 254. and Fm 256 are no presen t 9x10' .. . 3xO-'

n it is knownthat Sr 90. 125. t 126. t129( 131. t 133. Table 11only). Pb 2t0. Po 210. Ra 223. Ra 226. Ra 228. Pa 231. Th-nat. Cm 248. Ca 254.

and Fin 2S6 are not present -_ ._ 6x10' ...... 2x10"4 If It is known mat S, 90. 1 129 (1 125. 1 126.1 131. Table U only). Pb 210. Ra 226. Ra 228. Cm 248.and al 254 are not present- 2xiO'_ 6x 10' If it is known that ( 129. Table only).Ra 226. and Ra 228 are ot present__ _ 3x10' . __ 1 XO1' n it Is known that alpha-emitters and Sr 90.1 129. Pb 210. Ac 227. Ra 228.

Pa 230. Pu 241. and Uk 249 are not present - 3x10-' I__ IxlO-f it Is known that alpha-emitters and Pb 210. Ac 227. Ra 228. and Pu 241 are not present_ _I 3x10- _ 1x10-"_

H it is known that alpha-emitters and Ac 227 are not present 3x 10 . " x 10_

If it is known that Ac 227. Th 230. Pa 231. Pu 238. Pu 239. Pu 240. Pu 242.

Pu 244.Cm 248. Ct 249 and C 251 are not presen 3X1O- _ 1X10"

4. If a mixture of radionuclides consists of uranium and its daughters hi ore dust prior to chemical separation of the uranium from the ore, the values specitied below may be used ko uraniurn and its daughters through radium226. instead of those from paragraphs 1. 2. or 3 above.

a- For purposes of Table i. Col. 1-1x10O" pz/lmf gross alpha activity; or 5X10" La/lml natural uranium or 75 maicrograms per cubic meter of a; natural uranunL-2

b. For purposes of Table tl. Col. 1-3x10' pLC/rml gross alpha activity. 2x10':pClmt natural uranium; or 3 micrograms per cubic meter of air natural uranium.

• S. For purposes ot this, note. a radlonuclide may be considered as not present in a mixture if (a) the ratio of the concentration of that radionuclide in the mriixture (CA) to the concentration lit for tat radionuclide specified in Table It of Appendix B'" (MPQF) does not exceed VYo. (ie. CAAfPCQ51/10) and lb)the sum of such ratios for all the radionucdides considered as not present In the mixture does not exceed V. Le.

(CA/"PQA+C6/MPC,_.+ E Y).

B-n Kevislon L I