Text

Rev 12 to Ynps Off-Site Dose Calculation Manual
	ML20198G638
Person / Time
Site:	Yankee Rowe
Issue date:	02/07/1997
From:	Cumming E, Harvey R, Strun M; YANKEE ATOMIC ELECTRIC CO.
To:	;
Shared Package
ML20198G629	List: ML20198G624; ML20198G631; ML20198G638;
References
	PROC-970207, NUDOCS 9709040222
	Download: ML20198G638 (210)
	v • d • e

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YANKEE NUCLEAR POWER STATION OFFSITE DOSE C A L C U L A TI O N- M A N U A L i

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The YNPS ODCM contains the methods necessary to show compliance with YNPS Technical Specifications that directly limit environmental radiation doses and environmental radionuclide concentrations. The methods were developed by the Yankee Atomic Electric Company for use during normal plant operations. The original methods and the revision procedure were approved by the U.S.

Nuclear Regulatcry Commission.This ODCM contains sufficient documentation for use or tegeneration of

, the methods by an experienced Health Physicist.

. YANKEE ATOMIC ELECTRIC COMPANY

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YANKEE NUCLEAR POWER COMPANY OFF-SITE 00SE CALCULATION MANUAL YANKEL ATOMIC ELECTRIC COMPANY NUCLEAR SERVICES DIVISION 580 MAlH STREET BOLTON, MASSACHUSETTS 01740 l

l l REVIEWE0 FORC MEETING l-. PREPARED BY/DATE BY/DATE ,, NO./DATE REVISION 8 6 ' = 4- l 4 8 7-#' k M M '~,Heeting No. S2-72 Q _;;-- J . / 9, / ' 9-2 p ,ef _ 74 August 19, 1992

, ,,,,,, , E W (- m(~/,fM% .necting so. 93-22 Af q ir,f773 -

3 , jp . yy flay 18, 1993 REVISION 10 R. b 4t.t, , bq ge,,z.J.,

a-p-ap 2/h neeting no. 93-28

_73 June 22, 1993

, M M 6cdt; = -

Meeting,No. 96-63 g pp, jffg, October 31, 1996

\\ [\ @ MQ REVISION 12 F4 n , M' 9 ah/q. '*"" '

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' REVISION RECORD' l P.

Revision Date Description 0 12/01/82 Initial printing. Approved by PORC 11/23/82. I Submitted for USNRC approval 12/03/82. )

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

2 D7!30/85 Addition of Intercomparison Program description to

, Section 4.0. Reviewed by PORC 07/30/85.

3 03/19/86 Addition of a PVS I-131 inspection limit to -

demonstrate compliance witn Technical Specification -3.11.2.1.b.

4 05/21/86 Change in milk sampling location. Samples no longer available at Station TM-11.

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

4 6 02/18/88 Change in liquid dose factors to reflect additional dose pathways, change in gaseous dose factors to reflect five-yet average meteorology. Change in

,((j.

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gaseous dose rate . factors to reflect-a shielding factor of.1.0. Deletion of feod product location TF-12 (samples no longer required after 10/31/86).

Update of fence line location and several building names and locations in Figure 4-4.

's 05/21/90 Addition of App..J1x 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 transmittrd 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 muzc -ii- --

,. -REVISION RECORD Revision Date Description 4 8 08/19/92- -2. _Section 1,0,-Introduction updated to reflect' the change in scope of the ODCL;

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 definitior:s pertinent to'the relocated Technical Specifications:

5. Tables 1.9, OPERATIONAL MODES, and 1.10.

FRE0VENCY NOTATIONS, added to Section 1.0;

6. Technical Specification 3/4.11.1.1, now referred to as Control 2.1, relocated to Section 2.0;

7. Technical Specifications 3/4.11.1.2 I-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, reiocated to Section 3 0;

8. Techn; cal 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 3/4.3.3.7):

Revision 8 mum -iii- .-

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'fl REVISION RECORD 1 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 Annua 1> Radiological Environmental Monitoring Operating *

(Control 7.1) and Semiannual Effluent Release Reports (Control 7.2) and Major Changes to the Liquid and GASEOUS RADI0 ACTIVE WASTE TREATMENT SYSTEMS (Control 7.3); and

13. Corresponding Technical Specification Bases reloccted with Technical Specifications to become part of- '

controls. r

b. All pages r(numbered.

9 05/18/93- Replacement of milk sampling location TM-12 with TM-

, 14 in Table 4.4 and Figure 4-2.

a5 10 06/22/93 Technical Specification 3/4.3.3.3 (now referred to

! C" 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 Liquid 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 1 fission process with the shutdown of power ,

operations.

_l 12 02/07/97 The requirement to submit an annual summary of

' hourly meteorological data with the Semiannual Radioactive Effluent Release Report due 60 days after January 1 of each year was eliminated.

Revision 12

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tIST OF AFFECTED PAGES Page Revision /Date 2 l 1 12 02/07/97 11 through 111 8 08/19/92 I

l iv through v 12 02/07/97 v1 8 08/19/92

. v11 through xv 11 10/31/96 1 1 through 1-21 11 10/31/96

] 2 1 through 2-7 11 10/31/96

-3 1 through 3 45 11 10/31/96 4 1 th ' gh 4 18 11 10/31/96 4 49 0 08/19/92 4 20 11 10/31/96 4-21 8 08/19/92 5 1 through 5 27 11 10/31/96-6-1 through 6-8 11 10/31/96

) 7-1 8 08/19/92

l. 72 12 02/07/97 l 7 3 through 7-4 11 10/31/96 7-5 10 06/22/93 R1 8 08/19/92 A 1 through A-48 7 05/21/90 -

B-1 through B 11 11 10/31/96 o

t Revision 12

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

O Revision 8-amuo -vi- -,

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ABSTRACl The Yankee Nuclear Power Station (YNPS) 0FF-511E DOSE CALCULATION MANUAL (0DCM) contains the methodology and parameters used in the calculation of o*f site doses resulting from radioactive gaseous and liquid effluents. in the calculation of gaseous and 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

, l Environmental Monitoring Programs required by Section 6.7.5 and (2) l descriptions of the information that should be included in the Semiannual Radiological Environmental Operating and Annual Radioactive Ef fluent Release

, - l Reports required by Specifications 6.8.2.3 and 6.8.2.b. With initial approval by the U.S. Nuclear Regulatory Commission and the YNPS Plant Operation Review Committee (P0kC) and approval of subsequent revisions by PORC (as required by the Technical Specifications) this manual is suitable to show compliance where referred to by the Technical Specifications and controls listed in this document. >

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4 Revision 11.

ausuo -vii- "

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TABtf Of CONTfNTS Pace REVISION RECORD . . . . . . . . . . , . . . . . . . . . . . . . . 11 LIST OF AFFECTED PAGES .. . . . . . . . . . . . . . . . . . . . . .v DISCLAIMER OF RESPONSIBILITY , . . . . . . . . . . . . . . . . . . vi ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vii TABLE OF CONTENTS . . . . . . . . . . . . . . . . . . . . . , viii LIST OF CONTROLS .... . . . . . . . . . . . . . . . . . . . . . xii LIST OF TABLES . ..... . . . . . . . . . . . . . . . . . . . xiii LIST OF FIGURES . . . . . . . . . . . . . . . . . . . . . . . . . . xy

1.0 INTRODUCTION

...-........................ 11 1.1 Summary of Methods. Oose Factors. Limits. Constants.

Variables, and Definitions , . . . . . . . . . . . . . . . . 11 1.2 Applicability of Controls and Surveillance Requirements (SR) .. . . . . . . . . . . . . . . . . . . . . i2 j 2.0 RADI0 ACTIVE L10010 EFFLUENTS . . . . . . . . . . . . . . . . . . . 2-1 2.1 - Off Site Concentrations . . . . . . . . . . . . . . . . . . . 2-1 2.2 Method to Calculate Off Site Liquid Concentrations . . . . . 2-S 2.3 Method to Determine Radionuclide Concentration for Each Liquid Effluent Pathway . . . . . . . . . . . . . . . . . . . 2-6 2.3.1 Test Tank Pathway . . . . . . . . . . . . . . . . . . 2-6 2.3.2 Service Water Systems Pathway . . . . . . . . . . . . 2-7 2.3.3 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 RsJioactive Liquid and Gaseous Effluents . 3-3 3.3 Dose Rate Due to Radioactive Gaseous Effluents . . . . . . . 35 .

3.4 Dose Due to Noble Gases Released in Radioacti"e Gaseous 3.5 Effluents . . . . . . . . . . . . . . . . . . . . . . . . . . 3-9 Dose Due to Tritium and Radionuclides in Particulate form .

With Half-Lives Greater than Eight Days . . . . . . . . . . 3-11 3.6 Dose C61culation Concepts . . . . . . . . . . . . . . . . . 3 13 3.7 Method to Calculate the lotal Body Dose from liquid Releases ... .. . . . . . . . . . . . . . . . . . . . . , 3-14 3.7.1 Method 1 . . . . . . . . . . . . . . . . . . . . 3-14 3.7.2 Method 11 . . . . . . . . . . . . . . . . . . . . . 3-15 3.7.3 Basis for Method 1 . . . . . . . . . . . . . . . . . 3-15 Revision 11 oruro -vi11- "

TABLE OF CONTINTS (Continued)

Pace e 3.8 Method to Calculate Maximum Organ Dose from Liquid Releases .........................319 3.8.1 Method I .... . . . . . . . . .. . . . . . . . . 3 19 3.8.2 Method II . . . . . . . . . . . . . . . . . . . . . 3-20 S 3.8.3 Basis for Method 1 . .. . . . . .. . . . . . . . . 3 20 3.5 Method to Calculate the Total Body Dose Rate from Noble Gases . . . . . . . . . . . . . . . . . . . . . . . . 3+21 3.9.1 Method I . . . . . . . . . . . . , . . . . . . . . 3-21 3.9.2 Method !! . . . . . . . . . . . . . . . . . . . . . .

3 22 -

3.9.3 Basis for Method 1 . . . . . . . . . . . . . . . . . 3 22 3.10 Method to Calculate the Skin Dose Rate from Noble Gases . . 3 24 3.10.1 Method 1 . . . . . . . . . . . . . . -. . . . . . . . 3 24 3.10.2 Method 11 . . . . . . . . . . . . . . . . . . .. . 3-25 3.10.3 Basis for Method 1 . . . . . . . . . . . . . . . . . 3-25 3.11 Method to Calculate'the Critical Organ Dose Rate from l Tritium and Particulates with Half Lives Greater Than Eight Days ........ . . . . . . . . . . . . . . . . . 3 27 3.11.1 Method 1 ..... . . . . . . . . . . . . . . . . . 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 l Gases (Kr 85) . . . . . . . . . . . . . . . . . . . . . . . 3 30 g 3.12.1 Method 1 . . . . . . . . . . . . . . . . . . . . . 3 30 3.12.2 Method 11 . . . . . . . , . . . . . _ . . . . . . . . . 3-31 3.12.3 Basis for Method 1 . . . . . . . . . . . . . . . . . 3-31 3.13 Method to Calculate the Beta Air Dose from Noble Gases . . . 3 33 3.13.1 Method 1 .... . . . . . . . . . . . . . . . . . . 3 33 3.13.2 Method 11 . . . . . . . . . . . . . . . . . . . . . 3-33 3.13.3 Basis for Method 1 . . . . . . , , . . . . . . . . . . 3-34 Revision 11 moto - -ie "

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12 .

d' TABlf 0F CONTENTS (Continued)

Pace 3.14 Method to Calculate tb9 Critical Organ Dose from Tritium and Particulates . . . . . . . . . . . . . . . . . . . . . . 3 35 3.14.1 Method I .. . . . . . . . . . . . . . . . . . . . . 3-35 3.14.2 Method 11 . . . . . . . . . . . . . . . . . . . . . 3-36 3.14.3 Basis for Method I . . . . . . . . . . . . . . . . . 3 36 -

3.15 Critical Receptors and Long Term Average Atmospheric Dispersion factors for important Exposure Pathways . . . . . 3-41 -

3.15.1 Critical Receptors . . . . . . . . . . . . . . . . . 3-41 3.15.2 Yankee Atmospheric Dispersion Model . . . . . . . . 3 42 3.15.3 Long lerm Average Dispersion factors for Critical Receptors . . . . . . . . . . . . . . . . . . . . . . 3-43 L

3.16 Method to Calculate Direct Dose from Plant Operation . . . . 3 45 4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING . . . . . . . . . . . . . . . 4-1

,- s 4.1 Monitoring Program . . . . . . . . . . . . . . . . . . . . . 41 4.2 Land Use Census . . . . . . . . .

(". )

4.3 Intercomparison Program . . . . . .

. . . . . . . . . . . . . . 4-10

. . . . . . . . . . . . . 4-12 4.4 Environmental Monitoring Locations . . . . . . . . . . . . . 4-12 5.0 INSTRUMENTATION . . . . . . , . . . . . . . , . , , , , , , , , , , 91 5.1 Radioactive Liquid Effluents . . . . . . . . . . . . . . . . 5-1 5.2 Radioactive Gaseous Effluents . . . . . . . . . . . . . . . 5-10 5.3 Liquid Effluent Instrumentation Setpoints . . . . . . . . . . 5 16 5.3.1 Method . . . . . . . . . . . . . . . . . . . . . . . 5-16 5.3.2 Liquid Effluent Setpoint Example . . . . . . . . . . 5-17 ,

5.3.3 Basis . . . . . . . . . . . . . . . . . . . . . . . . 5-18 5.4 Gaseous Effluent Instrumentation Setpoints . . . . . . . . . 5-21 ,

5.4.1 Method . . . . . . . . . . . . . . . . . . . . . . . 5 21 5.4.2 Gaseous Effluent Setpoint Example . . . . , . . . . . 5-22 5.4.3 Basis . . . . . . . . . . . . . . . . . . . . 5 23 5.5 Meteorological instrumentation . . .

. . . . . . . 5 25 Revision 11

1hltD -X- ...

TABLE Of CONT [NTS (Continued)

Page 6.0 RADI0 ACTIVE WASTE TRrATHENT SYSTEMS. EFFLUENT PATHWAYS, AND RADIATION MONITORS . . . ..................... 61 6.1 tiquid Radioactive Waste Treatment . . . . . . . . . . . . . 61 6.2 Gaseous Radioactive Waste Treatment . . . . . . . . . . . . . 63

. 6.3 Liquid and Gaseous Effluent Streams. Radiation Monitors, and Radioactive Waste Treatment Systems . . . . . . . . . . . . . 65 6.4 In Plant Liquid Effluent Pathways . . . . . . . . . . . . . . 6-5

, 6.5 In-Plant Gaseous Effluent Pathways . . . . . . . . . . . . . 66 7.0 REPORTING REQUIREMENTS . ........ . . . . . . . . . . . . . 71

- 7 .1 Annual Radiological Environmental Operating Report . . . . . 71

,l 7;2 Semiannual Radioactive Effluent P.elease Report . . . . . . . 72 7.3 Major Changes to Liquid and Gaseous Radioactive Waste Treatment Systems . . . . . . . . . . . . . . . . . . . . . . 74 7.4 Special Reports . . . . . . . . . . . . . . . . . . . . . . . 7-5 REFERENCES ..

. . . . ....... . . . . . . . . . . . . . . R1 Appendix A - Disposal of Septage . . . . . . . . . . . . . . . . A1 Appendix B - Concentrationr, in Air and Water Above Natural Background (10CFR20.1-20.602. Appendix 0) . . . . . . B1 Revision 11 mun -xi- ~

y (IST OF CONTR0ls Control Title Page 1.1 Applicability of Controls and Survelliance Requirements 1-2 1.2 Applicability of Controls and Surveillance Requirements 12 l

2.1 Off-Site Concentrations 2-1 ,

3.1 Dose Duc to Radioactive Liquid Effluents 3-1 3.2 Total Dose Due to Radioactive Liquid and Gaseous Effluents 32

3.3 Dose Rate Due to Radioactive Gaseous Effluents 34 3.4 Dose Due to Noble Gases Released in Radioactive Gaseous Effluents 3 10 l 3.5 Dose Due to Tritium, and Radionuclides in Particulate Form With Half-Lives Greater than Eight Days 3 11 4.1 Monitoring Program 41 c.

!qlgg) 4.2 Land Use Census 4-11 4

4.3 Intercomparison Program 4 13 5.1 Radioactive Liquid Effluents 5-1 5.2 Radioactive Gaseous Effluents 57 5.5 Meteorological Instrumentation 5 26 6.1 Liquid Radioactive Weste Treatment 6-1 6.2 Gaseous Radioactive Waste Treatment 6-2 7.1 Annual Radiological Environmental Operating Report 7-1 .

l 7.2 Semiannual Ef fluent Release Repc~t 7-2 7.3 Major Changes to liquid and Gaseous Radioactive Waste Treatment Systems 7-4 7.4 Special Reports 75 Revision 11 anura -xii-

i llST OF TABLES Tabl e Title Pace 1.1 -Summary of Concentration and Setpoint Methods, and Method 1 Dose Equations for Normal Operations at the Yankee Plant 1-3 1.2- Dose Factors Specific to the Yankee Plant for Noble Gas Releases 16 a 1.3 Summary of-Radiological Effluent Controls and Implementing Equations 17 1.4 Summary of Constants 19 1.5 Summary of Variables 1 10 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 Flant for l Tritium and Particulate G6;eous 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 Liqaid Effluents at Yankee Rowe (Derived from Reference A) 3 3.3 Age-Specific Usage Factors for Various Liquid Pathways at Yankee Rowe 3-18 3.4 Age Specific Usage Factors

, 3 38 ;

3.5 Environmental Parameters for Gaseous Effluents at the

, Yankee Plant 3 39 l 3.6- Yankee Nuclear Power Station Five-Year Average Atmospheric Olspersion Factors 3-44 4.1 Radiological Environmental Monitoring Program 44 4.2 Reporting' Levels for Radioactivity Concentrations in E~/ironmental Samples 4-6 Revision 11 onm -xiii-

i

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llST OF TAlilES (Continued)

Table Title Page 4.3 Detection Capabilities for Environmental Sample Analysis 47 4.4 Radiological Environmental Monitoring Stations 4 13 5,1 Radioactive Liquid Effluent Monitoring Instrumentation 5-3 5.2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 5-7 5.3 Radioactive Gaseous Effluent Monitoring Instrumentation 5-12 5.4 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements 5 14 l 5.5 Meteorological Monitoring Instrumentation 5-26 l 5.6 Meteorological Monitoring Instrumentation Surveillance Requirements 5 27 f- , .

o Revision 11 intua -xiv- "

llST OF FIGilRES-1 Table Title Page 11 Yankee Atomic Electric Company SITE BOUNDARY LINES 1 20 12 Yankee Atomic Electric Company Effluent Disc'arge Points.

-Site Plot Plan 1 21 w 4-1 Yankee Plant Radiological Environmental Monitoring Locations Within 1 Mile (Airborne, Waterborne, and Ingestion Pathways) 4 15

42 Yankee Plant Radiological Environmental Monitoring Locations Withiri 12 Miles (Airborne, Naterborne, and Ingestion Pathways) 4 16 4 Yankee Plant Radiological Environmental Monitoring Locations Outside 12 Miles (Airborne, Waterborne, and Ingestion Pathways) 4 17 4-4 Yankee-Plant Radiological Environmental Monitoring Locations at

-the Restricted Area Fence (Direct Radiation Pathway) 4 18 j 4-5 Yankee Plant Radiological Environmental Monitoring Locations Within 1 Mile (Direct Radiation Pathway) 4-19 46 Yankee Plant Radiological Environmental Munitoring Locations Within 12 Miles (Direct Radiation Pathway) 4 20 47 Yankee Plant Radiological Environmental Monitoring Locations Outside 12 Miles (Direct Radiation Pithv1y) 4 21 6-1 . Liquid Effluent Streams, Radiation Monitors, and Radioactive Waste Treatment System at the Yankee Plant 67 6-2 Gaseous Effluent Streams, Radiation Monitors, and Radioactive Waste Treatment System at the Yankee Plant 68 Revision 11 nron -xv-

i

' 1. 0 INTRODUCTION ,

According to Definition of Terms (Table 1.6), the OFF SITE DOSE CALCULATION MANUAL (ODCH) contains the methodology and parameters used-in the calculation of off site doses resulting from radioactive gaseous and. liquid effluents, in the calculation of-gaseous and 11ould effluent monitoring alarm / trip setpoints.-and in the conduct of the Radiological Environmental

-Monitoring-Program. The ODCM also conta_ ins: (1) the Radioactive Effluent i .< Controls.and Radiological Environmental Monitoring Program required by-l_ _ l Section 6.7 of the Technical Specification document and (2) descriptions of-l- .the information that should be included in the Annual Radiological

' 7 ] - Environmental Operating _ and-Semiannual 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 operation

-which are used to show compliance with the numerical guides-for design.

Controls of bection-11 Appendix 1, 10CFR Part 50.

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

All changes to the ODCH shall be reviewed and approved by the Plant'

-Operation: Review Committee (PORC) in accordance with Technical Specification l- L 6.13 prior to implementation. Changes made to the ODCM shall be submitted to l the_ Commission for their information in tne Semiannual Radioactive' Effluent Release Report for the period _in_which the_ change (s) was made effective.-

1.1 Summary of Methods. Dose Factors. timits. 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 reao Chapters 2 through 5. The concentration and setpoint methods.are_ documented in Table 1.1, as well as the

-Method.1 dese equations. Where more accurate dose calculations are needed.

-use the' Method 11 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 l -constants, variables. 'special definitions, and FREQUENCY NOTATION used in the

' Revision 11 mmo 1-1 l

.. -.. = _-- - - - . ._ - . -- . - . -

l ODCH are in Tables 1.4. 1.5, 1.6, and 1.10, respectively. Lastly, F,1gures 1-1

{' ")

and 1-2 depict the Yanket plant site boundary line and liquid effluent discharge points, respectively, 1.E Applicability of Controls and Surveillance Reauirements ISR)

Control 1.1 The controls and ACTION requirements shall be applicable during

-l conditions specified for each control .

Control 1.2 Adherence to the requirements of the controls and/or 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.

l l SR 1.1 Surveillance requirements shall be applicable during the conditions

_ l 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 4

@ surveillance interval, and

b. A total maximum combined interval time for any three consecutive surveillance intervals not to exceed 3.25 times the specified surveillance interval.

SR 1.3 Performance of a surveillance requirement within the specified time

, interval shall constitute compliance v:ith OPERABILITY requirements for a control and associated ACTION <tatements unless otherwise required by the control. -

I 1

1 Revision 11 uruzo 1-2

TABlf 1,1 Summary of Concentration and Setpoint Methods and Method i Dose Eauations for Normal Operations at the Yankee Plant Equation Maximum Equation (*)

No.

2-1 Unrestricted Area Total ENG Fraction of MPC in Liquids. F ING Ci Except Noble Gases - { t MPC 3 22 Unrest.ricted Area. Concentration of Noble Gases in Liquids CNG - E C i"O g

31 Total Body Dose Due to Liquids Dtb (mrem) - K E Og DFL.itb t

32 Maximum Organ Dose Due to liquids Dorgan (mrem) - K J' 0; DFL,,o l 3-3 Tot P y Dose Rate Due to

' mrem '

- 7.83 D Kr -85 DF8 Kr 85

) r >

l 34 Skin Dose Rate Due to Noble ' mrem' "

Gases Ski" Kr 85 DFK 'r 85 yr ,

l 3-5 Critical Organ Dose Rate Duc ' 'mr em ' ~

l H 3 and Particulates with Oc 'C T% > 8 Days

yr '

l 3-6 Gamma Air Dose Due to Noble Gases .)y,, (mrad) = 0.25 OKr 85 DFlr 85 l 3-6.1 Gamma Air Dose Due to Ground DT level Noble Gas Releases 9fd (mrad) - (6.0 x 10-6) (0 Kr'85) l 3-7 Beta Air Dose Due to Noble Gases 0 E

. Datr (mrad) = 0.76 Okr-85 DF Kr 85

] 3-8 Critical Drgan Dose Due to H-3 and Particulates with Oco (mrem) - E 0 3DFG ico T4 > 8 Days Revision 11 usuro l'3 "

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

r 8

J.iBL E 1.1 (Continued)

Summary of Concentration and SetDoint Methods, and Method 1 Dose fauations for Normal Operations at the Yankee Plant "

Equation Maximum E'uation(8)

No.

l 51 Liquid Release Rate Reading r f'2 R (MPCc ) (St)

I ,

<l> ,

l 53 Gaseous Release Rate Reading for (500) (60) (Sgr.as)

Total Body Dose Limit Rtb "

(f) (7.83) DFB %.

l 54 Gaseous Release Rate Reading for Skin Dose Limit (3000) (60) (S Kr 85)

Rsk "

(F) DF Kr-as Note (a):

Ci - Concentration of radionuclide "i" in a mixture (pCi/ml).

F = Primary vent stack flow rate (cc/ min).

I

- Concentration of radionuclide "i", evcept noble gases, at the point C fNG of discharge.

C ,"' - Concentration of radionuclide "i", excep+ noble gases, at the point of discharge.

D F ," -_ Skin dose factor for radionuclide "1".

OF[ - Gamma dose factor to air for radionuclide "i".

DFf - Beta dose factor to air for radionuclide "i".

i Revision 11 urpio 14

TABlf 1.1 (Continued)

Summary of Concentration and SetDoint Methods, and Method 1 Dose Ecuations for Normal 0Derations at the Yankee Plant 0F8 i = Total body dose factor for radionuclide *i".

DFG ic, - Site-specific, critical organ dose factor for a gaseous release of radionuclide "1".

. OFG i 'co- Site-specific critical organ dose rate factor for a gaseous release l

of radionuclide "i".

l DFL t u) - Site specific, total body dose factor for a liquid release of radionuclide "i".

Dfli ,, = Site-specific, maximum organ dose factor for a 11guld release of radionuclide "1".

f 3 - Flow rate past the test tank .aanitor (gpm).

f 2 = Flow rate at the point cf discharge (gpm).

K = Deerfield River flow rate correction f actor.

MPC c - Composite MPC for the mix of radionuclides (pCi/ml).

}' C,

=

(Eq. 5-2)

C

})i MPC, 0, - Total release (Curies) for radionuclide "i".

l

. Qi - Release rate (pC1/sec) for r.:Minnuclide "1".

lSKr 85 - Gaseous instrumentation response factor for Kr-85 (cpm /(pCi/cc)).

St - Liquid instrumentation response factor (cpm /(pC1/cc)).

I Revision 11 urun 1-5 "

. _ - \

< TABtf i,2~

% ' ' ")

Dose Factors Specific to the Yankee Plant for Noble Gas Releases Gammt.

Total Body Beta Skin Combined Skin Beta Air Gamma Air Dose factor Dose factor Dose factor Oose f6ctor Dose factor mrem-m 3 3

OfS, ~mrn-m mrem-sec ' mrad-m3 mrad-m3 OfBi ~

DF, Ofp Dfl ,

pC1 -yr , , pf 6 -y r , C 1 -y r , pCi-fr , pC1 -yr j Radionuclide Kr 85 1.61 x IO'S 1.34 x 10'3 3.22 x 10 2 1.95 x 10'3 1.72 x 10'5 l

4 Revision 11 mun 1-6

1 1

l TARlE 1,3 Summary of Radioloaical Effluent Controls pnd Implementino Ecuations Control Cateoory Method limit 2.1 Off Site Total Fraction of Eq. 2 1 s 1.0 Concentrations MPC Excluding Noble of liquids Gases Total Noble Gas Eq. 2 2 Concentration s 2.00 x 10 4 pCi cc 3.1 Dose Oue to Total Body Dose Eq. 3 1 Liquid 5 1.5 mrem in a atr.

s 3.0 mrem in a_yr.

Effluents Organ Dose Eq. 3 2 s 5.0 mrem in a-qtr.

s 10.0 mrem in a yr.

3.2 Total Dose Oue- Total Body Dose Eq. 3 1 s 25.0 mrem in a yr.

to Liquid and Eq. 3-6 Gaseous Eq. 3 9 Effluents Organ Dose- Eq. 3 2 s 25.0 mrem in a yr.

Eq. 3 8

) Eq. 3 9 Thyroid Dose Eq. 3-2 s 75.0 mrem in a yr.

Eq. 3 8 Eq. 3 9 3.3 Dore Rate Due Total Body Dose- Eq. 3-3 mrem to Gaseous Rate Due to Noble s 500.0 Effluents Gases Skin Dose Rate Due Eq. 3 4 mrem to Noble Gases s 3000.0 yr Organ'00se Rate Due Eq. 3 5 mrem

'll to H-3 and s 1500.0 Particulates with TW > 8 Days 3.4 Dose Due.to Gamma Air Dose Due Eq. 3-6 s 5.C .arad in a qt r.

Noble Gases in to Noble Gases s 10.0 mrad in a yr.

f nts Beta . Air Dose Due Eq. 3 7 s 10.0 mrad in a qtr.

to Noble Gases s 20.0 mrad in a yr.

Revision 11 airuro 1-7

l l

1 3

TABlf 1.3 (Continued)

Summary of Radiolooical Ef fluent Controls

. a.,nd lmolementina Ecuations Control Catecory Method limit 3.5 Dose Due to Organ Dose Due to Eq. 3 8 s /.5 mrem in a qtr.

l Tritium and H 3 and s 15.0 mrem in a yr. .

Particulates in Particulates with ;

Gaseous 1% > 8 Days Ef fluents e 5.1 Liquid Effluent Alarm / Trip Setpoint Eq. 5 1 Control 2.1 Monittr Setpoint 5.2 Gaseous Alarm Setpoint for Eq. 5 3 _ Control 3.3.a (Total Effluent Total Body Dose Body)

Monitor Rate

P " Alarm Setpoint for Eq. 5 4 Control 3.3.a (Skin)

Skin Oose Rate 6.1 Liquid Total Body Dose Eq. 3 1 s 0.06 mrem in a mo.

4a, Radioactive Waste Treatment Organ Dose Eq. 3 2 s 0.2 mrem in a mo.

6.2 Gaseous Gamma Air Oose Due Eq. 3 6 s 0.4 mrad in a mo.

Radioactive to Noble Gases as'.e heatment Beta /ir Dose Due Eq. 3-7 s 0.4 arad in a mo.

to Noble Gases Organ Dose Due to Eq. 3 8 s 0.3 mrem in a mn. ;

e

' l H 3 and Particulates with T% > 8 Days O

P,evision 11 1-8 a m iro -

i TABLE 1.4

,$ummary of Constants gn' tant Definition Units 0*

l - (3.17 x 100) {pCi-yr) [X/037 (sec/m3) 01 sec L

, " J .L ) ? 5 1C*D (7.83 x 10'6) Ci yr Ci-m3

- (3.17 x 10*d) (:pCi -yr ) [X/0] (sec/m3 )

Cl-sec.

- (3.17 x 10+4) (2.39 x 10'5) pCi-yr Ci m 3 1.11 - Average ra'. O of tissue to air energy absorption ratic coefficient.

l. 7.83 - (10+6) (pCi/pCi) (7.83 x 10 6) (sec/m3 ) pCi sec pC1 m 3 8.69 -

( 1.1.' ) (Sr) [X/0'.17 (sec/m 3

) (1.00 x 10+6) (pCi/pci)

- (1.11; (1.00) (7.,83 x 10 6) (1.00 x 10+6) pCi-sec pCi-m3-l 23.90 - (1.00 x 10+6) (X/0) pCl sec pCi-m 3

- (1.00 x 1046) (2.39 x 10'S)

, 60.00 = Conversion factor. sec min

. 500.00 - Total body annual dose l'ait from ICRP-2. mrem 3000.00 - Skin annual dote limit from ICRP-2. mrem j '3.17 x 10+4 - Number of picocuries per Curie divided by pC1-yr the number of seconds per year Ci-sec Revision 11 druro 19

P 3 ,TASlE 1.5 2'mmarv of Variables Variable Definition- Units 1.00 x 10+6 -

Number of picocuries per microcurie. pCi pC1

=

(1.0 x 10 6) ' Ci ' *7) sec' 'Ci-sec ' "

,pCi,(3.154 x 10 , yr , pCi -y r ,

NG l C =

Total concentration of all dissolved and Ci

entrained noble gases from all station sources. cc C1 NG =

Concentration of radionuclide "1.". except noble C1 gases. at the point of discharge. cc

~

C, =

Concentration of radionuclide "i". pC1/m3 or pCi/cc p

D atr

=

Beta dose to air. mrad

- Gamma dose to air, mrad Dh Dy = Gamma dose to air from a ground level release, mrad grd D,e = Dose to the critical (rgan, mrem D organ - Dose to the maximum orgate, mrem D 3ggn - Beta and gamma doset 'o the skin, mrem Dtb- - Dose to the total body. mrem l DFB Kr-85 - Total body gamma dose factor for Kr-85.

mrem-m 3 pC1-yr l DFS Kr 85 - Beta skin dose factor for Kr-85. mrem-m 3 pCi-yr

- = Combined site-specific skin dose factor for mrem-sec DF Kr es l Kr 85- pCi yr Revision 11 unizo '1-10 -

l.

TABLE 1.5 (Continued)

Summary of Variables Variable Definition Units

) 0Fy. - Gamma air dose factor for Kr-85.

Kr 85 mrad-m3 pC1 yr l s DF Kr-85 eta air dose factor for Kr-85. mrad m3 pC1 yr DFG ic , = Critical organ gaseous dose factor for mrem radionuclide "i". Ci DFG ico

- Critical organ gaseous dose rate factor for mrem sec radionuclide *1".

pCi-yr l

DFli ,, = Maximum organ liquid dose factor for radionuclide mrem

i",

ct 0Fl itb - Total-body liquid dose factor for radionuclide mrem i) '1".

C1 l g"" - Critical organ dose rate due to tritium and mrem l particulates-yr g - Skin dose rate due to noble gases. mrem yr g - Total body dose rate due to noole gases. mrem yr

'l D/0 - Deposition factor for dry deposition of sec l particulates.

m2

-l F- - Primary vent stack flow rate.

cc sec I

Revision 11 attuto 1-11

I

_ - _ _ a

TABLE 1.5

(' ,

(Continued)

Summary of Variables Variable Definition Units F

3 - Total-fraction of MPC in liquid pathw6ys, p ENG " Iotal fraction of MPC in liquid pathways excluding 1

noble gases. -

l f 3 - Flow rate past the test tank monitor. gpm l- f 2 = Flow rate at the point of discharge ,

gpm

MPC c - Composite MPC for the mix of radionuclides. See ci Eqnation 5 2.

cc MPCg - Maximum permissible concentration of radionuclide Ci "i" (10CFR Part 20. Appendix B. Table 2. Column 2, cc l see Appendix B of the 00CM).

Oj = Release for radionuclide "i". Ci l g - Total r: lease rate of noble gas (Kr 85). pC1 sed l g - Release rate for radionuclioe "1". Ci sec X/0 - Average undepleted dispersi an fcctor, sec m3

[X/0]D = Average depleted dispersion factor, seC m3

[X/0F - Effective average gamma dispersion factor. sec m2 Sp- = Shielding factor.

Revision 11 anuto 1-12 '

h TABLE 1.5

)

(Continued)

Summary of Variables Variable Definition Units l S gr.g5 - Gaseous monitor response factor for Kr-85. cpm pC1/cc

l S,- - Liquid monitor' response factor. cpm pCi/cc l

L.

l.

Revision 11 n:uro 1-13 "

TABLE 1,6 xf -}

-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 CAllBRATION A CHANNEL CAllBRATION 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, lhe CHANNEL CALIBRATION shall encompass the entire channel, including the alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CAllBRATION may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK G

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

CHANNEL FUNCTIONAL TES_T_

A CHANNEL FUNCTIONAL TEST shell 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, I -

FRE0VENCY NOTATION the FREQUENCY NOTATION specified for the performance of surveillance -

l- l requirements snall correspond to the intervals defined in -iable 1.9.

I MEMBER (S) 0F THE PUBllC L

l MEMBER (S) 0F THE PUBLIC (for purposes of 10CFR50. Appendix I) shall include all persons who are not occupationally associated with the plant. This Cdtegory 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 urnro 1 14 "

l

TABLE 1.6

)

(Continued)

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

OFF-SITE DOSE CALCULATION MANUAL (00CM)

The 00CM contains the methodology _ and parameters used in the calculation of off-site doses resulting fron radioactive gaseous and liquid effluents, in the

=.

calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in, the conduct of _the Environmental Radiological Monitoring Program. The ODCH. also contains (1) the Radioactive Effluent Controls and Radiological l l Environmental Monitoring Programs required by Section 6.7 of the Technical Specification documer.t and (2) descriptions of the information that should be

-included in the. Annual Radiological Environmental Operating and Semiannual Radioactive- Ef fluent Release Reports required by Controls 7.1 and 7.2, respectively.

l OPERABLE - OPERABillTY 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 te 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 nerform its function (s) are also capable of. performing their related suppo. function (s).

I-SITE BOUNDARY The SITE SOUNDARY shall be that line oeyond which the land is not owned, leased, or 'otherwise controlled by the licensee. Any area within the SITE

, BOUNDARY used for re'sidential quarters or recreational purposes shall be

. considered to belbeyond the SITE B0UNDARY 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.

I Revision 11 anun 1-15 "

__________d

l_

s TABLE 1.6 (Continu3d)

Definition of Terms

. VENTILATION EXHAUST TREATMENT SYSTFM A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce radioactive material in particulate form in gaseous ef fluents by pass;ng ventilation or vent exhaust gases through HEPA filters for the purpose l

) of removing particulates from the gaseous exhaust stream prior to release to

the environment. Such a system is not considered to have any effect on noble l l gas effluents.

l l

Revision 11 neuro 1-16 ,

I

b g TABLE 1.7

^

Dose factors Specific to the-Yankee Plant for liauid Releases T

Total Body Dose Maximum Organ Factor Dose factor

, '0Fl rem' mrem'

,' itb Ci DFLt .o Radionuclide ( , Ci ,

1

H3 5.99 x 10'd 5.99 x 10'4 C 14 1.64 x 10+0 8.18 x 10+0

.l 4 Mn 6.07 x 10-2 5.47 x 10'l fe 55 3.46 x '10 2 2.11 x 10'l Co 58 4.76 x 10-2 1.81 x 10'l Co 60 2.79 x 10'l 9.04 x 10'l

'k Zn-65 1.65 x 10+0 2.71 x 10+0 Sr 89 2.30 x 10'l 8.04 x 10+0 Sr-90 6.97 x +1 2. 75 - x '10+2 Zr-95/Nb 95 1.40 x 10'3 2.87 x.10'l i

Ag 110m- 2.32 x 10 2 2.21 x 10+0 Sb 124 2.62 x 10-2 6.'48 x 104

.[

Cs-134 1.79 x 1041 2.40 x 10+1

.l.

. .Cs-137. 1. 0 7 . x '.0+1' 2,07 x 10+l 1' Ce-144 1.41 x 10'3 2.58 x 10+0 Revision 11 mun 1517 ~'.

=-

-TABLE 1.8

.s Dose and- 00se' Rate Factors Specific to the Yankee Plant for l Tritium and Particulate Gaseous Releases Critical Organ Critical Organ 00se Factor Dose Rate Factor DFG ico mrem' N Ci , , D F G,"c, .

l Radionuclide < pCi-yr ,

H-3 7.21 x 10'3 2.27 x 10'l ,

C-14 4.38 x.10+0 1.38 x 10+2 l

Mn-54 3.78 x 10+0 1.49 x 10+2 l'

Co-58 1.98 x 10+0 7.06 x 10+1 Co-60 4.08 x 10+1 1.81 x 10+3 Zn-65 1.99 x 10+1 6'. 4 3 x 10+2 Sr-89 6.10'x 10+1 1.92 x 10+3

@ Sr-90 2.36 x 10+3 7.44 x 10+4 Zr-95/Nb-95 3.77 x 10+0 1.24 x 10+2 1

Ag-110m 3.63 x 10+1 1.22 x 10*3 Sb-124 6.95 x 10+0 2.32 x 10+2 l

Cs 134 8.52 x 10+1 2.83 x 10+3 I

Cs-137 8.71 x 10+1 2.97~x 10+3 -

l Ce-144 2.10 x 10+1 6.65 x 10+2 Revision 11 nuun 1-18

. . . .. . . - - - -~

~ TABLE-1.9 Frecuency Notation Notation Frecuency

-S' At least once per 12- hours.

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

W At least once per 7 days.

. M At least once per 31 days.

O At least once per 92= days.

SA At least once per 184 days.

R At-least once per 18 months.

l P Prior to e3ch release.

N.A. Not applicable, h

j Revision 11 auuzo 1-19 "'

1 i

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1-20

. . ...i --m_ .

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. g !

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s'

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.**%.- l j 45 $p% Pn.

mary Liquid ' s, s , , .. - j :/ ,7

% Effluent Discharge . N :l 1*-

." i<- '

r*irn . ;.# ,[ ' #

g , . ~ , - /s=::5 :s

\, i \' \ Y., l T-nt << #

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(Secondary Discharge Point) f,

'<,--,3 -

[,, /

.y ,f , / , ,.. .:-

og l ./ [7 .i

/ . p// O ,.

\ ,Prirmry Vent ,i p ,

Sud

/\ i

\ ,*

- \ o v

\ i '

\ p '

1

% ~

FIGURE 1 ./'#

YANKEE NUCLEARPOWER STATION EFFLUENT DISCHARGEPODTTS

/ y-SITE PLOT PLAN /

Revision 1I

2.0 RADI0 ACTIVE L10VID EFFLUENTS I-2.1 Off-Site Concentrations l Control 2,1 In accordance with Yankee Technical Specification 6.7.5.a.,

l itens 2 and 3, the concentration of radioactive material releasec to l Unrestricted Areas (see Figure 1-2) shall be limited to the concentrations sp;cified in 10CFR Part 20 Appendix B. Table 11. Column 2. for radionuclides Cther than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2.00 x 104 pC1/ml total activity.

Applicability At all times.

w ACTION l With the concentration of radioactive material reicased from the site to j Unrestricted Areas exceeding the above limits, without delay, take actions to restore the concentratica to within the above limits.

{

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

SR 2.1.2 The results of radioactive analysis shall be used in accordance with the methods of the 00CH to assure that concentrations at the point of release are maintained within the limits of Control 2.1.

Bases

.l- Control 2.1 is provided to ensure that the at any time concentration of l radicactive materials released in liquid waste effluents from the site above l 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.

l Column 2 (Appendix B of the ODCM contains a listing of these values Ls taken

-l from the regulations). These requirements provide operational flexibility.

Revision 11 orure 2-1 .,

(T l- compatible with considerations of health and safety, which may temporarily

( l ~ result in releases higher than the absolute value of the concentration numbers j in Appendix B, but still within the annual average limitation of the revised l (January 1, 1993 effective date) 10CFR. Ptrt 20, regulation. Compliance-with l the design objective doses of Section II.A of Appendix ! to 10cFR, Part 50 l assure that doses are maintained ALARA, and that annual concentration limits l of Appendix B to 10CFR20.1001-20.2401 will not be exceeded.

I l The concentration limit for noble gases is based upon the assumption =

l that Xe-135 is the controlling radionuclide and that an effluent concentration l in air (submersion dose equal to 500 mrem /hr) was converted to an equivalent l concentration in water. '

47? -

M e

Revision 11 suruo 2-2 ~

j

TABLE 2.1 Radioactive liould Waste Sampling and Analysis Program Lower Limit Minimum Type of of Detection Sampling Analysis Activity LLD(*)

Liquid Release Type Frequency Frequency Analysis (pCi/ml)

A, Batch Waste P P Principal 5.00 x 10'7 Release TanksO) Gamma Each Batch Each Batch Emitters")

, Dissolved and Entrained Gases 1.00 x 10'S (Gamma Emitters)

P M Tritium 1.00 x 10 5 Each Batch Composite (*)

Gross Alpha 1.00 x 10'7 P 0 Sr-89, Sr-90 5.00 x 10-8 Each Batch Composite (*) pe.55 1.00 x 10-6 3 B. Plant Continuous y Principal 5.00 x 10'7 rb e uilding Continuous (d) Composite (d)

Sump ers "I g Tritium 1.00 U0-5 Continuous (d) Composite (d) hu WM 1R x M'7 0 Sr-89, Sr-90 5.00 x 10 8 Continuous (d) Composite (d)

Fe-55 1.00 x 10~8 C. Principal PlantCogtinuous M/2 5.00 x 10'7

Releases Gama Continuous (d) Composite (a)

Emitters {t)

-Auxiliary Service M Tritium 1.00 x 10'S Water Continuous (d) Composi te(d) gu gg 1g x g7 0 Sr-89, Sr-90 5.00 x 10'8 Continuous (d) Composite (') pe.55 1.00 x 10

Revision 11 anuzo 2-3 ,

4~ s.

-l

. l l

1 2

, TABLE 2.1 'j

[ -(Continued)

":Notatioq:

- +

{ La . The LLD is defined-in Table Notation (a) of Table 4.3:of SR:4.1.

!:: -b. A1 batch release is the' discharge of liquid' wastes of a discrete

-volume. Prior to sampling for. analysis, each batch shall:be

-.t . isolated and thoroughly * ' *.ed to assure representative sampling,

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 specimenLwhich is representative of the. liquids released,

d. Prior to analyses, all ' samples taken: forf the composite. shall 'be thoroughly mixed in order for the composite-sample to be I

- 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.

~

l input flow during periods when flow exist through the system. For n

the auxiliary service: water,: continuous composite sampling is- not L C ' required when;. liquid waste test tank. discharges are being .

conducted and/or the SFP cooling pumps are tripped.

Ef. The principal gamn.a' emitters for which. the LLD requirement applies exclusively are the: following adionuclides: Hn-54,-Co-58, Co 60,

.Zn-65. Cs-134'.JCs-137, and Ce-144. .This list does not mean that

- only these radionuclides are to be detected and reported.- Other peaks thats are measurable and . identifiable, together with the above radionuclides.- also shall be identified and reported.

~

- Radionuclides-that- are oeiow thF LLD for the . analyses should not be reported as treing present at the LLD level.-

Revision 11

- mano 2-4

]

L L

2.2 Method to Calculate Off-Site Liouid Concentrations

)-

The basis for plant procedures,that_the plant operator requires to meet Control 2.1, which limits the total fraction of MFC (Ff"0) in liquid pathways l

l (excluding noble gases) at__the point of discharge (see ODCM Figure 6-1) is I

discussed. (Ff"G) is limited to less than or equal to one, i .e. ,

Ci J- 13r l i MPC, The total noble gas concentration, CNG, is limited to 2.00 x 10 C1/ml.

NG Evaluation of (Ff"0) and C is required Concurrent with the sampling l and analysis prograc in Table 2.1 of Control 2.1.

Determine the _ total fractica of MPC in liquid pathways -(excluding noble gases) as follows:

)) Ff"6 -Ei MPC i

(Eq.2-1)

MPC, - Maximum permissible concentration of radionuclide "i" except for dissolved and entra'ned noble gases (10CFR Part 20.

-l Appendix B,-Table 2. Column 2. See Appendix B of ODCH for l listing).

ll Determine the total noble gas concentration such that the following l condition is met:

NG C

'l 12 2 E -4

l. C"G - Total concentration of dissolved and entrained noble gases from l all-. station sources (uCi/mi).

Revision 11-ataito 2-5 --

_a

f'( T.I Where:

l Ci - C fI + C "5 + C f""

3 (Eq.2-2)

C[T - Concentration at the point of discharge of radionuclide "1" except for dissolved and entrained noble gases; from the test tank.

ss Ci - Concentration at the point cf discharge of radionucli(.e "i"

l except for dissolved and entrained noble gases from the l Auxiliary Service Water System.

Cf* " = Concentration at the point of discharge of radionuclide "i" l except for dissolved and entrained noble gases from any other l significant sources which may be created during plant l decommissioning activitier

( l QYa - 2.3 Method to Determine Radionuclide Concentration for Each Liouid Effluent Pathway 2.3.1- Tes' Tank Pathway CfT 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.

l. Most periodic batch releases are from the two 5000-gallon capacity test l tanks. When test tanks are filled with liquid waste, they are isolated for .

l sampling and release. The volume of the tank's contents are determined from l the liquid level in each tank. A chemist extracts a s ..,,ie for radionuclide l' 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 l representative of the sample. Composite samples are analyzed at a minimum for l tritium and gross alpha activity. At a minimum, each test tank batch is Revision 11 anzuza 2-6 "

i

'l analyzed for .dissoived and entrained noble gases, and principal gamma l emitters.

I lI 2.3.2 Service-Water Systems Pathway 5

Cg"! is determined for each radionuclide above the analytical LLD from

.l the activity in composite samples from the effluent lines of the Auxiliary

.l Service Water System downstream of any potential inleakage source.

l .2.3.3 - Remainino Pathways other Ci -is determined for each of the remaining pathways as follows:

Miscellaneous batch releases of potentially contaminated water.

1.e.,

l rain water collected in the moat of the inservice radioactive waste-l tank, are treated according to Section 2.3.1.

A proportional cumposite of the Turbine Building floor drains is continuously collected, Since the water in this pathway is practically all Sherman Pond water (>99 percent) used for cooling purposes, it is unlikely

-that it will ever be a significant effluent pathway.

Revision 11 minno 2-7 ~'

r

?

l-

' 3 ~. 0 : DOSE / DOSE-RATE CONTROLS AND CALCULATIONS. ,

~

i; 13.ih Dose Due toiR adioact ve' tiauid: Effluents

- l Contr'o l 3,l' In accordance withiYankee:Tedhnical. Specification _6.7.5.a',

-items 14 and-5, the dose ot doseicommitment to a HEMBER OF:THE PUBLICEfrom. ,

radioactive. mates, 's in liquid; effluents-released from thersite (see

Figure lla2);to avap able uptake: pathwaysishall be limited:

-a. =During any_ calendar ~ quarter: :less than or equal to:1.5 mrem to '

.thehtotal body and_less than or equal to 5 mrem to any organ ;and r.

l. i b. During any1 calendar year: less than or equal to 3 mrem to :the -

-total body and less than ortequal:to 10 mrem to'any organ.

! ADDlicabilitY Atalht'mes.

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,J 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 t'he corrective actions takenito reduce the releases and the proposed-corrective! actions to be-taken to' assure:that' 1 subsequent releases will be within the above' limits.

h'

l;

,X .

Surveillance Reauirement' SR-3.1 EDose Calculations' - Cumulative dose contributions from _1.iquid

! effluents shall be determined in accordance with.the ODCH a' least once_per

- L31 days.

-Basesi 7

=Controli3.1. is provide'd to? implement the requirements of Sections ll.A.

~

_Ill.A; and IV.A of Appendix !. 10CFR Part 50. The' control implements the Revision'll t oturo __ 3-1 ",

. - - ~ . . . - . -

+ -

1

. guides set forth:in-'Section Il.A'.= The ACTION statements pr_ ovide the required i j 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.

l The' surveillance requirement implements the requirements in Section Ill. A of-Appendix I;that conformance with the_ guides of Appendix I be shown by I' calculational. procedures based on models and data such that the actual l

exposure _ of a MEMBER 0F T!:E PUBLIC through appropriate- pathways is unlikely to be-substantially underestimated. Existing pathways of liquid exposure to - '

{ MEMBER (S) 0F-THE PUBLIC which form the basis for calculating liquid _ doses _-_ in the:0DCM ~ are described in detail in Yankee Atomic Electric Company's design. ,

. report, " Supplemental Information -for the Purpose of' Evaluation of_- 10CFR

.Part_SO, Appendix'1", dated June 2.-1976 (with amendments). The_ point of exposure from.existi.ng pathways for dose calculational purposes is taken l downstream of Sherman' Dam in' the Deerf 0 eld River. The equations specified in

[ 'the ODCM for. calculating-the doses-due:to the actual release rates of g (radioactive materials in liquid ef fluents were developed from the methodology _

[ provided in Regulatory Guide-1.109, " Calculation of Annual Doses to Man from j Routine Releases of Reactor Effl"ents for the Purpose of Evaluating Compliance !

-with 10CFR Part-50, Append 1.x_I," Revision 1, October 1977,_and Regulatory ____

. Guide .1.113. " Estimating Aquatic Dispersion of Effluents from Accidental and i i

Routine- Reactor Releases for the Purpose of Implementing Appendix I." April-

[ :1977, Also, there is reasonable assurance that the operation of the facility i

l will not result.in radionuclide concentrations in finished drinking-water that

are in excess of: the requirements of 40CFR141. No drinking water supplies from~ the Deerfield River below-the plant have been identified.

l 1

1 p

Revision-11 i

- 32 4

3.2 Total Dose Due to Radioactive Liauid and Gaseous Effluents l ' Control 3.2 In accordance with Yankee Technical Specification 6.7,5.a.10, the dose or dore 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 th; thyroid, which is limited to less than or equal to 75 mrem) over a calendar year.

Applicability At all times.

t 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 'imits of Control 3.2 ..ve been exceeded. If such 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 l 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 if the release condition resulting in violation of 40CFR 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.

I Revision 11 mmo 3-3 "

7 Surveillahce Reautrement 4

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 dosc 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 1. 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 l been corrected) in accordance with the provisions of 40CFR Part 190.11, is considered to be a timely request and fulfills the requirements of 40CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40CFP 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

'l gaseous effluent controls.

Revision 11 nauze 3-4 -

y ,

q

'h.

Ur f3;3-_ Dose Rate Due'to:

- i Radioactive Gaseous Effluents--

.Ll[ Control 3.3 < Inlaccordance with Yankee Technical Specification 6.7.5.a.

items ~3 and 7 : the dose rate"due to? radioactive materials-released _ in gaseous y

effluents from the site to areasiat _ and,beyond the .' ? TE BOUNDARY (see

Figure.1-1);shall be limited to the following:

l T a: . For.. noble gases (Kr-85): less than or-equal to 500. mrem /yr-to the-

total- body. and 1ess than or--equal to 3,000 mrem /yr to the skin,

.-~

and

, l b. 1For tritium, and radionuclides in particulate. form with half-lives -

greater.' _than- 8 ' days: .less than or equal to 1,500 mrem /yr to any-

-organ.

E ' Applicability-

.At all:t'imes.

~ FACTION' With the Ldose: rate (s) exceeding the above limits, without delay, take actions to decrease the release rate to within the above limit (s).

Surveillance Reauirements ,

SR 3.3.1 The dose rate due to noble gases. In gaseous effluents shall be determined to be'within'the above' limits in accordance'with the methods and

_proceduresofEtheODCH.-

~

- l L SR ' 313. 2 The dose rate due to tritium, and 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 L the 0DCM. by; obtaining representative.-samples and performing analyses in

- accordance with the sampling and analysis program specified:in Table 3.1.

I i Bases-l} Theispecified limits as ~ determined by the Inethodology ir the ODCM, l-jrestrict .at all times, the corresponding gamma. and beta dose rates above i

! Revision:11 4

niun 3-5 e

' d

g-- l background to a member- of the public at or beyond the site boundary to (500) i s l mrem / year to the total body or to (3,000) mrem / year to the skin. This l- instantaneous dose rate limit allows for operational flexibility when of f

_ l normal occurrences _ may temporarily increase gaseous effluent release rates l from the plant, while still providing controls to ensure that licensee meets l the dose objectives of Appendix 1 to 10CFR50.

I l Control 3.3 also restricts, at all times. comparable with the length of l the sampling periods of Table 3.1 the corresponding maximum organ dose rate l above background to 1500 mrem / year for the highest impacted receptor to the l plant.

l Revision 11 Ril\170 3-6

TABLE 3.1 Radioactive Gaseous Waste Samolino and Analysis Procram Gaseous Release Type Sampling Minimum Analysis Type of Activity LLD Frequency Frequency Analysis Ci /ml at ) -

l A. Plant Vent (Primary- M. M Principal Gamma Emitters tb> 1.00 x 10

Vent Stack) Grab Sample Tritium 1.00 x 10~6 i Continuous (d) WIC) Principal Gamma Emitters tb) . 1.00 x 10'11 l Particulate Gross Alpha l Continuous (d) .

O Sr-89. Sr-90 1.00.x 10'11 Composite Particulate Sample l Continuous (8) Noble Gas Noble Gases 'I . 00 x 10-5 Monitor Gross Beta or Gamma ;

Revision 11 muro 3-7 i

ry TABLE 3.1 l

- fg i .

(Continued) 2 Table Notation 1

a. The LLO is defined 'in-Tablet N otation (a) of Table 4.1 of-Control 4.1- .

~

- b . ;. :The principal gamma- emitters _for which the LLO control ~ applies ll fexclusively are the following radionuclides: Kr-85,for gaseous

~

l. emissions and Mn-54. Co-58..Co 60, Zn-65. Cs 134. Cs-137, and Ce-144 for particulate emissions. Th!s list-does not mean that only these radionuclides are to be detected and reported. Other peaks which are .

measurable er.d identifiable, together with the above radionuclides, also shall be identified and reported. Radionuclides which are below the LLD

'for the analyses should not be ported as being present at' the LLD-level =-for that radionuclide.

l -' '

.l~-c - Samples shall be changed = a't least once per 7 days, and analyses shall be

? l= completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing or after removal' from samplers.

L=.i

. l Ld. -The-ratio of the sample flow rate to the sampled stream flow rate shall '

be known.for the1 time period covered by'each dose or dose rate:

T (f~ calculation made in accordance with Controls 3.3, 3.4, and 3.5.

l s

Revision 11 airuto 3-8 .,

o

l 3.4 _ Oose Oue- to Nctle Gases Released in Radioactive Gav gus Effluents l Control 3.4 In accordance'with Yankee Tecnnical Specification 6.7.5.a,-

items 5 and 8 the air dose due to_ noble gases released in gaseous effluents from the site to areas at and beyond th? SITE B0UNDARY (see Figure 1-1) shall be limited- to the following:

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

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

ApolicabilitY i

At all times.

ACTION

a. - With the calculated air dose from radioactive noble gases 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 Repcrt which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases _and the proposed corrective actions to be taken to assure that subsequent releases will be within the above limits.

I Surveillance Reauirement SR 3.4 Doce Calculations - Cumulative dose contributions for current calendar quarter and current calendar year shall be determined in accordance with the

_, ODCH at least once evary 31 days. i fpses Control 3.4 is provided to implement the requirements of Sections II.B.

Ill.A and IV.A of Appendix 1, 10CFR Part 50. The control implements the guides set forth in Section II.B. The ACTION statements provide the requirtd Revision 11 anuzo 39 .,

(TS operating flexibility and at _the same time implement the guides set forth in

\- Section IV.A to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable." The surveillance requirement--implements the requirements in Section III. A of Appendix l- that conformance with the guides of Appendix 1 be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways 'is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous ,

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 1," Revision 1. October 1977 -and Regulatory Guide 1.111. " Methods for Estimating Atmospheric _ Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1. July 1977.

The ODCM provides for determining the air doses at the SITE B0UNDARY based upon the historical average atmospheric conditions.

G 6

I i

Revision 11 mmo 3-10

3.5- Dose Due to Tritium and Radionuclides in Particulate Form With Half-Lives Greater than Eight Days l Control 3.5 -In accordance with Yankee Technical Specification 6.7.5.a.

.l items 5 and 9. the dose to a MEMBER OF THE PUBLIC from tritium, and radionuclides in particulate form with half-lives greater than 8 days in .

gaseous effluents released from the site to areas at and beyond the SITE i

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

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

Applicability At all times.

ACTION hj a. With the calculated dose from the release of radioactive materials.

in pa.ticulate form, or radionuclides other than noble gases 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.

I Surveillance Reouirement

, SR 3.5 Dose Calculations - Cumulati';e dose contributions for the current l ' calendar quarter and current calendcr year for tritium, and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the ODCM at least once every 31 days.

Revision 11 uruze 3-11

-~

s

Bases "

Control 3.5 is provided to implement the requirements of Sections II.C.

lil.A. and-IV.A of Appendix 1, 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 1 to assure that the releases of radioactive materials 'in gaseous effluents will be kept "as low as is reasonably a chi evabl e . " 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 HEMBER OF THE PUBLIC through appropriate pathways is -

unlikely to be substantially underestimated. The equations specified in the ODCH for calculating the doses due to the actual release rates of the subject materials were developed using the methodology provided in Regulatory Guide 1.100 " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix 1," Revision 1, October 1977, and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1. July 1977.

These equations also provide for determining the actual doses based upon the

() historical average atmospheric conditions. The release rate specifications l _for tritium, and 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 gree, leafy vegetation with subsequent consumption by man. (3) deposition onto grassy areas where milk and meat animals graze with consump, tion of the milk and meat by man, and (4) deposition on the ground with oubsequent exposure of man.

Revision 11 mum 3-12

n ;

!3.6 Dose Calculation Concepts I

Controls 3.l _ through 3.5 either limit dose or dose- rate. The term-

" dose" for ingested _orginhaled radioactivity means the dose commitment.

teasured: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 doseLcommitment is?

evaluated is 50 years. The. phrases " annual dose" or " dose in one year" then

~ ,

refer to the fif ty; year dose commitment from one year's worth of releases.

"Doselin' a quarter" similarly means a fifty-year dose-commitment from one

_i quarter's releases. 'The-term " dose " with-respect to external exposures, such' l

as to noble gac clouds, refers only to the doses received during the' actual tice' period-of exposure to the radioactivity-released from the plant. Once p 1 the source of the radioactivity is removed,- there is no longer any additional L - accumulation to the- dose commitment, 4

t-

- The-_ quantities D and - C are introduced to provide calculable quantities.

Iz

-related to off-site dose or dose rate which demonstrates compliance with the controls.

The dose.:D. is the quantity calculated by the Chapter 3 dose equations, f

~ The'D calculated by " Method I" equations'is not necessarily the actual-dose !

received by a real individual, but usually provides an upper bound for a given release-because of the conservative margin built into_the dose factors and the:

zselection' and definition of the critical receptors. The radioisotope specific dose factors in each " Method I" 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 iatake-

, 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.

Ll Revision.11 mmo - 3-13

.- n

&T,,G~~T~~"~~""~^""~~""~~~"~~~~~~""~'^~~~~"

l l

I

3.7!'Hethod to Calculate the Total Body Dose from Liauid-Releases

'~ .

Control 3.1 limits the . total body dose commitment to a MEMBER OF THE:

i- 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 -

i: PUBLIC from all station sources-(including liquids) to 25 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.

- ;li -- , ;

}To evaluate; total body dose for Control 6.1 add the total' body dose from l _ today's expected releases to the total body dose accumulated for; the- time -

l: _ period of. interest.

I i j 3.7.1 Method-I -l

[ The total body ~ dose from~a'11guid release is:

D tb

- K.E Oi 0Fl i - .- i tt, (Eq. 3-1)

s. :

_( mrem). g j=

Where:

[ DFlitb ' Site-specific total body = dose factor (mrem /C1) for liquid release. See Table 1.7.

It 0'i .- Total activity (Curies) re' leased to liquids of radionuclide "1" "

during the period of interest. 'For i - Fe-55, Sr-89, Sr-90, or' H-3 use the best estimates (such as the most recent. ' '

measurements).

W .

K - 366/F o where Fe is the averege (typically monthly. average) dilution flow of the'0eerfield River below Sherman Dam (in-ft 3/sec). If F dcannot'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 ft 3/sec). .

Revision 11-mua 3-14 e

__m g> - n.- --

m n 2

i n ,

s J 'A

. Equation 3-1.can;be applied under the._ following conditions' (otherwise,-

I 5-justify Method!I or1 consider Method II):

~

[ ; a . -- : Liquid releases:to the; service water pathway;to Sherman Pond or to-the west? storm drain pathway to the Deerfield River,

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

. 3.-7.2 Method'Il

. 4 If Method 1 cannot_ be applied or if.. the' Method -I dose exceeds the 11mit'

,, _ or if a more exact. calculation is _ required, then Method 11 should be applied.

Method 11 consists of the models. input data, and assumptions in- Regulatory _

Guide 1.'109. Rev.-I (Reference A), except:where_ site-specific models, data. or

< assumptions are more applicable. _The base case analysis is aLgood example of -

the use of Method II. It is an acceptable starting point for_ a Method Il-analysis, i3 ..7 3' Basis-for Method I .

Method I may be-used to show that the controls which limit off-site

~

p total body dose from liquids (Controls .3.1, 3.2, and 6.1)'have been met _ for--

- releases.over the appropriate periods. These requirements arc based on. design objectives _ and standards in P)CFR Pai t 50 and .40CFR Part 190. Control 3.1 is-based on the ALARA design controls in 10CFR_ Part 50, Appendix I, Subsection -II

A. LControl .6.1 is an " appropriate' fraction", determined by the NRC, of the

ALARA designz control. ' Control 3.2.is based on Environmental Standards for the-

lJ-Uranium Fuel" Cycle in 40CFR Part 190 which applies to direct radiation as well as lilquidiand gast.ous effluents. - Method I- applies only to the liquid contribution.

l-

-Hethod I.was developed such that "the actual exposure of.an

!individua1 ... is unlikely'to be substantially underestimated (10CFR Part 50.

,.. Appendix.I). -The_ definition of a single " critical receptcr" (a hypothetical individual..whose-behavior results in an unrealistically-high dose) provides part of4 the conservative margin to the calculation of total body dose in

- MethodJI. Method 11 allows that actual-individuals with real behaviors be taken into; account for any given release. In' fact, Method I was based on a.

Methodsil_ analysis for the criti:al1 receptor and annual average conditions 11nstead'of any real individual. The analysis was called the " base case"; it Revision:11 wuo -3 215

.--w , ,,,, ,. m

7 was_then' reduced to form Hethod 1. The base case, the method of reduction.

-and-the assumptions and data used are presented.

h The steps performed in the Method I derivation follow. First, in the base case, the doso impact to the critical receptor (in the form of dose factors in mrem /C1) 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-16.

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 l above.

i for any liquid release during any period, the increment in annual average total body dose from radionuclide "1" is:

ADig - (0 ) (DFlitb) 3 where DFLtu, is the total body dose factor for radionuclide "i", and 0, is the activity of radionuclide "1" released in Curies.

Hethod I is more conservative than Method 11 because it is based cn the following reduction of the base case. The dose factors, DFltu>, used in Het5od I were chosen from th' 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 unm 3-16

~~

l

..= -.: *" >=

qi TABLE 3.2 , .

,, 4 .

Environmental Parameters for Liauid Effluents at Yankee' Rowe . l

~

2 (Derived-from Reference A)-

d-

, j <

Food Grown With Contaminated Water.' I 4 ,l.

]- .-Variable Aquatic' Shoreline: . .

' Leafy ~

Food ActivityL Vegetables ~' Veg .' ' Meat' Cor Mil k;

j- MP. Mixing Ratio")' 'O.84 0.84' O.84 0.84. l 0.84 -' O.84 [

,w<

j '. TP- Transit Time -(hrs) '24.00. 0.00- 0.00 -0.00 f480.00L 48.00 YV Agricultural -

2.00 2.00 2.00. 2.00 -i (jl Productivity (kg/m ) 2 l P Soil Surface Density '(kg/m 2) -

240.00 7240.00 240.00' 240.00l ,

l IRR-- Irrigation Rate (1/m2 /hr)'. - -

O.15 0.15 0.15 0.15 l TE Crop Exposure Time. .(hrs) - -

1440.00 1440.00 1440.00. 1440.00

l: TH Holdup' Time (hrs) - -

-1440.00 24.00 2160.00 2160.00 l OAW Water Uptake Rate * -

50.00 L60.00 .f l for Animal -- (1/d)

-t Feed Uptake Rate

~

1 OF - - "-

50.00 50.00 .

-l f.or Animal' (kg/d) l Location of. Critical Individual lB610w Below Below 'Below Below Below f Sherman Sherman Sherman Sherman Sherman - Sherman f Dam Dam Dam Dam Dam Dam - j l!

)

+

[

") Listed mixing ratios appl to Method I' dose factors. Method II analyses can appi river flow and plant discfarge dilution flow which exist over the period of actuaf calculated I release. mixing ratios: based on= :

Revir. ion 11 ;

I ,

anura 3-17 1

, m O!

,. TABlf 3.3

.I

Ace Sriecific (Isace Factors 'or Various licuid Pathways at Yankee Rowe (from Reference A. Table E-5. Zero where no pathway exists)

Leafy Potable Age Veg. Veg. Hilk Heat fish invert. Water Shoreline Group (kg/yr) (kglyr) (1/yr) (kg/yr) (kg/yr) (kg/yr) (1/yr) (hr/yr)

Adult -520.00 64.00 310.00 110.00 21.00 0.00 0.00 12.00 ,

Teer 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 iQ.

Revision-11 nryro _ 3 ..

3.8 J.lethod to Calculate Maximum Oroan Dose f rom tiauid Releases Control 3.1 limits the maximum organ dose commitment to a MEMBER Of 1HE PUBLIC from radioactive material in liquid effluents to 5 mrem per quarter and 10 mrem per year. Control 6.1 rquires liquid radioactive waste treatment when the niaxi.num organ dose estimate exceeds 0.2 ihrem in any 31 day period.

Control 3.2 limits the maximum organ dose commitment to any real MEMBER Of THE PUBLIC from ell station sources (including liquids) to 25 mrem in a year except for tht thyroid, which is limited to 75 mrem in a year. Dose e

evaluation is rtquired 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 orgar, dose accumulated for the time period of interest.

3.8.1 Me t h od._J.

The maximum organ dose from a liquid release is:

} Dorga - K E 0, Of Li .,

1 (Eq. 3-2)

(mrem)

Where:

DF Li ,3 = Site-specific maximum organ lose factor (mrem /Ci) for a liquid release. See Table 1.7.

Oi - Total activity (Curies) released to liquids of radionuclide "1" during the period of interest. For i - fe-55, Sr-89, Sr-90, or H-3, use the best estimates (such as the most recent

, measurements).

K

~ 366/fd : where f dis the average (typically monthly average) dilution flow of the Deerfield River below Sherman Dam (in f t3 /ses). If f cannot o be obtained or f ois greater than 366 K can be &ssu.ned to equal 1.0. The value. 366. is the ten-year Revision 11 attuto 3-19

c. - - - . . . . - - - . . - - - - - -

l

,-- minimum monthly average Deerfield River flow rate below Sherman Dam (in ft3 /sec).

Equation 3 2 can be applied urder the following conditions (otherwise, justify Method I or consider Method 11):

l a. Liquid releases to Sherman Pond or to the west storm drain pathway to the Deerfield River, i

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

3.8.2 Method 11 ,

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 11 should be applied. Method 11 consists of the models, input data, and assumptions in l Regulatory Guide 1.109. Revision 1 (Reference a), except where site-specific 1

models, data, or assamptions are more applicable. The base case analysis is a

good example of the use of Method 11. It is an acceptable starting point for i a Method 11 analysis.

() 3.8.3 Basis for Method I

.l 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:

AD,,, = (0j ) ( OFL,,o) where DFL ,, is the maximum organ dose factor for radionuclide g *1", and Oi i s the activity of rad?onuclide "1" released in Curies.

The dose fc. tors. OFtj,o, used in Method I were chosen from the base case to be the t,1ghest ef .the sat 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 11 emua 3-20 t

,, ,~.m--, w ,,n s v - - - - -e>- - ~ - ~~a----

3.9 Method to Calculate the Total Body Dose Rate from Noble Gases Control 3.3 limits the dose rate at any time to t'ie total body from l noble gas at any location at or beyond the SITE BOUNDARY equal to or less than 500 mrem / year.

l l Compliance with the dose rate limits fof noble gases is continuously demonstrated when effluent release rates are below the plant ent stack noble gas activity monitor alarm setpoint by virtue of the fact that the alarm setpoint is based on a value which corresponds to the off site dose rate limit of Control 3.3 or a value below it.

Determinations of dose rates for compliance with Control 3.3 are performed when the effluent monitor alarm setpoint is er.ceeded, and the corrective action required by Control 3.3 is unsuccessful, or as required by the ACTION to Table 5.3 when the stack noble gas monitor is inoperable.

3.9.1 Method 1 The total body dose-rate due to noble gases can be determined as f ollows:

)

Du, " rem'

- 7.83 6ge.85 0FBxr 85 (Eq. 3-3) yr j Where:

l 0xr 85 - The release rate from the plant vent stack (pC1/sec) of Kr-85.

The release rate at the stack also can be stated in the

'following equation:

f 3 D = (H)

I

. (f) br 85 s (Eq. 3-10)

'pCi' ,gcp ,) fpti/cc ' ' cc '

sec, cpm ,,sec, Revision 'I airuro 3 21

M i

-- Where: l I

H -

Plant vent stack monitor count rate (cpm).

I l S Kr es Gaseous monitor response factor for Kr 85 (cpm /(pC1/cc)).

f -

Plant vent stack flow rate (cc/sec).

l l OfB ge.g3 - Total body dose factor for Kr 85. See Table 1.2.

Equation 3 3 can be applied under the following conditions (otherwise, )

justify Method I or consider Method !!): . l i

l .a. Normal conditions. -

l b. Kr-85 gas releases via the plant vent stack to the atmosphere.

l 3.9.2 Method ll ;

^

If Hethod I cannot be applied, or if the Method I dose exceeds the r limit, or if a more exact calculation is required, then Method 11 may be applied. Method 11 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 11 analysis. 1 3.9.3 Basis for Method I l

Method I may be used to show that the Control 3.3 limit for the total body dose rate from noble gases released to the atmosphere has been met for the peak noble gas release rate.

I Method I was derived from Regulatory Guide 1.109 as follows: .

O T - (3.17 x 10*4) (X/0) (S )F E Og OfB :'

1 Revision 11 amua 3~22

'f e

, -,m-,.,n ..._.,n.4-n.- -- ._ ,. g. ,,. - , ,

lhe equation was derived by combining Equations B 4 and B S from Regulatory I

Guide 1.109, assuming X/0 - X/0D for noble gases, and some simplification in the notation. T Assuming that Dginite - DI [X/0]T/[X/0) and that i

Di e - Orgnit,

0 (pci/sec) 31.54/0 (C1/yr), we get:

l Otb (mrem /yr) - (1.0 x 10 6) (Sr) [X/0)Y o rr 85 DFBrr.es Substituting:

-- 1.00 (shielding factor).

Sr

[X/0]Y - Long term average gamma dilution factor.

7.83 x 10 6 (sec/m3 ).

l O Kr es -

Release rate of Kr-85 (pC1/sec).

i Gives:

l 013 (mrem /yr) - 7.83 o ,,.85 r 0FBgr.g$ (Eq. 3 3)

Revision 11 nruro 3-23

3.10 Met hod to Calculat,e t he skin Dose Rat e f rom Noble Gases g t,

l Control 3.3 limits the dose rate at any time to the skin f rom noble l gases at locations at or beyond the SITE BOUNDARY to 3,000 mrem / year.

I Compliance with the dose rate limits for noble gases is continuously demonstrated when effluent release rates are below the plant vent stack noble gas activity monitor aiarm setpoint by virtue of the fact that the alarm

~

setpoint is based on a value which corresponds to the off site dose rate limit or a value below it.

Determinations of dose rate for compliance with Control 3.3 are performed when the effluent monitor alarm setpoint is exceeded, and the corrective ACTION required by Control 3.3 is unsuccessful, or as required by the notations to Table 5.3 of Control 5.2 when the stack noble gas monitor is inoperable.

3.10.1 Method 1 The skin dose rate due to noble gases (Kr-85) is:

Ol l Dskin (mrem /yr) = OKr-85 0Fx,.85 (Eq. 3 4)

Where: ,

6 Kr 85 -

The release rate from the plant vent stack (pCi/sec) of l _Kr-85. The release rate at the stack also can be stated in the following equation:

e 8 Q-(M) I (f) b

l Kr-85,

'pC1 '

=(cpm) 'pCi/cccc '

,sec, cpm , sec, Revision 11 mun 3 24

h Wherei: !

M = Plant vent stack monitor count' rate (cpm)..

l S Kr 85

' Gaseous monitor response factor for Kr 85'(cpm /(pC1/cc)). ;

F -Plant vent' stack flow rate (cc/sec).

l -- j

.l - Df'e.85 t - Combined skin dose-factor for Kr 85. See Table 1.2.- !

. Equation 3 4 can be applied under the following conditions (otherwise.

, justify Method 1. or consider Method. II):

l l a. Norma 1' conditions _(not. emergency event). I

.b. Noble gas releases via the plant vent-stack to the atmosphere. -

3.10.7 Method 11 If Method I cannot be applied, or if the Method I dose exceeds the- l L . limit, or. if a more_ exact calculation is_ required, then Method 11 may be applied. Method 11. consists ofJthe 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 11. It is_an acceptable starting = point for

-a Method.Il analysis. <

3.10.3:'BasisforMethod} f l: The methods to calculate the skin dose rate parallel the' total body-dose ;

rat'e methods'in Section 3.9.3. Only the differences'are presented here.

l i

. Method I was derived from Regulatory Guide-1.109 as follows:

i

D 3 . (3.17_ x 10") [(x/0). (1.11)l(Sr) E 0 DFJ + (X/0) E Oi 0FSi ] 1 1 t

4 i

Revifion 11 i

enuro 3-25 ..

~

The equation was derived by corrbining Equations B 4. B-5, and B-7 from Regulatory Guide 1.109, assuming that X/0 - X/0D for noble gases, and making some simplifications in notation. Assuming that 0/ininte - DT [X/0]T/[X/0) and l that b$ tin -D 5 +

0 (pci/sec) -

31.54/0 (Ci/yr) yields, and that Kr 85 is l the only noble gas left in plant inventory:

D$ g,n (mrem /yr) - (1.11) (Sr) (1.00 x 10*') [X/0]7 ogr.85 Dfir85 -

+ (1.00 X 10+6 (X/0) Drr.85 Ofstr 85 Where:

[X/0]7 - 7.83 x 10 6 sec/m3 .

X/0 - 2.39 x 10 5 sec/m3 .

Sr - 1.00 (shielding factor).

Substituting gives:

) 056tn (mrem /yr) - 8.69 ogr.85 DFlr 85 + D Kr 85 0fSxr+s5 23.9 x 10+1

= D Kr 85 [8.69 Offr 85 + 23.9 DFS Kr 853 Define:

OfK 'r 85 - 8.69 (Of 7) + 23.9 (DFSgr.85)

= 8.69 (1.72 x 10'5) + 23.9 (1.34 x 10'3)

' = 3.22 x 10 2 ' mrem - sec ' ,

pCi - yr ,

i Then:

l Osk$n (mrem /yr) = Oge.85 DFKr 85

Drr 85 3.22 x 10 2 (Eq 3 4)

Revision 11 ainuo 3-26 ,

j 3.11 Method to Calculate the Critical Oroan Oose Rate from Tritium and i Particulates with Half lives' Greater Than fioht Days l Control 3.3 limits the dose rate at any time at location at or beyond l the Site Boundary from H 3, and radionuclides in particulate form with half lives greater than eight days to 1,500 mrem / year to any organ. The peak l release rate averaging time in the case of particulates is commensurate with l the time the particulate samplers are in service between changeouts.

I 3.11.1 Method I

. The critical organ dose rate can be determined as follows:

Dc ,- E oi DFG ico ,

i

' mrem ' , 'pci ' ' mrem sec ' 9'

, yr , ,sec, pC1 yr ,

Where:

)

0, -

Stack activity release rate determination of radionuclide l "1" (tritium, and particulates with half-lives greater than eight days) in pC1/sec. For 1 - Sr-89, Sr-90, or H-3, use the best estimates (such as most recent measurements).

OFG[co -

Site-specific critical organ dose rate factor mrem sec'

, pCi - yr ,

for a gaseous release. See Table 1.8.

Equation 3-5 can be applied under the following conditions (otherwise.

).

justify Method I lor consider Method II):

l a. Normal conditions (not emergency event).

4 l b. Tritium and particulate releases via the plant vent stack to the atmosphere.

1 Revision 11-l eituzo 3-27 ,,

e 3.11.2 Method 11 k!

If Hethod I cannot be _ applied, or if the Method I dose exceeds the control limit, or if a more exact calculation is required, then Method 11 may be applied. Method 11 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 11. It is an acceptable starting point for l a Method 11 analysis.

l 3.11.3 Basis for Method 1 l The methods to calculate the critical organ dose rate parallel the total body dose rate methods in Section 3.9.3. Only the differences are presented here.

Method I may be used to show that Control 3.3.b which limits organ dose l rate from tritium and radionuclides in particulate f orm with half lives l greater than eight days released to the atmosphere has been met for the peak tritium and particulate release rates.

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

D co - E OsDFGico 1

~ 'Ci ' ' mrem ' (Eq. 3-8)

[ C1 ,

Applying the conversion f actor, 31.54 (Ci sec/pci yr), and converting 0 to 6 in pCi/sec yields:

oco - 31.54 E D, DfG,co i

' mrem ' ,, 'Ci- sec ' 'pCi ' ' mrem '

yr , pC1 yr, sec, C1 Revision 11 mmo 3 28 1

Equation 3 5 is rewritten in the form:

oco- E 0, orG,'co i

Where:

, DFG co - (DFGico) (31.54)

' mrem-sec ' , ' mrem ' 'Ci-sec '

y pC1-yr , C1

, pCi yr, Should Method 11 be needed, the analysis for critical receptor critical l 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.

h.

Revision 11 emuo 3-29 ,,

g l 3.12 Method to Calculate the Gamma Air Dose from Noble Gases (Kr-85) s Control 3.4 limits the gamma dose to air from noble gases at any location at or beyond the SITE BOUNDARY to 5 mrad in any quarter and 10 mrad in any year. 00se evaluation is required at least once per 31 days.

I 3.12.1 Method 1 4 ,

The gamma air dose from plant vent

tack releases is: ,

Dltr(mrad) = 0.25 Ogr.85 Offr85 (E9

3'f) . ,

Where: -

l O Kr 85 - Total Kr-85 (Curies) released to the atmosphere via the plant l vent-stock during the period of interest.

i l Of[r es - Gamma air dose factor for Kr-85. See Table 1.2.

() Equation 3 6 can be applied under the following conditions (otherwise, justify Method I or consider Hethod 11):

l a. Normal conditions (not emergency event).

4

b. Noble gas releases via the plant vent stack to the atmosphere.

3.12.1.1 Ground tevel Releases for ground level releases, the garoma air dose is: .

0,Tre (mrad) - (6.0 x 10 6) (Orr es) (Eq. 3 6.1) ,

d Where:

l OKr 85 = The ground level release (in curies) of Kr-85.

Revision 11 neuro 3-30 ..

3.12.2, Method 11 if Hethod I cannot be applied, or if the Nethod I dose exceeds the I limit, or if a more exact calculation is required, then Method 11 may be applied. Method 11 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 ex6mple of the use of Hethod 11. It is an acceptable starting point for a Hethod !! analysis.

3.12.3 Basis for Method I

+l Method I may be used to show that Control 3.4, which limits the off-site gamma air dose from gaseous effluents, has been met for releases o'!er appropriate periods. Control 3.4 is based on 10CFR Part 50, Appendix 1, Subsection B.1, which limits the estimated annual gamma air dose at unrestricted area locations.

l l $ With Kr-85 being the only noble gas potentially available for release, l the dose can be taken from Equations B 4 and B 5 of Regulatory Guide 1.109 '

with the added assumption that 0/inite - D [X/0]Y/[X/0):

- 3 D

'PCi-yr ' 3 mrad-m3 1r (mrad) = 3.17 x 10+4 [X/0]7 (sec/m ) Orr.es (C1) 0Flr 85 Ci-sec, pCi yr ,

Where:

[X/0]7 - Long-term average gamma dilution factor.

- 7.83 x 10 6 (sec/m3 ).

l 0 gr.g5 - Number of Curies of noble gas (Kr-85) released.

l l

l Revision 11 I

oruro 3 31 I

,e, Which 1 cads tos (l

0 ir (mrad) - 0.25 O rt 85 0 Fir 85 (Eq. 3 6) l The gamma air dose from a ground level release is determined by using the same Regulatory Guide 1.109 equation to derive Equation 3 6. The only l -differences are l

[X/0)Y - 1.10 x 10 5 sec/m3 , which is the long term average ground level [X/0)Ybased on the time period from May 1977 through April 1982. +

l l .0Flres - the gamma air dose factor for Kr 85 (see Table 1.s) f i mrad m3 l - to 1.72 x 10 5 pC1-yr ,

Substituting the above into the Regulatory Guide 1.109 general equation gives:

44 pCi-yr' [X/0]Y (sec/m 3) DFy erad m3 0lrd (mrad) - 317 x 10 0 (C1)

~C1-sec ~ ~pCi-yr '

(Eq. 3-6.1.'

= 3.17 x 10+4 x 1.10 x 10-6 x 1,72 x 10 5 x 0xr 85

- 6.0 x 10-6 OKr 85 (C1) e

Revision 11 airuro . 3-32 .,

1 3.13 Method to Calculate t he Beta Air Oose fiom Noble Gases Control 3.4 limits the beta dose to air from noble gases at any location at or beyond the SITE B0UNDARY to 10 t,irad in any quarter and 20 mrad in any year. Dose evaluation is required at least once per 31 days.

I 3.13.1 Method 1 The beta air dose from plant vent stack releases is:

Oftr (mrad) = 0.76 Ogr.as DFfr es (Eq. 3-7)

, Where l Offr 85 = Beta air dose factor for Kr-85. See Table 1.2.

l 0 xr 85 - Total Kr-85 (Curies) released to the atmosphere via the plant vent stack during the period of-interest.

Equation 3-7 can be applied under the following conditions (otherwise.

justify Method I or consider Method 11):

l a. Normal conditions f.not emergency event),

b. Noble gas releases via the plant vent stack to the atmosphere.

3.13.1.1 Ground level Releases For ground level released, the beta air dose can be determined by using

'l Equation 3-7 for stack releases. Equation 3 7 results in doses that are approximately ten percent more conservative than calculating releases using

. ground level methodology, 3.13.2 Method 11 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 11 may be applied. Method 11 consists of the models, input data, and assumptions in

-Revision 11-nuun 3-33

l, . Regulatory Guide 1.109, Revision 1 (Reference A), except where site specific k_, ) models, data, or assumptions are more applicable. The base case analysis is a good example of the use of Hethod 11. It is an acceptable starting point for a Method 11 analysis.

3.13.3 Basis for Method 1 l The methods to calculate the beta air dose parallel the gamma air dose methods in Section 3.12.3. Only the differences are presented here. ,

Method I may be used to show that Control 3.4, which limits the off site beta air dose from gaseous effluents, has been met for releases over .

appropriate periods. Control 3.4 is based on 10CFR Part 50, Appendix 1 Subsection B.1, which limits the estimated annual beta air dose at

, unrestricted area locations.

I l With Kr 85 being the only noble gas potentially available for release.

l the dose can be taken from Equations 6-4 and B 5 of Regulatory Guide 1.109:

Ofir (mrad) - (3.17 x 10d) [X/0) OKr 85 Of Kr 85 O Substituting:

X/0 - 2.39 x 10'5 sec/m 3 We have E

l Dftr (mrad) = 0.76 O Kr 85 DF Kr 85 (Eq. 3-7)

Revision 11 ausuo 3-34

_ _ _ _ _ _ ___.m. . _ - - _ - - _ _

3.14 Nethod to Calculate the Critical Orqan Oose from Tritium and Particulates Control 3.5 limits the critical organ dose to a HENBER OF THE PUBLIC l from radioactive tritium and 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 HEMBER OF THE PUBLIC frcm 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.

l 3.14.1 Method 1 The critical organ dose from a gaseous release is:

(mrem) = { 0 DFGico (Eq. 3 8)

Where:

, 0g - Total activity (Curies) released to the atmosphere of

) radionuclide "1" during the period of interest. For i - Sr-89, Sr 90, or H-3, use the best estimates (such as the most recent measurements).

OFG ic , = Site-specific critical organ dose factor (mrem /C1) for a gaseous release. See Table 1.8.

Equation 3 8 can be applied under the following conditions (otherwise, justify Hethod I or consider Method !!):

'{ a. Normal conditions (not emergency event),

,l b. Tritium and particulate releases via the plant vent stack to the atmosphere,

c. /sny continuous or batch release over any time period.

Revision 11 muro 3-35

i i

3.14.2 Method 11 h t

[

if Method I cannot be applied. or if the Nethod 1 dose exceeds the ;

limit, or if a more exact calculation is required, then Method 11 should be applied. Method ll consists of'the models,. input data, and assumptions in ;

Regulatory Guide 1.109. Revision 1 (Reference A), except where site specific models, d9ta, or assumptions are more applicable.- The Dase case analysis, i documented below 'is a good example of the use of Method II. It is an l acceptable starting point for a-Method !! analysis. , j i

3.14.3 Basis for Method I l . !

Hethod I may be used to show that Controls 3.2 and 3.5, which limit !

off site organ dose from gases. have been met for releases over the '

l appropriate periods. Control 3.5 is based on' the ALARA requirements in 10CFR

{

Part 50. Appendix 1. Subsection 11 C. Control 3.2 is based on Environmental l standards-for Uranium Fuel Cycle in 40CFR190 which applies to direct radiation '

j . as well as to liquid and gaseous effluents. These methods apply only-to tritium-and p;rticulates in gaseous effluents.

1 Hethod 1 was developed such that "... the actual exposure of an i 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 ll allows that actual individuals with real behaviors be taken into '

account for any given release. In fact, Method I was based on a Method !!

analysis of the critical receptor for-the annual average conditions. For purposes of complying with Controls 3.2 and 3.5, annual average dilution ,

factors are appropriate for batch and continuous releases. The analysis was called the " base case": it was then reduced to form Method I. The base case. ,

the method of reduction, and the assumptions and data used are presented i below.- i The steps performed in the Method I Gerivation follow. First, in the . .

base case, the dose impact to the critical receptor in the form of dose !

l factors. OFG ico fmrem/C1), for a one Curie release of each tritium and [

. particulate radionuclide to gaseous effluents was derived. Then Method I was q determined using simplifying and further conservative assumptions. The base -l case analysis uses the methods, data, and assumptions in Regulatory l- Guide' l.109-(Equations C-2. C-4 and C-13 in Reference a). Tables 3.4 and 3.5

' Revision 11 i

mun 3-36

+

outline human consumption and 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 3.15.

l for stack gas releases during any period, the dose from radionuclide *i" is:

Dico - (DFGsco)(0j) where DFG,co is the critical dose factor for radionuclide "i", and 0, is t'ne activity of radionuclide "i" reieased in Curies.

).I i

l l

)

F 9

Revision 11

'32020 3-37

-_..___m_..__...

TABlt 3.4 '

i )

Ace-Specific Usaae Factqrl (from Regulatory Guide 1.109. Table E 5)

Leafy

/.ge Vegetables Vegetables Hilk Heat Inhalation l Group (kg/yr) (kg/yr) (1/yr) (kg/yr) (m3 /yr)

Adult 520.00 64.00 3*0.00 110.00 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 by /-

Revision 11 estuzo 3-38*

TABLE 3.5

.5,,ironmental Parameters for Gaseous Effluents at the Yankee Plant (Derived from Reference a)

Vegetables Cow Milk Goat Milk

M2at variabic Stored Leafy Pasture Stored Pasture Stored Patture Stored YV Agricultural 2 (kg/m ) 2.00 2.00 0.70 2.00 0.70 2.00 0.70 Productivity 2.00 P Soil Surface (kg/m 2) 240.00 240.00 240.00 240.00 240.00 240.00 Density 240.00 240.00 T Transport Time (hrs) - -

48.00 48.00 48.00 48.00 480.00 480.00 to User TB Soil Exposure (hrs) 131400.00 131400.00 131400.00 Time") 131400.00 131400.00 131400.00 131400.00 131400.00 l

TF Crop Exposure (hrs) 1440.00 1410.00 720.00 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 OF Animals Daily (kg/ day) - -

50.00 50.00 6.00 6.00 50.00 50.00 Feed FP Fraction of Year - -

0.50 -

l on Pasture (2)

FS Fraction Pasture - -

1.00 -

l When on Pasture (M FG Fraction of Stored 0.76 - - - - - - -

Veg. Grown in Garden Revision 11 nituto 3-39

.I

g

()i TABLE 5.5 (Continued)

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

Veoetables Cow Milk Goat Milk

Meet Va riabl e Stored Leafy Pasture Stored Pasture Stored Pasture Stored FL Fraction of Leafy -

1.00 - - - - - -

Veg. Grown in Garden Fi Fraction Elemental - - - - - - - -

Iodine - 0.5 lH AbsoluteMI (gm/m3 ) - - - - - - - -

Humidity - 5.6 oPathway is not included in Method I.

It is listed for informational purposes and the possible use in a Method II 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 2.433 hours 0.0145 weeks 0.00333 months (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 enimals are assumed to be on pasture is zero (nongrowing season). ror 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/m 3

) shall be used to reflect conditions in the Northeast (

Reference:

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

' Revision 11 mun 3-40 O 6 0 6 r -- -- ._v

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

3.16 Critical Receptors and lono-Term Averace Atmosoheric Olsoersion Factors for 1moortant Exposure Pathways The gaseous effluent dosc equations (Method I) have been simplified by assuming an individual whose behavior and living habits inevitably lead to a higher dos.: than anyone else. The following pathways of exposure to gaseous 1 l effluents as listed in Regulatory Guide 1.109 (Reference a) have been I considered. They are:

.- a.- Direct exposure to contaminated air.

b. Direct exposure to contaminated ground.

c. Inhalation of air.

d. Ingestion of vegetables.

e, -Ingestion of cow milk. and

f. Ingestion of meat.

Section.3.15.1 details the selection of importen,t off-site locations and receptors: Section 3.15.2 describes the atmospheric model used to convert meteorological data into dispersion factors: and Section 3.15.3 contoins the i resulting descriptions of the critical receptors and their dispersion factors as a function of exposure pathway.

7 3.15.1 Critical Receptors t

r i

.The most limiting SITE BOUNDARY location'in which individuals are or are

[ likely to be located was assumed to be the receptor for all the gaseous i , : pathways considered. This provides a conservative estimate of the dose to an j _ individual-from existing and potential gaseous pathways for the Method 1 j analysis. '

i This point is the SSE sector. 800 meters.

4 j Revision 11

.inute 3*4l '

{z o ;

i 4

m

. _ . . ~ ~ . . . _ . . _ ,. . _ . . , _ . _ . _ . _ . _ . . , , . , _ , . . _ . _ . , . _ . . . ._._.~.....m.. .._ .. -- .- .- ... -

3.15.2 Yankee Atmospheric Dispersion Model j

\

l The annual average dispersion factors are computed for routine j (long-term) releases using the lankee Atomic Electric Company's (YAEC) AEOLUS 1 l (Reference b) computer code.

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

a. X/0, nondepleted dispersion factors for evaluating ground level

concentrations;

~

b. [X/030 , depleted dispersion factors for evaluating ground level l concentrations of particulates;

c. X/07, effective gamma dispersion factors for evaluating gamma dose rates from a sector averaged finite cloud (multiple energy, -

undepleted source): and l d. D/0, deposition factors for dry deposition of particulates, lhe AE0LUS diffusion model is described in the AE0LUS manual l (Refereice b). AEOLUS is based, in part, on the straight line airflow model l as discussed in Regulatory Guide 1.111 (Reference c).

One difference is that the gamma dose rate is calculated througnout this ODCH using the finite cloud model presented in Meteorology and Atomic Energy l 1968 (Reference h. Section 7-5.2.5). That model is implemented through the l definition (Reference b. Section 6) of an effective gamma dispersion far,

X/07, and the replacement of X/0 in infinite cloud dose equations by the / .

l Another difference is that the relatively narrow valley in which the -

plant sits is considered by the model. 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 tic valley sectors, the contributions from the other six balley sectors are averaged into the particular valley receptor. This is done for distances greater than 500 meters from the primary vent stack where the valley effects are assumed to cause channelling.

Revision 11 urur: 3-42

_ _ _ ___ _ . . _ . . y _ ~ _

3.15.3 lono Term Averaoe Dispersion Factors for Critical Receptors Actual measured meteorological data for the five year perind, January 1981 through December 1985, was analyzed to determine tt '; cations of the aaximum off site atmospheric dispersion factors. Each dose and dose rate calculation incorporates the maximum applicable off-site, long-term average attospheric dispersion factor. The values used and their locations are su marized in Table 3.6.

e l

(

r Revision 11 1

3.c I

L TABLE 3.6 Yankee Nuclear Power Station Five-Year Average Atmospheric Dispersion Factors (II Dose to Critical Dose Rate to individual Dose to Air Oroan Total Body Skin Critical Organ Gamma Beta Thyroid l

sec 2 19 x IO'S - -

2.19 x 10~5 X/0 9epleted l

r 3 -

2.39-x 10~5 .- -

2.39 x 10 5 .

X/0 Undepleted sec u

e ' - -

5.02 x 10-8 - -

5.02 x 10'8 0/0

\* >

r ' 7.83 x 10-6 7.83 x 10~6 5'C 7.83 x 10~6 - -

X/07

.j"SSESITEBOUNDARY.800metersfromtheprimaryventstack.

Revision 11

.inm 3-44 a

, g p e

\

3.16 Metnod to Calrulate 'i, et Dose from Plant Operation Control 3.2 restricts the dose to the whole body and any organ of any l real HEMBER OF THE PUBLIC at and beyond the Site Boundary from all station

-l sources (including direct radiation) to the limit of 25 mrem in a year, except l for the thyroid which is limited to 75 crem in a year.

l Estimates of direct exposure above background in areas at and beyond the l site boundary (or in residential areas inside the site boundary) can be l determined from measurements made by environmental TLDs that are part of the

l Environmental Monitoring Program (see Table 4.4). Alternctively, direct dose l calculations from identified fixed sources on site can be used to estimate the

,) off-site direct dose contribution where TLD information may not be applicabic.

Revision:ll muino 3-45

~ . .-.

.0 "aol0 LOGICAL ENVIRONMENTAL MONITORING 4.1 Monitorino Prooram i

j fControl 4.1 in accordance with Yankee Technical Specification 6.7.5.b.1, the l

Radiological Environmental Monitoring Program shall be conducted as specified !

in Table 4.1.

=

Applicability At all times.

ACTION

a. With the Radiological Ervironmental Monitoring Program not being conducted as specified in Table 4.1, prepare and submit to the Commission in the Annual Radiciogical Environmental Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrente.

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.

11 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 envir'- .au' sampling media at one or more of'the locations specified ic ist ' 1 exceeding the reporting levels of Table 4.2 when averaged oser 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 ev-luation 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 4.2 are detected in the samplint medium, this report shall be submitted if:

P concentration (1) + concentration (2) + . . 2 1. 0

reporting le~el (1) reporting level (2)

Revision 11 oruro 4-1

^

r When radionuclides other than those in Tttie 4." are detected and

(/ arE the result of plant effluents, this repu 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.

l c. With milk samples no longer available from one or more sample

locations required by Tabic 4.1, identify new location (s), if available, for obtaining replacement samples and add them to the ,

Radiological Environmental Monitoring Program within 30 days. The specific location (s) from which samples were no longer available may then be deleted from the monitoring program. Pursuant to Control 7.2, identify the cause of the samples no longer being available and identify the new location (s) for obtaining available replacement samples in the next Semiannual Radioactive Effluent Release Report and include revir.ed ODCM figure (s) and table (s) reflecting the new location (s).

Sinveillance Reouirement 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) 0F THE PUBLIC resulting from the station operation. The monitoring program implementsSection IV.B.2 of -

Appendix I,10CFR Part 50, and thereby, supplements the Radiclogical 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 provided by the Revision 11 em 4-2 ,,

4

~

l 1

' Radiological Assessment Branch Technical Position on Environmental: Honitoring,=

_j -Resision 1. November 1979. ; Program changes may be initiated based-on

operational experience,

i. - - -

-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 Cefined as an a priori (before the fact) limit' E

= representing the capability of a measurement' system and not as'an a costeriori 4

-(after the' fact) limit for a particular measurement. This does not preclude-

the calculation of an~ a costeriori' LLD for a particular measurement _ based upon -

7 he-actual parameters for.the sample in question.

L~

E .

-t *

-. J

)

Revision 11

~

enture 4-3

TABL .1 Radiological Environmental Monitoring Progr?m Exposure Pathway Number of Sample Sampling and Type and Frequency and/or Sample Locations Collection Frequency of Analysis

1. AIRBORNE l a. Particulates 5 Continuous operation of sampler Gross beta radioactivity

~

with sample collection as fcilowing filter change.

required by dust loading, but at Composite (by location) for least once per two weeks. gamma isotopic at least once per l

quarter.

2. DIRECT 24*** Quarterly Gamma dose, at least once per l

RADIATION quarter.

l l

3. WATERBORNE

a. Surface 2 Composite sample ** collected over Gross beta and gamma isotopic l

j a period of one month at analysis of each sample.

downstream location: monthly Tritium analysis of composite l

grab sample at upstream control sample at least once per l

location. quarter.

l

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 l

Shoreline _ sample.

Revi.sion 11 4-4

nruro , ,

, , TABL .1 , ,

(Continued)

Radiolootcal Environ": ental Monitorino Program

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

.of Analysis t

4. INGESTION l a. Milk 2 At least once per month. Gamma isotopic analysis of each ;

sample.

b. Fish 2 Commercially and recreationally Gamma isotopic analysis on important species. Seasonal or edible portions.

semiannually. if not seasonal.

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

1. Tuberous vegetable

2. Above ground vegetable l 3. Fruit Specific sample locations for all media are specified in the ODCH and reported in the Annua 1' Radiological Environmental Operating Report.

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

I l Does not include Restricted Area Fence locations.

l l One sample from downstream area with existing or potential recreational value, t

Revision 11 Nzuro 4-5

m ,

@ 4..

TABLE 4.2 Reportina levels for Radioactivity Concentrations in Environmental Samples l

Water Airborne Fish Hilk . Food. Products l Analysis (pCi/l) Particulates (pC1/kg, wet) (pCi/1) (pCi/kg. wet) l (pC1/m3 )

H-3 3 x 10+4 - -

Mn-54 1 x 10+3 -

3 x 10** - -

Co-58 1 x 10+3 -

3 x 10+4 - -

l Co-60 3 x 10+2 -

1 x 10** - -

Zn-65 3 x 10+2 -

2 x 10+d - -

Zr-ftb-95 4 x 10+2 _ .

3 x 10+1 1 x 10+1 1 x 10+3 6 x 10+1 1 x'10+3 l Cs-134 5 x 10+1 2 x 10*1 2 x 10+3 7 x 10+1 2 x 10'3 Cs-137 1

Reporting levels for nondrinking water pathways.

Revision 11 .

oruro 4-6 O 9 g e Y

e-. .

qn/

-- i '( .g *-

' TABLE 4.3 _

'i

-Detection Capabilitiesfor Environmental Samole'Analysistanc) l Airborne-' -

Water Particulates ....s t Fish Milk Food Products.-- ' Sediment

-l' Analysis .W '(pCi/l) '(pC1/m3 ) .(pC1/kg, wet)

't (pCi/l)' ~~(pC1/kg. wet)? ' (pC1/kg; dry); -j

. Gross' beta 4:x 10+0' 1 x: 10-2 H-3 2 x '10+3, - - - - I :

.Mn-54 1.5 x 10+1 -

1.3 L x '10+2 . . .

l Co-58','-60 1.5 x 10+1 -

1.3 'x 10+2 . .. .

.t Zn-65 3 x 10+1 -

2.6 x 10+2 : .. ..

l Zr-Nb 1.5 x 10+1(D3 - - -

Cs-134 1.5 x 10+1' '5 x 10-2 -1.3 x 10+2 1.5 x 10+1 6 x 10+I 1. 5' x '10+2 - - .,

Cs-137 1.8 x 10+1

' l 6 x 10-2 1.5 x 10+2 1.8 x 10+1- 8 x 10+1 4

.1.8 x 10+2 ; ;

I :

(

i i

1 r

. (

4

.i i

.i Revision 11 nruro 4-7 '

3

, ,, , , . , - , . - - , - - - - ~ . . - -- . . , ~ - , , . .

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

l l

i l

' ; ,,5 H -TABLE 4.3

. i-} ;1:

1 -

(Continued) .

Table Notation ,

a. ~ The LLO.is the smallest. concentration of radioactive material in a sampleLthat 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.

JFor a' particular measurement system (which may include

[ radiochemical separation):

l

-(4.66) (Sb) '

LLD =

(E) (V) (2.22) (Y) [Exp(- Aat))

1 Where: I LLO = A priori lower limit of detection as defined above

,,, (microcuries or picocuries per unit _ mass or- volume),

ik}\ S b = 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.

l 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 1

.-Revision 11 nuuro 48 ,

e 1

r. . TABLE 4.3 (Continued)

Table Notation spectrometry, the background shall include 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 LLDs 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 caser, 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 orlori (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 nosteriori 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.

l

[ b. Parent only.

l 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 LLO.

j 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 Contrcl 7.1.

l-Revision 11 anuro 4-9

.~

.-.. 4.2 Land Use Census

- Y:

l Control 4.2 in accordance with Yankee Technical Specification 6.7.5 b.2, 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.

Applicability 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 4

l currently being calculated in SR 3.5, identify the new location (s) l in the next Semiannual 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 (Eth 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 Semiannual l Radioactive Effluent Release Report. ,

l Surveillance Reauirement 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, In lieu of the garden census, broad leaf vegetation sampling may be performed at the site boundary in the direction sector with the highest D/0.

Revision 11 anuro 4-10

9 aerial survey, or by consulting local agriculture authorities. 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 B0UNDARIES are identified and that modifications to the tonitoring program are made if required by the results of the land use census. The census satisfies the requirements of Section IV.B.3 ef

. 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 gardcn 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 D/0 may be substituted. The use of the maximum off-site D/0 value predicted for v

gaseous effluents from the plant stack (the plant stack does not qualify for an elevated release as defined in Regulatory Guide 1.111 March 1976) will 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 11 emue 4-11

4.3 Intercomparison Procram

('.s-)

~

} Control 4.3 in accordance with Yankee Technical Specification 6.7.5.b.3, analyses shall be performed on referenced radioactive materials supplied as l part of the quality assurance Laboratory Intercomparison Program.

Applicability At all times.

ACTION l 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.

I Surveillanct Reouirement 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 Radiciogical l 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-1, 10CFR Part 50.

4.4 Environmental Monitorino 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.

I Revision 11 anur: 4-12

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

i..-

TABlf 4.4 l -

i l: Rediolooical Environmental Monitorino Stations * '

4-

. Exposure Pathway._ ' Sample ~ Loc'ation l ~j Distance From Direction-

-and/or S mDie and Designated Code **

. the Plant (km) From Plant

! L1. AIRBORNE j -- - l ;

(Particulates)

-AP-11 Observation Stand 0.50 NW AP-12 Monroe Bridge 1.10 SW l

AP-13 Rowe School 4.20 t SE

.AP-14 Harriman Power Station 3.20'

', N AP.-21 Williamstown, MA 22.20- W l . 2. WATERBORNE'-

t a. Surface WR-11 Bear Swamp Lower 6.30 F Downriver Reservoir WR 21- Harrirr.an Reservoir -10.10 Upriver

! b. Grciund WG-11 Plant Potable j_ On-Site Well WG-12 Sherman Spring .0.20 NW f c. Sediment SE-11 -Number 4 Station 36.20 From Downriver -

SE 21 Harriman Reservoir 10.10 Upriver-g Shoreline

3. INGESTION

a. Milk TM-13 Whitingham, VT 8.40 ENE

. l- TM-21 Williamstown, MA 21.00 WSW

b. Fish FH-11 Sherman Pond 1.50 At Discharge and Inverter-Point FH Harriman Reservoir -10.10 . Upriver-brates

c. Food TF-11 Monroe Bridge 1.30 SW

. Products- TF-13 Monroe, MA: 1.90 WNW TF-21 Williamstown, MA 21:00 WSW

-l Revision 11 naute 4-13

1 i

TABLE 4,4 l .(Continued)

Radiological Environmental Monitorina Stations * ~

Exposure Pathway' Sample Location, ' Distance-From Direction ,

j and/or Samole and Desionated Code ** the Plant (km) From Plant i

4. DIRECT- GM 1 Furlon House; ~ 0.80 'SW ;

RADIATION GM-2 Observation Stand 0.50 NW GM-3 Rowe School 4.20: SE GM 4 Harriman Station 3.20 N .-

GM-5 Monrce 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 11- _ 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-

.z GM-18 Restricted Area Fence 0.14 ESE

440 GM-19 Restricted Area Fence 0.16' ' SE-

~,

i i - GM-20 Restricted Area Fence 0.16 SSE

[ GM 21 Restricted Area Fenco 0.11 SSW l .. GM Heartwellvilla 12.60- NNW

_l GM-23 Williamstown Substation .22.20 W j GM 25 Whitingham, VT- 7.70 NNE

_ l- GM-27 Number 9 Road 7.60 ENE i L GM 29 Route 8A 8.20 ESE

GM Legate Hill-Road 7.60 SSE.

GM 32 Rowe Road 7.90 S'

GM-33 Zoar Road 6.90~ SSW j l GM-35 Whitcomb Summit 8.60 WSW l GM Tilda Road 6.60 W '

l~ GM-38 West Hill Road 6.60- NW

l GM-40 Readsboro Road 0.50 !!

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

I -

loo

Station?1X's are indicator stations, and Station 2X's are control stations (excluding

_ the direct radiation stations). y l

9

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Figure 4-1 Yankee Plant Radiological Environmental Monitoring Locations ~'

Within'l Mile (Airborne, Waterborne and Ingestion Pathways)

Revision 11 ^

N

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Figure 4-2 Yankee Plant Radiological Environmental Monitoring Locations "

Within 12 Hiles (Airborne, Waterborne and Ingestion Pathways) i Revision 11 4-16

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0 5 10 15 20 t e i e l PITVSFIELD e

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

l Revision - 11 4 17

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Figure 4-4 Yankee Plant Radiological Environmental Mc,nitoring Locations at *'

tne. Restricted Area Fence (Direct Radiation Pathway)

Revision 11 4-18 4

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Figure 4-5 Yankee Plant Radiological Environmental Montwaring Locations *'

Within 1 Mile (Direct Radiation Pathway)

Hevision 8 4-19 i I

I N

N NN1;

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l 5- ""

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A gg d GM$l ssw '& . ss 4'et CHARLEMONT O 5 10 s 1 1 1 1 t 1 1 KILOMETERS ssE 3

SHELEURNE FALLS , j Figure 4-6 Yankee Plant Radiological Environmental Monitoring Locations Within 12 fiiles (Cirect Radiation Pathway)

Revision - Il 4-20

._ _. .. . _ _ . . - .1.. . _ _ . - - _ .

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, N

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/ 0 5 10 15 20 g i t 1 I KILOMETERS PITTSFIELD Figure.4-7 Yankee Plant Radiological Enviro.' mental Monitoring Locations .,

Outside 12 Miles (Direct Radiatica Pathway)

Revision 9 4 21

I i

15.0 INSTRUMENTATION 5.'l-L ' Radioactive liauid -Effluents

. l ' Control 5.1 In' accordance with . Yankee' Yechr.ical . Specification 6.7.5.a.1, the

~ . radioactive liquid effluent monitoring instrumentation channels shown in Table L5.lishall be OPERABLE with their alarm / trip setpoints: set to ensure that the olimits of Control 2.1 are.not exceeded. The alarm / trip setpoints of these schannels shall be determined in accordance with the ODCH.-

Acolicability c 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 '0PERABLE, 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 ~1n the next Semiannual Radioactive Effluent -

_ Release Report the reason for the delay in correcting the inoper bility. '

.- ' Surveillance Reautrement 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 11 anure 6I .,

,. hases

(-

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 instrumentat. ion is consistent with the requirements of General Design-Criteria 60, 63, and 64 of Appendix A, 10CFR Part 50.

l A gross radioactivity monitor which provides for automatic isolation of liquid discharges on detection of radioactivity concentrations in excess of l the values of 10CFR Part 20 (see Appendix B of ODCM), is included on the

l liquid radioactive waste effluent discharge line from the plant's test tanks.

l 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 (g Control 2.1 being exceeded.

l The sam'e gross radioactivity monitor provides alarm and spent fuel pool l cooling pump trip on the detection of radioactivity in the Auxiliary Service I Water effluent line (normally clean) when spent fuel pool cooling via the l spent fuel pool cooling heat exchanger is in operation.

l Composite samples are provided on continuous potential radioactive l effluent pathways (Turbine Building Sump and ASW discharge) to give assurance l that potential radioactive liquid releases to the environment are accounted l for (See figure 6-1).

Revision 11 e uc 5-2 .,

TABLE'5.1

-)

1

~ Radioactive Liouid Effluent Monitorino instrumentation

Minimum-Channels instrument OPERABLE :Ano11cability ACTION

1. ' Gross Radioactivity Monitors .

Providing Automatic Isolation a . -. Liquid Radwaste Effluent -(1) At All Times 1

~; l ' Linel ,

l 2. - Gross - Radioactivity Monitors not

. .- ll Providing Automatic Isolation Lj l a.-- Auxiliary Service Water (1) At All Times 4

.l~ . Ef fluent Linel l 3. -Continuous Composite Samplers

[ a . -- : Auxiliary Service Water (1)

5 l Effluent Linez

-l -b. ' Turbine Building Sump (1) 2 ll 4.- Flow Rate Measurement Devices r

f-[ _ -a. . Liquid Radwaste Effluent Line (1) -3 l b. ' Service Water System (1) **

6 l- 01scharge 3 l: c. Auxiliary Service Water (1) **

6

-l System Effluent 3 l=

A common radioactivity monitor (ASW-RM-001) provides' indication-for both

-the Liquid Radwaste-Effluent Line and the' Auxiliary Service Water Effluent Line _(servicing the. Spent Fuel Pool (SFP) Heat Exchanger operation).

e Liquid radwaste effluent discharges from the batch effluent. test tanks.

~

will> not occur when SFP cooling is active via the SFP Heat Exchanger.

Actuation of the radioactivity monitor alarm will 1solate test tank

releases (if in progress), and' trip SFP cooling pumps (if _in operation).

2.

t l The composite sampler is intended to provide water samples from the j ~. ' expected-clean flow-path associated-with SFP cooling. . When radwaste

.l - discharges from the' test tanks occur. the ASW composite sampler is secured --

_l ..from. operation.

2 Revision 11

. a!2\l20 5-2 4 .,

+

k 4

,.r -....,-~-. 7 , . , , . . , , - . ny,. , . . . . . - . . . . , , , ~ - .,., -...

. ~ _ . .m. .. . . , . _ ... _ .._ . __ _ .s_...._.. . .. .._m_m. ...__._.

V -_

L t

TABLE 5.1-(Continued)~ ,

~

Table Notation l- Pump and valve position-curves may be utilized to estimate flow.

~

In such

-- cases, the ACTION. statement-is not required.

Via this pathway when SFP cooling is active 'an the SFP heat exchanger.

~

l(ta. this . pathway duringl releases.

Y f

s Revision 11;

. nruao 5-4 * ,,

. < , , .-' - , , + -

m.. _

-TABLE 5.1

[ (Continued)

-ACTION Statements 1.- With the number of channels OPERABLE less than required by th~e

^

j. ACTION

. minimum channels OPERABLE requirement, effluent releases from the 1

tank may continue, provided that_ prior to initiating the release

-l. a. the conditions of ACTION 4 are met,.

f-.-l- 'b. - At it:t.st two independent samples of the tank's contents are

-analyzed in accordance with SR 2.1.1 -

-- l c. At least two technically qualified members of the facility. ;

staff independently verify the release rate calculations and discharge line valving, otherwise, suspend release of radioactive effluents via this pathway.

Ll ACTION 2,- With the number of- channels OPERABLE less than required by the minimum :hannels OPERABLE requirement, effluent releases via this

l. pathway may continue, provided grab samples are taken at least l once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and analyzed for grus radioactivity (beta or gamma) at a limit of detection of at least 1.0 x 10'7 l microcuries/mi at least weekly on a composite sample.

l _ 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. Pump curves may be used to estimate flow.

l ACTION 4 - With the number of channels OPERABLE less than required by the ;

l_ minimum channels OPERABLE requirement, the contaminated system-

,-l being cooled through a noncontact heat exchanger (s) shall be

-l- _ shutdown and the,ASW flow isolated by shutting down the ASW pump.

.)= '

I ACTION 5 i' With the number of channels OPERABLE less than _ required by the

-l minimum c5annels OPERABLE requirement, operation of the potential !

Revision 11

. amuo .- 5-5 a w w , ~-- -

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

i i N TABLE 5;1:

p.jf_ (Continued) -

ACTION Statements-t l; .' effluent release pathway may. continge, provided that the- '

l_ l associated effluent monitor is _ verified to be operating and grab -

ll- ' samples are taken at least once per ~24 hours and analyzed for

~

l'- gross radioactivity (beta or gamma) at a limit of detection of at~

, -l- least 1.0 x-10i microcuries/mi- at least weekly an a composite l sample.- *

-- l l ' ACTION 6 - - With the number of channels OPERABLE'less. than required' by the ,

l minimum-channels OPERABLE-requirement . operation of the potential- <

j. effluent release pathway may continue, provided that the flow rate l is estimated at least once per four. hours. Pump and. valve l position curves may be used to estimate flow, s

I 1

Revision 11 amuo 56 ,,

~ ~ . -. .

TABLE-5.2 Radioactive Liauid' Effluent'Monitorino !fstrumentation Surveillance Recuirements t

' MODES.in? :

~ CHANNEL iCHANNEL- >Which e.

. CHANNEL SOURCE. CALI- FUNCTIONAL 1 Surveillance'-

-Instrument CHEcr. CHECK BRATION. , TEST '

is Reauired-1 Gross" Beta or Gamma Radioactivity Monitors.Providing.

Alarm and Automatic Isolation ;

l a. Liquid Radwaste. Effluent; Line(a) . D' P R(2) -0 (I)

At- All.. Times' l 2. Gross Radioactivity Monitors Providing Alarm But~Not' l

l Automatic Isolation of Release ,..

a. Auxiliary.' Service Water Effluent Line(a.) .(b).

l .

(b) M- (b) . At_All Times l 3. Continuous Composite Samplers '

j ***

a. - Auxiliary Service Water Effluent Line .D NA NA 0 l l b. Turbine' Building Sump. D NA NA 0 l 4. Flow Rate Measurement Devices-

a. Liquid Radwaste Effluent Line D(3) NA R~ NA-l b. Service Water System Disch'arge D(3) NA' R NA .

i

-i Revision 11' airuzo 5-7

, ,. , . . . .. ww, + . - - - -

O T.

TABLE 5.2 (Contint.ed) 7adioactive ticoid Effluent Monitorinc inst ?ntation Survet11ance Rec cements MODES in CHANNEL CHANNEL Which l CHANNEL SOURCE CALI- FUNCTIONAL Surveillance Instrument CHECK CHECK BRATION TEST is Recuired

*^

l c. Auxiliary Service Water Discharge 0(3) NA R NA l Pump and valve curves may be utilized for flow rate determination.

Via this pathway during raleases.

l Via this pathway when SFP cooling is active on the SFP heat exchanger.

l(a) A common radioactivity monitor (ASW-RM-001) provides indication for both the Liquid Radwaste Effluent Line and l

Auxiliary Service Water Effluent Line. Actuation of the radioactivity monitor alarm will isolate test tank releases l (if in progress). and trip SFP cooling pumps (if in operation).

l(b) Channel Check. Channel Calibration and Channel Functional Test surve111ance requirements stated in 1.a. for this l commcn monitor.

Revision 11 n ru ,,

5-8

TABlf 5,7 (Continued) s.

Table Notation (1) The CHANNEL FUNCTIONAL TEST also shall demonstrate that automatic

l isolation of the 11guld radwaste effluent line, automatic trip of SFP l l cooling pumps, and Control Room alarm annunciation occurs if any of the following conditions, except as noted, exist

1

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

b. Circuit failure.

c. Instrument indicates a downscale failure (automatic pathway l 1 solation, and Control Room alarm indication.

(2) - The CHANNEL CAllBRATION 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 3 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 except where pump curves are used to estimate flow.

When pump curves are utilized as means of detereining flow, no CHANNEL CHECK is required. The CHANNEL CHECK shall be made at least once daily on any day on which continuous, periodic, or batch releases are made.

Revision 11 neure S9

5.2 Radioactive Gaseous Effluents I;

l Cont rol 5.2 in accordance with Yankee Technical Specification 6.7.5.a.1. the radioactive gaseous effluent monitoring instrumentation channels shown in j Table 5.3 shall be OPERABLE with their alarm setpoints set to ensure that the i li,mits of Control 3.3 are not exceeded. The alarm setpoints of these channels shall be determined in accordance with the ODCH.

Applicabillix As shown in Table 5.3.

ACTION l

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm setpo, int less conservative than a value which will ensure that the limits of Control 3.3 are met, without delay, take actions to suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel-inoperable, or change the setpoint, se it is acceptably conservative.

With less than the minimum number of radioactive gaseous effluent

, b.

monitoring instrumentatien channels OPERABLE, take the ACTION shown in Table 5.3. Exert reascaable efforts to return the instrument (s) to OPERABLE status v! thin 30 days and if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report the reason for the delay in correcting the inoperability.

I Surveillance Reauirement SR 5.2 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CAllBRATION, and CHANNEL FUNCTIONAL TEST operations during the HODES and at the frequencies shown in Table 5.4. ,-

Revision 11 ensuo 5-10

Esses The radioactive gaseous effluent instrumentation in the primary vent stack is provided to monitor, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm setpoints for these instruments are set conservatively to ensure that the limits of Control 3.3 are not exceeded. 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.

The primary vent stack exhausts building ventilation air, as well as gaseous _ process streams, to the atmosphere and as such, cannot be isolated due !

, to building ven{11ation requirements.

I e

Q Revision 11 oruro 5 11

O O TABLE 5.3 Radioactive Gaseous Effluent Monitorino Instrumentation Minimum Channels -

Instrument OPEPABLE ADDlicability Parameter ACTION

1. Primary Vent Stack l a. Noble Gas Activity Monitor (1) Radioactivity Rate 9 Measurement l b. Particulate Sampler Filter (1) Verify Presence of Filter 8 I c. Effluent System Flow Rate (1)* System Flow Rate Measurement 7 Measuring Dtsice l d. Sampler Flow Rate Measuring (2)* Sampler Flow Rate Measurement 7 Device

At all times.

i

One channel per pathway (VC. and PAB ventilation flow paths) required.

l b

One flow device channel is for the noble gas channel, and a separate channel is for the fixed particulate filter.

Revision 11

.irura 5-12 4 4 9 e

. - . e -

l TABtf 5.3 (Continued)

ACTION St at ement s, I l ACTION 7 + With the number of channels OPERABLE less than the minimum channels OPERABLE requirement, effluent releases may continue, l provided the flow rate is estimated at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l ACTION 8 - With the number of channels OPERABLE less than the minimum channels OPERABLE requirement, effluent releases via this pathway may continue, provide' m91es are continuously collected with auxiliary sampling ew'!wa*, <is required in Table 3.1.

l ACTION 9 - With the number of channels OPERABLE less than the minimum channels OPERABLE requirement, effluent releases via this pathway may continue, provided grab samples are taken at least once per

, l 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

O O

Revision 11 i

ausuo 5-13

1

. l l b l

l l TABLE 5.4 l

1 f Radioactive Gaseous Effluent Monitoring instrumentation Surveillance Recuirements l

CHANNEL MODES in which CHANNEL 50URCE CHANNEL FUNCTIONAL Surveillance Instrument CHECK C*IECK CALIBRATION TEST is Recuired

1. Primary Vent Stack l a. Noble Gas Activity Monitor D M RW Om Particulate Sampler Filter W NA NA NA l b.

c. System Effluent Flow Rate Measuring D NA NA NA l

Device

d. Sampler Flow Rate Measuring Device D NA R 0 l

l

At all times.

Revision 11 eiruze 5-14 I

TABtf 5.4 (Continued)

Table Notation (1) The CHANNEL FUNCTIONAL TEST also shall demonstrate that Concrol Room alarm annunciation occurs if any of the following conditions exist:

l a. Instrument indicates measured levels above the alarm setpoint.

1

b. Circuit failure,

c. Instrument indicates a downscale failure.

(2) The initial CHANNEL CAllBRATION for radioactivity measurement instrumentation shall be performed using one or more of the reference standards certified by the NIST or using standards that have been obtained from suppliers that participate in measurement assurance '

activities with NIST. Subsequent CHANNEL CAllBRATION sources that have been related to the initial calibration can be used at intervals of.at t

least once per 18 months.

l

)

i c

Revision 11 au,na .

5 15

.; n . ~

~

5.3 tiouid Effluent instrumentation Setnoints ;

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

l Use the method below to determine the setpoints for the required I instrumentation. -

5.3.1 Method

-J l The liquid Radwaste Effluent monitor response (cpm) for the limiting !

concentration at the point of discharge is the setpoint, denoted R, and is !

determined as follows* '

1 1

f %

l R= (MPCc ) (St) (Eq. 5 1)

Where:

f3 - Flow rate past the test tank monitor (gpm).

I l f 2

- Flow rate at the point of discharge (gpm).

S, = Instrument response factor (cpm /( C1/ml)).

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

l MPC c = E Cs/E Cg/MPC, = Et f /E fi /MPCi = 1/(E f /MPCs) i (Eq. 5-2) *

, 1 i i i i 1 .

Where:

HPC = MPC for radionuclide "1" from 10CFR Part 20. Appendix B.

3 l Table 11, Column 2 ( Ci/ml). See ODCM Appendix 8.

l Revision 11 mmc 5-16

% C, = Concentration of radionuclide *1* in mixture (pC1/ml).

p )- '

W .

f, = fraction of radionuclide *1* in mixture.

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

l The setpoint. R. may be administrative 1y set lower to accommodate l -j pathw',ys which normally are nonradioactive (Auxiliary Service Water). The l auxillary Service Water is a normally clean ' system. The same radiation l effluent monitor provides detection of the presence of an off normal condition j that may have unexpectedly introduced radioactive contamination to this clean j system. The alarm setpoint when only ASW cooling flow is in operation is set l 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 l discharges are made. This requirement allows the common radiation monitor to j see either the expected clean ASW pathway flow when SFP cooling operations are '

l on going (potential source), or the expected radioactivity in the test tank l effluent flow when this discharge pathway is in operation.

5.3.2 Liauid Effluent Setooint Example l The effluent monitor for the test tank release pathway is geana l

_ l sensitive monitor. It has a typical sensitivity. S. of 2.8E+8 cpm per pC1/mi of gamma emitters which emit one photon per disintegration and a typical l background of about 330 cpm.

l The composite HPC and setpoint can be calculated based on the following

-l- example data:

i f, HPCj (pCi/ml) l Cs 134 0.02 9 x 10 6 l Cs-137 0.18 2 x 10'5

,- 1 Co-60 0.80 3,x 10 5 Revision 11 mun 5-17 E

,s r 5,

, k. ) '

HPC*

I f,/MPC i -

1 l . 1 (Eq. 5 2)

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

HPCc = 2.6 x 10 5 (pC1/ml) l l For this example, normal liquid effluent flow rate (f ), is assumed to be g .

l 2.8 gpm. Dilution water flow, f r is assumed to be 200 gpm (equivalent to l total flow of both contaminated and clean water). The setpoint for the j monitor when the test tank effluent pathway is operating is then calculated

( l for these example conditions to be:

ig R= (HPCc ) (Sf) 200 gpm' (Eq. 5-1) l -

(2.6 x 10 5 Ci/ml) (2.8E+8 cpm /(pC1/ml))

g, ,2.8 gpm ,

- 520,000 cpm I

l This setpoint value may be administratively set lower than the maximum l 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:

(c m)" (b')Ir f t 51) (Cnon) (Eq. 5-5)

Revision 11 urau 5-18

l J

Where variables are the same as those in Section 5.3.1 except:

,,3-- .

-C n ou ~' Total concentration (pC1/ml) seen by the monitor, s, ~ ~ Ratio of response from equal activities of radionuclide *1" to a reference radionuclide.

Calibration of the radiation monitors have established that the gross gamma l detector. response. S E fg sg was fairly independent of gamma energy as '

I expected. Thus, the response is a function of radioactivity concentration and the gamma yield of the mixture. Since E f, si is approximately one: {

t .

R = (St) (Cnon). (Eq. 5 6)

-ll for simplicity, assume that the monitor looks at a flow for f . We know that:

3 C=

't1 7 (Cnon) (Eq. 5-7)

Where:

C - Total concentration at the point of discharge.

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

non r, ,

l R= (C) (St) , (Eq. 5 8) r >

Revision 11 i m iro -

5-19

l We defined C = E Cy and define the composite HPC c such that:

4 i C ,g C' HPC c t MPC, (Eq. 5 9) l The right side of the equation is the sum of the ratios of the HPC limits in l 10CFR Part 20 ( Appendix B of the ODCH). Solving for HPCc, the composite HPC. ,

for the mixture. We get the definition of HPC -c E C; '

i HPCc = , (Eq. 5 2) g 61 i MPC, Substituting HPCc into Equation 5 6, we get the response of the monitor as HPCc is reached at the point of discharge, which is the setpoint:

r 3

'S l R- b f (HPCc ) (54) (Eq. 5-1)

<1>

Revisior 11 nrurs 5-20 4

4 r

5.4 Gaseous Effluent' Instrumentation Setootnts y .

Control 5.2 requires that the radioactive gaseous effluent

-instrumentation in Table 5.3 have their alarm setpoints set to ensure that Control 3.3.a is not exceeded.

I Use the method below to drtermine the setpoint for the noble gas activity monitor.

l l 5.4.1' Nethod l-.

The noble gas activity monitor response (cpm) at the limiting noble gas-T-

dose (either total body or skin off+ site) is the setpoint, denoted R. and is

l. determined for Kr 85 as the only noble gas left in the. spent fuel _ following l, permanent shutdown in October 1991.

~

R is the lesser of: '

(500) (60) (Sgr.85) l Rtb - (Eq._5 3)

-(f) (7.83) 0FB Kr 85 And:- -

p , (3000) (60) (SKr es)

(Eq. 5 4)

.(F) DF Kr 85 Where:

1. 7.83 - 1046 (pCi/pC1) 7.83 x 10 (sec/m3 )

l- 0Fg "r 85 - Combined Skin dose factor for Kr-85. See Table 1.2.

[- DFB gr.85

Tctal body' dose factor for Kr 85. See Table 1.2.

I Revision 11 aiture 5-21

f. F = Primary vent stack flow rate (cc/ min),

l. '

l S tr 85 - Gaseous monitor response fattor for Kr 85 (cpm /(pC1/cc)).

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

5,4,2 Gaseous Effluent Setooint Example The primarf vent stack noble gas activity monitor is an off line system ,

consisting of a beta sensitive scintillation detector, electronics, an analog ratemeter readout, and a digital scaler which counts the detector output l l pulses. System characteristics are: -

l a. Typical sensitivity - I cpm - 3 x 10 8 pCi/cc of Kr 85: that is -

l S - 3.3 x 10*7 cpm /(pC1/cc) l b. Typical background 100 to 200 cpm Under normal plant stack flow. F, of 5.80 x 10 48 cc/ min (20,500 cfm x 28,300 cc/ft 3).

Applying Equations 5-3 and 5-4:

(500) (60) (3.3 x 10+7) - 1.35 x 10+7 cpm Ra == (5.80 x 10'8) (7.83) (1.61 x 10'6) l Rg - (3000) (60) (3.3 x 10+7) - 318,000 cpm (5.80 x 10+8) (3.22 x 10'2) l The setpoint, R is the lesser of R tb and Rg; therefore, it is equal to l 318,000 cpm. Since Kr 85 is the only noble gas radionuclide left for ,

l potential release, the skin dose rate limit calculated in the example is the l maximum setpoint permitt,d.

Revision 11 muzo 5-22

5.4.3 Basis The noble gas activity monitor setpoint must ensure that Control 3.3.a is not exceeded. Sections 3.9 and 3.10 show that Equations 3-3 and 3 4 are l acceptable methods for complying with Control 3.3.a. The derivation of Equations 5-3 and 5 4 starts with the general equation for the response. R l (cpm), of a radiation monitor and the basic assumption that only one noble gas j radionuclide is expected to be available for detection (Kr-85) for the j long term shutdown mode:

R-(SKr es) (C)

(cpm) - (cpm /( C1/cc)) (1) (pC1/cc) (Eq. 5-5) i Expanding for the concentration:

i R _(SKr 85 (6) (60/f) l (cpm) - (cpm /(pC1/cc)) (1) (pC1/sec) (sec/ min)/(cc/ min) (Cq. 5-10)

()

y The response of the monitor at the release rate which causes the total body dose rate limit to be reduced. Rtb. begins with Equation 3-3.

! btb

7*03 bKr85 DFB Kr 85 l where 7.83 is equal to the limiting off-site gamma X/0 from Table 3-6 times a l unit conversion factor of 10+6 pCi/pC1.

Rearranging to solve for 0 :

D l 0-7.83 0FB Kr 85 (Eq. 5-13)

Substituting Equation 5-13 into Equation 5-10 and substituting the total body dose rate limit gives:

Revision 11 sieuro 5-23 e 9

SKr es (cpm /(pCi/cc)) 500 (mrem /yr) 60 (sec/ min)

( l Rtb " (Eq. 5-3)

F (cc/ min) 7.83 (pCi-sec/pci-m ) 3DFBKr 85 (mrem m3 /pC1-yr)

The response of the monitor at the release rate which 'auses the skin dose rate limit to be reduced. R,g. begins with Equation 3 4:

l Osk = 0 DF g'p.85 Rearranging to solve for 0:

0= 6 (Eq. 5-15) l .

DF Kr es Substituting Equation 5 15 into Equation 5-10 and substituting the skin dose rate limit of 3.000 mrem /yr gives:

p ,SKr 85 (cpm /(pC1/cc)) 3000 (mrem /yr) 60 (sec/ min)

F (cc/ min) DFKr 85 (mrem-sec/ Ci-yr)

Revision 11 muro 5-24 l

_...--__.__.____.m ..____-_.- . - . _

s 5.5 Meteoroloaical Instrumentation Control 5.5 The meteorological monitoring instrumentation channels shown in =

l; Table 5.5 shall be OPERABLE. l Aeolicability At all times.

ACTION

a. With one or more required meteorological monitoring channels inoperable for more than 7 days, prepare and submit a Special Report-to the Commission pursuant.to Control 7.4 within the next 10 days outlining the cause of the malfunction and the plans for-restoring the channel (s) to OPERABLE status.

Surveillance Reauirement SR 5.5 Each of the above meteorological _ monitoring instrumentation channels

shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK and

) l CHANNEL CAllBRATION operations at the frequencies shown in Table 5.6.

Bases The OPERABILITY of the' meteorological instrumentation ensures that sufficient meteorological data are available for estimating potential radiation doses as a result of release of radioactive materials to the atmosphere. This capacity is consistent with the recommendations of Regulatory Guide 1.23, "On-Site Meteorological Programs."

.~

Revision 11 neuro 5-25

n. - _ , . _ . _ _ , . . . ,.. _ _ . . , _ , . _ - - ~ _ . ...~ , . - . - . _ _ , _ .._,

l TABLE 5.5 Meteoroloaical Monitorino Instrumentation ,

HINIMUM J;STRUMENT LOCATION OPERABLE

1. WIND SPCEO a. Nominal Elev.199 feet 1

b. Nominal Elev. 35 feet 1

2. WIND J1 RECT 10N a. Nominal Elev. 199 feet 1

b. Nominal Elev. 35 feet 1

3. AIR TEMPERATURE: a. Nominal Elev.197-33 f'eet 1 ,

DELTA T .

Revision 11 ausac 5-26 ,,

l TABLE 6,6 Meteorolooical Monitorinq Instrumentation Surveillance Recuirements 1 CHANNEL CHANNEL INSTRUMENT CHECK CAllBRATION

1. WIND SPEED

a. Nominal Elev.199 feet D SA

b. Nominal Elev. 35 feet D SA

2. WIND DIRECTION

a. Nominal Elev. 199 feet D SA

b. Nominal Elev. 35 fett D SA

3. AIR TEMPERATURE: DELTA T

a. Nominal Elev. 197 33 feet D SA

?

Revision 11 nu.us 6-27

4 6.0 RADIDACTIVE WASTE TRt. 9ENT SYSTEMS. EFFtVENT PATHWAYS. AND RADIATION

[) NONITORS .

6.1 Liouid Radioactive Waste Tr itment j . Control 6.1 -In accordance with Yankee Technical Specification 6.7.5.a.6, 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 O' 06 mrem to the total body or 0.20 mrem to any organ.

. Acolicability At all times. I ACTION a.

With liquid waste being discharged.without processing through appropriate treatment systems as defined in the ODCH and estimated doses-in excess of the above limits, and if not applicable to g 10CFR Part 50.73, prepare and submit to the Commission within P 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 Reauirement SR 6.1 Doses due to 11guld releases shall be estimated at least once per 31 days in accordance with the ODCH. No dose estimates are required if the-Revision 11 unuo 61

l l

Liquid Radioactive Waste Treatment 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.

Basel ;

The control that the appropriate portions of the Liquid 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 11.0 of Appendix 1. 10CFR Part 50. The specified -

limits governing the use of appropriate portions of the Liquid Radioactive Waste Treatment System were specified as a suitable fraction of the dose design controls set forth in Section ll.A of Appendix 1, 10CFR Part 50, for liquid effluents.

O l

Revision 11 s.inio G-2 *

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

_ _ _ . _ - _ m. _ _ _ _ _

I 1

6.2 Gaseous Radioactive Waste Treatment

)

i i--

l Control 6.2 In accordance with. Yankee Technical Specification 6.7.5.a.6. the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive ;

materials in gaseous waste prior to their discharge when the estimated doses due to gaseous effluent releases from the site to areas at and beyond the SITE .

COUNDARY (see Figure 5.1 3) would exceed 0.30 mrem to any organ over 31 days.

Aeolicability i

At all times. - '

.- ACTION

a. With gaseous waste being discharged without processing through appropriate treatment systems as defined in the ODCH and in excess of the above limits and if not applicable to 10CFR Part 5f.73, prepare and submit to the Commission within 30 days. purtaant to Control 7.4. a Special Report that includes the followint information:

{} 1. Explanation of why gaseous r8J)oactive waste was being F discharged without treatment. identification of any t inoperable equipment _ or subsystems, and the reasons for the inoperability:

. 2. Action (s) taken to restore any inoperable equipment to OPERABLE' status, and *

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

I-Surveillance Re'ouirement

'SR 6.2_' Doses due to gaseous releases from the site shall be estimated at

, least once per 31 days in accordance with the ODCH.

2 j Revision 11 s

j esiuro 63

-+-l, -

,w- ,,n,,,,-e,w.ww e w n ~.-- , , , , ,a-w r, ,--w -

s- ,...,,-.,w,, . ,,e- ., ,v,,w m ,,,m-- m,r,,w--w,w--,

d 4

{ Bases l The control that the appropriate porticns of the VENTILATION EXHAUST TREATMENT SYSTEM be used when s.ecified provides reasonable as:urance that the releases of radioactive materials in gaseous ef fluents will be kept "as low as is reasonably achievable." Control 6.2 implements the requirements of 10CFR

.Part 50.36a, Ger.eral Design Criterion 60 of Appendix A, 10CFR Part 50, and the l design controls of Appendix 1,10CFR Part 50.

e 5

i 4

4 i

4 Revision 11

17\ UO b-4 e

<-.y. -- .--riwe , - .m F-

6.3 tiauid and Gaseous Effluent Streams. Radiation Monitors and Radioactive

,) Waste Treat' ment Systems l Figure 61 shows the liquid effluent streams, radiation monitors, and j the appropriate Liquid Radioactive Waste Treatment System. Figure 6 2 shows the gaseous effluent streams, radiation monitors, and the appropriate l Ventilation Exhaust Treatment System.

6.4 in Plant '! quid Effluent Pathways

.l A number of auxiliary coolers and heat exchangers discharge service water (Sherman Pond water) into a common underfloor discharge header. The

Turbine Building floor drains and Auxiliary Boiler Room floor drains drain to

-this header. The water from this drain header discharges without further dilution into a tributary of the Deerfield River outside the controlled area.

- A composite sampler collects a sample of the water whenever there is discharge (water in the pipe).

Batch effluent tanks called

test tanks' collect the distillate from the i l-atmosphericliquidradioactivewasteevaporator. Normally, liquid waste I l accumulates at about 0.3 gpm ,and is processed at about 1 gpm. When the test j t9nks are full, they are sampled, analyzed, and released at a nominal 5 gpm.

l Auxiliary service water provides dilution water flow. Flow rate is variable and estimated by pump curves. Typically, flow rates range from l 175 gpm to 200 gpm for normal auxiliary service water use. Additional l dilution flow at point of discharge to Sherman Pond may be provided by service l- water pumps (nominal 2000 gpm) if operating.

l- The maximum discharge rate for the Turbine Building pathway is estimated j.to be 150 gpm. All piping is buried and inaccessible, so flow is estimated

. j - from periodic observation.

1 Calibrations of the radiation monitor has established that the gross gamma detector recponse was fairly independent of the gamma energy, as expected. Thus, the response is a function of the radioactivity concentration

.and tp gamma yield of the mixture, but not the gamma energies of the mixture.

l The electronics of the monitor channel has an adjustable alarn, setpoint.

Revision 11 emin 6-5

.s i

t . .

6.5 In Plant Gaseous Effluent Pathways

(' .

The primary vent stack noble gas effluent monitor is an off-line system consisting of a beta sensitive scintillation detector, electronics, an analog ratemeter readout, and a digital scaler which counts the detector output l pulses. Calibration data is provided by the mantfacturer which indicates the response of the beta sensitive detector to various gaseous radionuclides. The calibration data was verified on installation and periodically thereafter.

System characteristics are:

l a. Typical sensit'vity 1 cpm - 3 x 10'8 pCi/cc Kr-85 l b. Typical background 100 200 cpm

I I

Revision 11 nrore 66

FIGtlRE 6-1 Liauid Effluent Streams. Radiation Honitors, and Radioactive Waste Treatment System at the Yankee Plant j ----+ se erwewsm

. ==

n U ne.re.r l

^=a va-**w r < . r- 1

cre H i emh , ,.

~

s , ,

e i I so1.a e om.r j %*

. %,s.r s-4 t

's

. w r.,- - -

l

% s Am -

we.m=.

e

g Twnte.mid.

nm

. u

.I ca o.* '

n s-we-de v en-e 9 l h T-e.

m n.

c.r .

% o.m.v-v stWm Dr.me n t.6s-.,ew ,

1 -

vc ==

- was n a

- em-vv e m.* u.m =w

- ma

= mp.

n Tar.uh D. .

Dr.ti

-a e s-i,

-t.mt.n.masa ma n-,w aL - en m i.easie.r a

- . rw ew

- e.a n

..................................................,............M.l......v........

ee ! o.-a.w w sh.a.a

. cassnu maa.nemam **'*

Revision 11 attuto 6-7 t

1 i

_ _ _ _ _ _ _d

__ ....____.._.._._.._.__._..._~-._._-._._m___... _

~. . _ .._.. _.

l FIGllRE 6-2 4

Gaseous Effluent Streams Radiation Monitors. and Radioactive Waste Treatment System at the Yankee Plant i

. Atmo.phere

, JL

!- O TritiumSampler b- - Noble Particulate Sampler Gac Monitor 4

)

1 I l 3

l l I

j I O* Fan ---

i HEeA eaE i vc veatriation i i.

' I I'

) I vr mpy."t.r

::ji

~

t i a E ,I

.y ,

i i

- - n.d Wa.i. oi.po.ai Bunding

- Primary AuxHiery Butdmg

! Fan

, l - Spent Fuel Pool Building l HEPA PRE ,

- Red Waste Evaporator DetRiate Tank Wnt l

l - Aerated Uquid Wo.te Tank Vents l

l g - Rad Waste Evaporator Trailer I I I

I Gaseous RadW g nd l Treatment System i

l. l I

I I

u__________ l n

Turbine Building Wntilation

4 COGTRUM6 FIG 6.2 (R12\120) '

1

- Revision 11' uruto 6-8

t 6

,,,-r w s + - . '. . , - - , ,a.v.- >c.w,-y s,vw..-,._. -,.-,.. ~

---e---m , e .. -.a. --, =

7.0 REPORTING REOUIREMENTS L

7.. Annual Radioloaical Environmental Operatino Report b Cont rol 7.1

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

2

5. The Annual Radiological Environh.2ntal Operating Report shall include summaries, interpretations, and an analysis of trends of

, the results of the radiological environmental surveillance e

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 shCIl include the results of the land use census rsquired by 3 Control 4.2.

The Annual Radiological Environmental Operating Report shall lg 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 explair_ing the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.

The report also sh.91 include the following: a summary description of the kodiological Environmental Mont toring 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.

L a

Revision 8 neuro 7-1 O

. _A

r______ -____---___________-____--_ _ _ - -

O 7.2 Semiannual Radioactive Effluent Release Report

[

i- Control 7.2

a. Within 60 days after January 1 and July 1 of each . :.or a report shall be submitted covering the radioactive content of effluents released to unrestricted areas during the previous six months of the year.

b. The Semiannual 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 if# f " Measuring. Evaluating, and Reporting -

Radioactivity in Solid W.utes and Releases of Radioactive Materials in Liquid and Las 30us 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 Semiannual Radioactive Effluent Release Report l to be submitted 60 days after January 1 of each year 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. The Semiannual Radioactive Effluent Release Report sula ttted within 60 days of July 1 each year need not contain any

(*) In lieu of including in the Semiannual Radioactive Effluent Release l

Report required to be submitted within 60 days after JanJary 1 additional information that covers the assessment of radiation doses, 6 supplemental report is permitted to be submitted within 90 days after January I conttining this information.

Revision 12 7-2

dose estimates from the previous six months effluent releases.

'\ This report also shall include an assessment of the radiation dore; from radioactive effluents to MEMBER (S) 0F 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 l 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 ODCH.

l The Semiannual Radioactive Effluent Release Report to be submitted

, l 60 days after January 1 o'," each year also shall include an assessment of radiation doses to the likely most exposed real MEMBER (S) 0F THE PUBLIC from reactor releases (including doses from primary effluent pathways and direct radiation) for the previous calendar year to show conformance with 40CFR190,

" Environmental Radiation Protection Standards for Nuclear Power Operation," if Control 3.2 has been exceeded during the calendar year.

l The Semiannual Radioactive Effluent Release Report shall include a

{bg 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.

l The Semiannual Radioactive Effluent Release Report shall include any changes made during the reporting period to the ODCH, 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 11 ausue 7-3

, -7.3 Ma.ior Chanoes to Liouid and Gaseous Radioactive Waste Treatment Systems h .

Control 7.3 Licensee initiated major changes to the liquid and gaseous

~

radioactive waste-systems:

l a. Shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period .in which the evaluation was reviewed by 1he PORC. 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 10CFR l Part 50.59. -

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

lL

.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 e(pected maximum exposures to MEMBER (S) 0F THE PUBLIC et the SITE-BOUNDARY and to the general population-that differ from 1 those previously estimated in the license appl! cation 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 cs a result of the change. and i!

Revfsion 11 R12\120 7-4 i

8. Documentation of the fact that_the change ..s reviewed and found acceptable by the PORC.

b. Shall become effective upon review and acceptance by the PORC.

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. Liquid Effluents. Controls 3.1 and 6.1.

b. Gaseous Effluents. Controls 3.4. 3.5, and 6.2.

c. Total Dose. Control 3.2..

d. Radiological Environmental Monitoring, Control 4.1.

l e. Meteorological Honitoring. Control 5.5

[ Revision 10 R12\ l20 7-5

,, REFERENCES a.

Regulatory Guide 1.109, " Calculation of Annual Doses to Man f rom 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., 'AE0LUS 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-1120, January 1977.

c.

Regulatory Guide 1.111, " Methods for Estimating Atmorpheric Transport and Dispersion of Gaseous Effluents in Routine Releases From Light-Water Cooled Reactors," U.S. Nuclear Regulatory Commission March 1976.

d. NEP 1 and 2 Preliminary Safety Analysis Report, New England Power Company, Docket Nos. STN 50 568 and STN 50-569.

e. Yankee Atomic Technical Specifications, f.

Yankee Atomic Electric Company Supplemental Information for the Purposes of Evaluation of 10CFR Part 50, Appendix 1. Amendment 2, October 1976 (Transmitted by J. L. French - YAEC to USNRC in letters dated x June 2, 1976: August 31, 1976; and October 8, 1976).

nt g.

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.

1. TDR-122374, " Isotopic Standard zation of Yankee Rowe Vent Stack Monitor."

e Revision 8 ausuo R-1

l

\- .

l l

\-

APPENDIX A DISPOSAL OF SEPTAGE O

e%

h

1 4 0 tac oq'o

+<

UNITED STATES fgg p, NUCLEAR REGULATORY COMMISSION

}

, ~. . ;eE

WASHINGTON,0. C. 20555

( % * *%.* ***

,/

MAY 171990 l Docket No.50-029 11r. George Papanic, Jr.

Senior Project Engineer - Licensing Yankee Atomic Electric Company 580 Main Street -

Bolton, Maso thusetts 01740-1398

Dear Mr. Papanic:

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 cwned treatment works is acceptable.

Sincerely, o

Patrick Sears, Project Manager Project Directorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Enclosed:

As stated cc w/ enc 1:

See next page HAY 211990 '

Revision 7 - Date: A-2 Approved By: > 4

Mr.- Gurge PapInic, Jr.

Yankee Rowe

_j cc:

~

Dr. Andrew C. Kadak, President

' and ' Chief Operating Officer -

Ycnkee Atomic Electric Company 580 Main Street B31 ton, Massachusetts 01740-1398 Thomas Dignan, Esquire Ropes and Gray 225 Franklin Street'

, Boston, Massachusetts 02110 i Mr. T. K. Henderson i

Acting Plant Superintendent Yankee-Ator.ic Electric Company Star Route Rowe, Massachusetts 01367 Resident Ir.spector Yankee Nuclear Power Station U.S. Nuclear Regulatory Commission Post Office Box 28 .

Monroe Bridge,. Massachusetts 01350 Q Regional Administrator, Region I y- U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406 4

Robert M. Hallisey, Director Radiation Control Program Massachusetts. Department of Public Health 150 Tremont Street, 7th Floor Boston, Massachusetts 02111 Mr. ' George Stertinger Commissioner Vermont Department of Public Service 120 State Street, 3rd Floor

> Montpelier, Vermont 05602 Ms. Jane M.. Grant

Senior.Engineei - PLEX Licensing Yankee Atomic Electric Company 580 Main Street Bolton, Massachusetts 01740-1398 1

Revision 7 - Date: A-3

'[

ApprovedBy:gg[,,,,,[, , y

( ..

~

e

)-

-SAFETY ASSESSNENT BY THE OFFICE OF NUCLEAR REACTOR REGUtATION 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 C

with, and disposed as part of the sarritary waste generated from many sources..

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

In the submittal, the licensee addressed specific information requested in accordance with-10 CFR 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.

Revision 7 - Date: A-4 Approved By: /I ,-s d I ..

i i

~

-2 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 w'aste-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.

Ruvision 7 - Date: 2I D A-5 Approved By: M! A D7

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2.2 Radiological Properties 4

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The plant's sanitary. system septic tank collects waste from the lavatories, showers, and janitorial facilities outside the Ra'diological 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 j 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 (uC1)

Co-60 1.94 .

Mn-54 0.057 Cs-134 0.082 Cs-137 0.248 TOTAL 2.33 Revision 7 - Date: N U 2 1 1990 A-6 ApprovedBy:[ 4

4 j 3.0 PROPOSED DISPOSAL HETHOD 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 f;rtilizer, 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 h' tank pumpers incinerate their ::ludge 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 Recuirements and. Limits

, The licensee will perform a gamma isotopic analysis on a representative sample of. waste from the s'eptic tank no more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 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

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-tank by taking a volumetric column sample that will allow the licensee to Revision 7 - Date: N4Y 2119M A-7 Approved By: Af[ -

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{)- 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)

requiremers 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 h herein before disposal. The following limits apply to these analyses:

1. The concentration o,f radionuclides detected in the volume of septage to be pumped to a dis'posal truck shall be limited to a combined sum of fractional Maximum Permissible Concentrations in Water (HPC)-(as listed in 10CFR Part 20, Appendix B. Table II, Column 2), sumed 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: H4Y 2 1 19n9 A-8 Approved By: Af./ 4LN c- ..

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3.2 Administrative Procedures The licensee will maintain complete records of each disposal. In addition to c: pies 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 cotal cccumulated 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, rcdionuclide concentrations in the liquid and solid fractions, and the total h cctivity 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 1cnd 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 R: vision 7 - Late: NSY 2 I I990 A-9 Approved By:,-ff b 4 t ..

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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 1MPACTS The'licenset has evaluated the following potential exposure pathways to members p

of the general public: (1) inhalation of resuspended radionuclides, (2) ingestion b of food grown on the disposal site, (-3) 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 ari consistent with regulatory Guide 1.109.1 The staff finds the assessment methodology acceptable. -

Revision 7 - Date: 1930 A-to Approved By: ~Arfm w Q l

1 Regulatory evide 1.109, "Calculetion of Annual Doses to Man from Routine Releases "

of teactor1,"

Appendiy Effluents for the Purpose of Evaluating Compliance Vith 10 CFR Part 50, Revisicn 1, October 1977. ,

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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 i

i of 2.3 pC1,- more than 80% of which is Co-60):

Maximally Exposed Ir.dividualNhole Body (Child)

Pathway (mrem / year)

Ground Irradiation 0.099 Inhalation 0.0001

. Stored Vegetables 0.0214 Leafy Vegetables 0.0011-Hilk 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 followingtable:

Revision 7 - Date: 1 990 A-11 Approved By: 4/ J %)

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Maximally Exposed Individua1Nhole Body Pathway (mrem / year)

Ground Irradiation 0.980 -

inhalation 0.0004 Stored Vegetables- 0.13 Leafy Vegetables 0.007 s

TOTAL 1.1 Based on this same total- activity, the dose to truck drivers and SVTF 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 19'0.10(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 standatd 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: NU 2 I 030 A-12 Approved By:/fs[ ;

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An alternate means of disposal would involve the treatment cf the septage as radwaste. Such a disposal would require that the licensee stabilize, solidify, cnd 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 taximum value of 1 millirem per year, Yankee ensured that radiological risks from the proposed disposal method are insignificant.

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). The proposed release limits repr'esent 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 tjormal liquid waste to the environment. Any resulting dose to any individual in the public is less than exposures caused by natural background radiation.

Based onjur review of the proposed disposal of septage, the staff makes the following conclusions: (1) the radionuclide concentrations in disposed septage will be a small percentsge 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 45 O h e

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- facility and will require only insignificant' changes in the handling or transport of radioactive material (septage);; and (4) the licensee's procedures with commitments as documented in the submittal are' acceptable, provided that the submittal is permanently incorporated into the licensee's Offsite Dose Calculation

Manual.(00CH) as an Appendix, and future modifications will be reported to NRC.

L-in.accordance with' licensee commitments regarding 00CM changes.

Contributors: J.' Minns-P.-Sears

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Revision 7 - Date: 2 1 1990 a.14- Approved ny: g// ww)

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YANKEEATOMICELECTRIC COMPANY "*' "' l' *) "***' "

TWX 710-380-7619 y

580 Main Street, Bolton, Massachusetts 01740-1398 m ,

April 11, 1990 BYR $90-42 United States Nuclear Regulatory Commission Attention: Document Control Desk W2shington, D.C. 20555 R3ferences:- (A) License No. DPR-3 (Docker No. 50-29)

Subject:

10 CFR 20.302 Application

Dear Sirs:

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

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

Very truly yours, YANFEE ATOMIC ELECTRIC COMPANY George Papanic, Jr. /

Senior Project Engineer Licensing Enclosure GP/emd Revisiort 7 Date N 2I D Approved By* M MXZ A, -

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i- YANKEE NUCLEAR POWER STATION APPLICATION FOR APPROVAL TO ROUTINELY DISPOSE OF SEPTAGE UNDER 100FR20.302

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I Revision 7-- Date: MAY 211930 A-16 Approved By: /

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i L TABLE OF CONTENTS t =.1 h ^

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TABLE OF CONTENTS................................................. 11 LIST OF TABLES.................................................... iii IJ ST OF FI GURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv=

1.0 INTRODUCTION

...................................................... 1

- 2.0 WASTE STREAM DESCRIPTION.......................................... 2

2.1 Physical / Chemical Properties................................ -2 2.2 Radiological Properties..................................... 3

-3.0 PROPOSED DISPOSAL METB0D........................................... 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 D.

P- '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 Carden Pathway Scenario......................... 4.. 14 5.3.4 Incineration Pathway Scenario....................... 20 5.4 Maximum Releasable Activity................................. 21 6.0

SUMMARY

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

7.0 REFERENCES

........................................................ 24

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Ravision 7 - Date: A-17 Approved By: 4 , y ud .

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.. LIST OF TABIE Number riti. y,g, 1 Landspreading: Ingestion Pathways (Adult) 25 2 Landspreading Ingestion Pathways (Teen) 26 3, Landspreading Ingestion Pathways (Child) 27 4 Landspreading Ingestion Pathways (Infant) 28

-lii-MAY ? 1 I?53 Revision 7 - Date: _ A-18 Approved By: < ~s *

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LIKT OF FIGURES-Number Title gagg 1: Yankee Plant Sanitary Waste Disposal Process 29

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Revision 7 - Date: A-19 Approved By: M '

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0 1

i a YANKEE NUCLEAR POWER STATION '

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Annilention for AnnrovAli .

i to Routinelv Dianone of Eentmee finder 10c71t20.302 l~ 1.0- INTRODUCTION t .

Yankee Atomic Electric. Company (YANKEE) requests approval, pursuant to *

10CFR20.302(a) 'of a method proposed herein for'the routine disposal' l (typically, once every one to two years) of septic tank waste containing very - -

!! Ilow levels of license,d material over an extended period of time of;30 years.-

Yankee proposes to periodically dispose of accumulated. septic waste' solids I from.the plant's sanitary system septic tank by transferring--it to a public i Sanitary Waste-Water Treatment- Facility (SWTF) where it will be mixed with, processed, and-disposed off as part of sanitary waste generated from many sources. _ This is analogous to other Nuclear Regulatory Commission (NRC) -

licensed faciliti,es who have their sanitary waste -systems connected directly j 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.

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- Routine maintenance of Yankee's septic system is necessary to ensure l

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 e.pabilities of -

the' field.-

This application addresses specific information requested.in 10CFR20.302(a), and demonstrates that the periodic disposal cf septage from *

, . Yankee'stSanitary 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.- !

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~ Revision'7 - Date: MAY 21 IM A-20 Approved By: A /4 ,/ <

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I 2.0 WAETE WATER STREJL*1 DESCRIPTION g .

P 2.1 Physical /Chemient Pronerties The waste involved in this application consists of residual septage (cccumulated settled and suspended solids, and scum) associated with the canitary sewerage _ collection and treatment system at the Yankee plant. The Yankee plant utilizes an on-site septic system supplemented with off-site tr atment 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 pr: cess.

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 tha solid particles settle to the bottom and form a layer of sludge, where

) grecses 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 1siching 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 '

esptage stored in the septic tank is reduced to liquid by bacterial action in tha septic tank, but the rest of the septage remains essentially untreated.

This material must be pumped out at regular interrals to prevent it from cvarflowing the tank and entering the leaching fi>1d (Reicrences 1, 2, 3, 4, 5, 6, 7, 8, 9, and 10) where it will clog the soil e d etentually lead to

,esptic system failure.

In general, sept! age 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 ssptage 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.

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Rcvision 7 - Date: A-21 Approved By: [ / '

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

2.2 Radiolorical Pronerties i_ The plant's sanitary system septic tank collects waste from the lavatories, showers, and janitorial facilities outside the Radiological Contrcl Area (RCA). No radioactivity is intentionally discharged to the

septic system. However, plant investigations into the so cce 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 radiosctivity releases from the P.CA, 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 vash water) is now either poured through a filter bag to remove suspended solids and dirt before the water is l released into a janitorial sink, or the wash water is returned to the RCA for

{ disposal.

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

Activity Concentration Nuclide (eci/ke wet +/- 1 siema)

West Manhole East Manhole Samele Location Samole loention Co-60 92.4 3.9 13.2 i 2.2 Cs-134 5.9 1 1.3 --

Cs-137 9.2 1 1.5 3.2 1 1.0 After the initial analysis of the composite samples noted above, the samples were subsequently centrifuged irto 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.

.M Revision 7 - Date: A-22 Approved By: -e - -Q

An:1ysis of t'ne resulting solid fraction of the septage indicated the f0110wiar radionuclide enacentrationst

%ctivity Concentration Euclide t oci/kc vet +/- 1 nima) _

West Manhole East Manhole Samole Leention Sample toentien Co-60 1,588 2 42 Ma-54 528 4 26 Cs-134 47113 -

67 .t 11 -

Cs '.?? 203 A 17 100 1 13 The original septic tank samples were volumetric samples representative cf the distribution of solids e.nd 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 cf total solids content of the septic tank. For the vaste sample fr'om the v:st manho'le, the solid fraction of the composite sample was found to bs s 0.024, or 2 t 3 %.

For the east manhole, the solid frac *too of the total h sample was 0 i, or 4.6 wt.%.

The principle radionuclide is (daalt-60, which cccounts for appror!aately 82% of all plant-related activity detected in the c:ptage.

The total radioactivity content of the septic tank can be estimated by cniculating the mass of solids present in the 7,000 gallon tank by taking the higher (conservative) solids fraction determined from the sam,,le data. This is multiplied by the mass of septage in the tank and by the highest activity c ncentration determined in the solids. As a result, the estimated maximum total activity ist Total Activity Nuclide (uci) 9 Co-60 1.94 Mn-54 0.05/

Cs-134 0.082 Cs-137 0.248 TOTAL 2.33

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R vision 7 - Date:

I:4Y 2 1 IS93 A-23 Approved By: ,Ma

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3.0 rROPostn on,POSAL MrTHOD '

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Upon approval irom the U.S. Nucicar Regulatory Commission (NRC), Yankee proposes to periodically dispose of accumulated 6 ptags from its septic tank by contracting with a. town-approved (Board of Health, Rowe, Massachusetts) septic tank pumper icr the removal and transfer by truck of the septage to a Massachusetts SWIF fcr treatment. At the SWTF, the septage vould typically be mixed and diluted with other raw sewage and introduced e hner 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 reruit in the processed sludge being landspread as a fertilizer, though u nerally 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 .f treatment.

This method of tank pun. ping and transfer to an SWTF is identical to h, 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 10CFR20.303. 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 syrtem 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 har.dle sanitary waste. In this case, a tank truck acts as a sewer line in transferring septage to a SWTF. The quantity and fonn (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.

Y.X 2115% ~

Revision 7 - Date: A-24 Approved By: _

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3.1 Sentie Tank Waste Procedural Recuirements and Limits Cassoa isotopic analysis of septic tank waste shall be made prior to tren der of the waste by a contracted septic tank pumper to a SWIT by cbtaining a representative sample from the tank no more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 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 detemined and applied to the gamma isstopic analysis in order to estimate the total radioactivity content of the tank.

These games isotopic analyses of the representative samples will be perfomed at the Technical Specification Environmental lever Limit of Detection (LLD) requirements for liquids (see Technical Specification T;ble 4.12-1, " Detection Capabilities for Environmental Sample Analysis") in crd:r to document the estimation of radiological impact from septage disposal.

1 The radionuclide concentrations and total radioactivity identified in th3 septage will be compared to the concentration and total curie limits cat:blished 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 Pemhsible Concentration of Water (MPC) (as listed in 100FR, 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 c21 mrem to any individual in the public).

U Rsvision ? - Date: ,A-25 Approved By: A WM +

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if the total activity limit is met, comp 1'ance with the self-imposed

( dose criteria will have toen demonstrated since the radiological impact (section 5) is t 'ised on evaluating the exposure to a maximally exposed hypothetical inkidual such that his annual whole body dose would be limited to approximately 1 area. -

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. Tf 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.

t 3.2_ Administrative Procedgggg i t

Complete records of each disposal vill be maintained. In addition to copies of invoices with approved septic tank' pumpers, these records.will h.includetheconcentrationofradionuclidesintheseptage,thetotalvolumeof 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.

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Revision 7 - Date: A-26 Approved By: AgM _

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4.0 EVAmATION OF ENVIRONMENTAL IMPACT O

The proposed method for disposal of septage is the same as currently us:d by all fscilities designed eith septic tanks for the collection of septic w:ste.

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 t:pography, geology, meteorology, hydrology, or nearby facilities by the pre).osed method of disposal. ,

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Rsvision 7 - Date: A-27 Approved By: dd -<+<

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5.0 EVALUATION OF RADIO 1hCICAf> IMPACT k~

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 SWIT. The processed sludge would typically be buried ir, 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 -

l consider the maximum potential impact of long-tem buildup of activity }

resulting from 30 years of placing septage vaste in the same SWIT, with all .I the processed sludge assumed to be buried in one landfill disposal cell. l 5.1 Septic Tank Sample Analvale Data

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The analysis of the septic tank's measured radioactivity, and its distribution between liquid and solid fractions, provides the bases upon which s

a dose a'sessment 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 co' lid 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 (ocl/ke) Wet foci /kci Wet Mn-54 -

47 Cs-134 -

67 Cs-137 100 203 Co-60 528

~9-1.588 ..

Revision 7 - Date A-28 Approved By: #

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. With the s:ptic tank v31ume t ken cs cypr:ximatoly 7,000 gs11cas I s (16,500 liters), and essuming the maximum solid fraction (0.046) and maximum -

i radionue11de concentration applies to the total cank's content, the total seminum radioactivity content is estimated to bos factopa E=1 f-13 f a De (cil Mn-54 _312.2 day 5.73 E48 00-60 5.271 yr 1.94 E-06 Cs-134 1.065 yr -

8.17 E-48 Cs-137 30.17 yr- 2.48 E-07 5.1 PmNav **aa-ura se. =*ian. -

Radiological evaluations were perfomed >r both the expected cctivities associated with handling, processing, and disposal of septage waste -

3

-ct a 5WTF, and a hypothetical event causing undiluted'septage release. The beunding case was determined to be associated with a hypothetical event which ic:d'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

. s:ptic tank pumpers will direct that_ septage be disposed of at a SWIT 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 ecse activities could occur which lead to a long-tem buildup of activity on a single minimum size garden plot.

, In addition, since incineratica o'f septic waste is not a current' prcctice in the local area, the potential exposures associated with incineration are not of current concern. However, the establishment of a-censervative total whole body dose criteria for release of sanitary waste, via

'th) above-noted garden scenario, assures that the potential resulting whole brdy dose'due to incineration would not be expected' to result in significant deses to'any individual.' This assessment is further detailed in Section 5.3.4.

~

The contributing pathways of- exposure for the normal' SWIF disposal

-process includet '

1. External exposure to a truck driver.

2. External exposure to a SWTF worker.

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. 3. Ext:rnal exposuro tis an individual standing on ths SWIT landfill

-)

af ter 30 years of buildup and decay.

l

. The following garden exposure pathways were addressed for the maximally

= exposed hypothetical individual:

1. 6tanding 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-eilk-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 l

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, th'refore, e considered independently.

5.3 Dose Assessments 5.3.1 Externat Ernemure to a Truck Driver /swrr Worker The external dose rate from a 3.500-gallo'n tank truck filled with septage containing the total measured activity in the septic tank (2.33 901)

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 gansna 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 .,

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truck driver making o 100-milo trip to o tr:ctment fccility et an cvarego cf 4 20 miles per hour plus a three-hour waiting period at the SWIF. is estimated to be 9'.5E-04 mrom. 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 s:ptage handling, a SWIF employee's direct exposure time is kept to a ninimum. Using the dose rate estimated for the truck driver above, and conservatively assuming that it requires an employee at the SWIF a full cight-hour day to process each truckload of vaste, and not taking any credit

fer dilution or increased distance from the vaste, a vaste processing facility employee's dose is also estimated to be S.5E-04 mrem.

If the maximum activity content proposed to be disposed of each year w re assumed as the source term (20 pC1), the dose to the truck driver /SWIF '

varker is estinated to be less than 1.0E-02 mrem using the same assumptions as n:ted above. '

5.3.2 Ext emm1 Excesur. Due to Lone-Tem Buildup P .

In order ,to assess the potential impact from the postulated buildup of cctivity 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 0.211 hours 0.00126 weeks 2.8918e-4 months per year). -

For regional SVTFs vaste sludge is typically mixed with san 5 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, cs requir'ed by Massachusetts Department of Environmental Protection

, ccgulations (Reference 19 . The landfill disposal cells range in site from cbout one acre up to about five acres. Af ter a cell is full, a final layer of ctmpacted material is required to be placed over the entire surface of the call to a minimum depth of two feet (Reference 16).

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Analytically, if Qo is the amount of radioactivity per tank full of b./ septage for a give nuclide, then the total accumulated radioactivity Q,(max) disposed of after 30 pumpouts is given by:

Q,(max) = Qo (1 + E + E2,g3+E4 + .... + E29)

= Qo (1 - E29)/(1 - E) (A) where E = exp(-Mt)

, 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:

A Qo Qe(max)

Nuclide Half Life (1/vear) fuci/ 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 ,ith SWTF sludge that formed a minimum one foot layer whir- 's placed immediately below the two-foot disposal cap of the cell, the ts- ting _

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___ _ __--..---.-_-- ---- --- - --- - - ----- - ~

dose rate one cater cbove the ground surface was calculated to be 3,

6.4E-07 cres/ hour. If it is also essumad that en iridividual remained on the lendfill for a full year (8.760 hours0.0088 days 0.211 hours 0.00126 weeks 2.8918e-4 months ) without taking any credit for shielding by a rt.sidential 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 recult 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 af ter 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 SWIF, 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 Carden 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 war. given to the following:

1.

The computation of an effective self-shielding factor ta account for the effect provided by the soll after the vaste is plowed or mixed in the top 15 cm surface layer.

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2. The definition of an annual activity release rate, which following

(, a year's time of continuous release, vould yield the ground deposition expected to Prevail af ter a tank pump-out and spreading on the 500 ft2 garden. ,

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

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

fendepreadine. Pecuepenelen. nnd Occupancy Factors

If it is assumed that the garden plot is limited to a surface area of 2

500 f t , then the land deposited radioactive material S (Ci/m2 ) following landspreading will be equal to:

S, = Q (Ci)/(500 ft2

0.0929 m 2 /ft2)

(B)

Tbe denominator of this equation is equivalent to the (D/Q) deposition factor normally employed in the airborne impact assessment of deposited radionuclides; that ist (D/Q) = 1/(500 ft2

0.0929 m 2

/f t2) .

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

Following the application of undiluted septage on the garden, some of .

the radioactivity may becomo 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 A, (C1/m3 ) and the surface deposition is:

A, = S, (Ci/m ) x K (1/m) (D)

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l wheros K is the resuspension fccesr and is tek:n to bo oqual to 1.0E-06 (1/a) !

(Reference 11) which is believed conservative due to the limited i i surface area icvolved and the irrigation provided to a garden which '

1 minimises airborne dust.

The 500 ft1 garden s'ise was selected based on the minimum surface area cecessa'ry 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 oculd be expected to produce sufficient food to support the uptake assumption on food consumptionn'oted below.

In addition, by limiting the garden surface area to 500 ft2 (a circle I with a 3.85 m radius) the concentration of radioactivity in the garden is :

maximised 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 i

500 f t . 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 place'd over it.

As for the occupancy factors for direct exposure to the ground dsposition and for immersion inithe' resuspended radioactivity -360 hours was ussd 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 ,

o plot of. land or garden during the growing season in the northeast- (average two hours a day for six months).

Ce.rdenpathwaydataandusagefactorsasahplicabletotheareainthe v$cinity.of the plant are shown below. These are the same facters as used in tha plant's ODCM assessment of the off-site radiological impacts due to routine releases from the plant, with the following exceptions; l

p I

~ - . . . . , - , , m., . . . - . , , . . . ~ . . -- -r-.,,,.

1. De c2ii exposure t'me i w:s chang:d from 15 ysers to 1 y :r to

.} account for the discrete application of septage on a garden plot.

4

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 0.0444 hours 2.645503e-4 weeks 6.088e-5 months to 0 hours0 days 0 hours 0 weeks 0 months to reflect the condition that no radioactive material would be dispersed directly on crops for human or sniv.a1 consumption, the deposition on crops of resuspended radioactivity being .

insignificant 1y small; that is, crop contamination is only through

, root uptake.

1l SAGE FACTORS Vegetables Leafy veg. Milk -

Individual (ke/vr) (kr/vri (liters /vri Inha}ation*

(m /vr)

Adult 520 64 310 329

, Teen 630 42 400 329 re Child 520 26 330 152 a Infant 330 58

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

VEGETABLE PATIN&Y Stored Leafy Vecetables Vecetables 2

Agriculturalproductivity{kg/m)

Soil surface density (kg/m )

2.0 2.0 240.0 240.0 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 i Fraction of stored vegetables

grown in garden 1.0 Fraction of leafy vegetables grown in garden -

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_ _ - _ _ _ _ _ _ _ - _ _ - - - - - - - - - - - - -~

COW-MILK PATHWAY Pasture Feed Stored Feed 2

Agricultural Soil surface productivity density (kg/m{kg/m) ) .7 2.0 240.0 240.0

' 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 vaste shows that all the activity is associated with the suspended or settled solids fraction, and not dissolved in the liquid portion, trinsport of activity through groundwater vould not be expected to lead to drinking water supplies being impacted by septage placed on farm lands. It is, therefore, not anticipated that the groundwater pathway could result in doces comparable to the direct' surface exposure pathways. As confirmation of this, however, a methodology'for groundwater analysis, as developed by 7 ,

Kannedy, 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 pC1, with the organ being the LLI vall.

These results are several orders of magnitude below the direct surface exposure doses as detailed below. The groundwater pathway is, therefore, not significant.

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Direct cround Plane Ewoosure To account for the gama attenuation provided by the soil, it was necessary to carry out an appropriate shielding calculation. This was cecomplished 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/ce, which is equivalent to ths Regulatory Guide 1.109 value of 240 kg/m2 for the effective surface dsasity 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 /br. *Ihe total dose from standing on the garden area for 360 hours0.00417 days 0.1 hours 5.952381e-4 weeks 1.3698e-4 months each year is seen to be 0.099 mrem from the total activity content measure in the septic tank (2.33 pC1) being placed on the garden, f carden Pathway Total Dose The maximum individual ingestion / inhalation exposure assessments resulting frem 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 af ter a tank pump-out and spreading on a garden at a concentration IcVel equivalent to the measured concentrations in septic waste are as follows:

Radiation Excesure Individual /Orcan Maximum Exposed Individual 0.122 mrem /yr Child /Whole Body 0.157 mrem /yr Child / Liver Revision 7 - Date: 2I A-38 Approved By: M , . ad L

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

The individus1 pathway contributions to the total dose are as follows:

Pathway-Dependent critical Orcan Doses

Maximally Exposed Maximally Exposed Individual / Organ Individual /Whole Body Pathway (Child / Liver) (Child)

(erem/venri (mrem /vear)

Cround Irradiation 0.099 0.099 Inhalation 0.0003 0.0001 Stored Vegetables 0.055 Leafy Vegetables 0.0214 0.0028 0.0011 Milk Ingestion

f0.019) (0.0036)

. TOTAL 0.157 0.122 Tables 1 through 4 detail the internal dose breakdown by radionuclide l' and pathway of exposure. As can be seen in the results, the whole body and maximum exposed organ dose are appropriately equivalent. This is 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.

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power reactor low level wastes.

This distribution was similar' to the measured distribution in ths Ycnkee septage in that Co-60 and Os-137 were the '

predominant gamma emitters.

The results of their analyses show that the transport worker receives the highest dose frna the incineration scenario. Ibe transport worker dose is approximately a factor of 5 higher than either the maximum incinerator worke~r 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

. trcnsport verker dose will not change if the 'septage is incinerated, since it was conservatively assumed that the worker spends 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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.

f 5.4 Maximen 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 5or about 96% 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 su:q of all measured gamma emitting nuclides can be' assessed as Co-60 to determine the quantity of gross activity that, if p 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: A-40 Approved By: MI . M .

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

The breakdown by expesure pathway for this sce' n ario, assuming an activity release of 20 microcuries in the form of Co-60 is as follows:

Maximum Exposed Pathway' Individual /Whole Body (erem/vear)

Cround Irradiation 0.980 inhalation 0.0004 Stored Vegetables 0.13 Leafy Vegetables 0.0068

. TOTAL 1.1 All other scenarios for tLe 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.

f The following sumary compares the calculated whole body doses I

associated with normal handling of septage with the 1 mrem bounding event garden scenarlo.

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 I crem/ year:

~ Maximum Whole Body Annuni Dose Seennrfe (erem)

(a) Septic truck driver /SWTF worker. 1.0E-02 (20 uCi Co-60 per year) -

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

9 9

p- . - .

_ _ _ _ - _ _ _ - _ _ - _ - - - - - - - ~ ~

6.0 StMMARY AND CONCLUSIONS Ihe disposal of septage by transferring it to a public SWTF is in accordance with standard practices for treatment of the type of waste material ganerated by a septic tank / leach field sanitary'vaste system. Periodic pumping of the septic tank is necessary for the maintenance.and continued operation of Yankee's sanitcry waste syst'em. Approval for disposal of septic waste from the Yankee sanitary system is requested to prevent failure of the senitary system to ad'equately handle plant domestic waste.

Alternate means of disposal of the septage vould 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 vaste far outweigh the concerns due to any radioactivity which is present.

By setting release limits which restrict the exposure to a maximum bypothetical individual of .1 mrem per year, it is ensured that radiological risks from the proposed disposal method are of no significance.

I The proposed release limits represent a small fraction of NRC limits permitted for disposal of similar vaste by licensed facilities who have their sanitary systems connected direct 1y to a public sanitary sewerage system.

These proposed limits are also within the plant's current allowable release limits for discharge of nomal 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.

g "

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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-600/8-60-032, August 1980.

3. " Handbook - Septage Treatment and Disposal," U.S. Environmental Protection Agency, EPA-625/6-64-009, 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 Eavironmental 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. Hacuer, " Water Supply and P'o11ution Control," International Textbook Company, 1971.

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

11. Cember, H.,

1969.

" Introduction to Health Physics " Page 321, Pergamon Press,

12. Kennedy, W. E., Peloquin, R. A., " Residual Radioactivity Contamination From Decocaissioning," NUREC/CR-5512, January 1990 (Draf t ' Report for Coenent ).

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

14.

J. N. Hamavi, " 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 ERC 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).

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TA8LE 1

IMDSPREADINO ikOESTION PATWATS (ADULT)

(2.33 001 TOTAL ACT!YITT)

.I .

(MAEM)

PA?WAY S0dt L!ntt KIONET LUWO DI Lt1 THYtotD VROLE toof ..

I m LATION 54 M C.00E+00 2.93E*06 7.21E*07 1.04E*04 5.72E 06 0.00E+00 4.66E 07 40 00 0.00E40 2.11E 05 0.00E+00' 1.09E 02 5.21E*04 0.00E+00 2.71t 05 135 OS 3.17E*05 7.22E 05 2.44E*05 4.31E 06 8.23E*07 0.00E+00 6.19E 05 137 CS 1.07E 04 1.39E*04

, 4.98E 05 1.6st 05 1.taE 06 0.00E40 9.$tE 05 T67ALFORPATWAY 1.39E 04 2.35E*04 7.49E 05

).11E 02 5.30E 04 0.00E40 1.E5E44 O

STORED YECETA8LES

$4 m 0.00E+00 3.10E 04 9.21E*C5 0.00E+00 9. TEE 04 0.00E+00 5.91E 05 60 00 0.00E+00 1.72E*03 0.00E+00 0.00E+00 3.34E*02 0.00E+00 3.92E 03 134. CS 2.24E*03 5.33!*03 1.72E*03 5.72E 04 9.32E 05 0.00E+00 4.35E 03 137 CS 9.25E*03

1.27E 02 4.29E 03 1.43E 03 2.45E*04 0.00E+00 8.2VE 03 107AL FDL PATWAT 1.15E 02 2.01E 02 6.11E*03 2.00E 03 3.47E 02 0.00E40 1.64E 02 LEAF 7 YECETAILES 54 M 0.00E40 4.34E*05 1.29E*05 0.00E+00

1.33E*04 0.00E+00 E.29E 06 60 CD 0.00E+00 2.24E44 0.00E+00 0.0fM0 4.20E*03 0.00E+00 4.93E*04 136 CS 2.91E*04 6.92E*04 2.24E 04 7.44E 05 1.21E 05 0.00E+00 5.66E*04

%7 CS 1.14E 03 1.56E 03 5.31E-04 1.76E*04 3.03E 05 0.00E+00 1.02E 03 TOTAL FOR PATWAT 1.43E 03 2.52E43 7.68E 04 2.51E 04 4.33E 03 0.00E+00 2.09E 03 C3tMILE 54 M 0.00E40 2.39E*06 '7.10E 07 0.00E+00 7.31E*06 0.00E+00 4.53E 07 60 CD 0.00E+00 5.33E*05 0.00E+00

0.0X+00 1.00E-03 0.00E+00 1.1EE*04 a 134 CS 8.11E*04 1.93E 03 , 6.25E 04 2.07E 04 3.38E*05 0.00E 40 1.5EE 03 137 03 3.31E*03 4.53E*03 1.54E 03 5.11E*04 8.77E-05 0.00E+00 2.97E 03 y TOTAL TOR PATWAT 4.12E 03 6.51E 03 2.16E*03 7.18E 04 1.13E-03 0.00E+00 4.66E 03 25 Fevision 7 Date I'N 2 I III A 4'4 Approved Dy: 44[ . *

/

TMLE 2 LAND $PREADlWG IWCEST!0N PATWAYS (TEEN)

(2.33 UCI total ACTIVITT)

(HREM)

PA?WAT SONE LIVER CIONET LUWG CI LLI TNTR0!D VWott SCOT ..

IkKALA710W 54 M 0.00E+00 3.78E 06 9.41E 07 1.47E44 4.94E*06 0.00E+00 6.21E L7 60 CD 0.00E+00 2.77E*05 0.00E+00 1.60E*02 4.75E*04 0.00E40 134 CS 3.63C 05 4.28E 05 9.60E*05 3.19E 05 1.25t*05 137 C2 8.31E*07 0.00t40 4.67E 05 1.50E 04 1.90E 04 6.80E 05 2.70E*05 1.90E 06 0.00t+00 6.96E 05 101AL FOR PATWAY 1.93t*04 3.17E*04 1.01t-04 1.62E 02 4.82E 04 0.00E+00 1.53E 04 SfotED VTCETABLES l 56 M O.00t+00 4.54E 04 1.44t 04 0.00E+00 9.93E 04 0.00t40 9.60E 05 i 60 CD 0.00E+00 2.83E 03 0.00E+00 0.00E+00 3.69E 02 0.00t+00

! 136 0s 6.3 N 03 3.65t*03 8.59E C3 2.73t 03 1.04t* 03 1.07E*04 137 Cs. 0.00E40 3.98E 03 1.57t 02 2.10E*02 7.13t*03 2.77E 03 2.9CE*04 0.00E+00 7.30E 03 10TAL FOR PATWAT 1.94E 02 3.29E 02 1.00E 02 3.81t 03 3.83E 02 0.00t+00 1.75E 02 LEAF 7 t!!CETABLEs 54 M 0.00E+00 3.68E 05 1.10E*05

0.00E+00 7.53E*05 0.00E+00 7.30E 06 60 CD 0.00E+00 1.93E 04 0.00E+00 0.00E+00 2.51E*C3 0.00E+00 934 Cs 4.34E 04 2.57E*04

6.0$E 04 1.921 04 7.34E 05 7.52E*06 0.00E+00 2.81E*04 137 CS 1.0$E 03 1.40E C3 4.77E 04 1.25E 04 1.99E*05 0.00E+00 4.SSE 04 107AL 70R PATWAT 1.31E-C3 2.24E-03 6.8 @ 04 2.59E 04 2.61E-03 0.00E+00 1.21E*C3 CoJ MILC 56 M 0.00E+00 3.98E 06 1.19E*06 0.00E+00 8.15E 06 0.00E+00 7.SEE 07 60 CD 0.00E+00 9.C3E 05 0.00E+00 0.00E+00 1.18E*03 0.00E40 2.C3E 04

,134 C3 1.41E 03 3.3tE C3 1.05t 03 4.02E*04 4.121 05 0.00E+0c 1.54E*C3 13? CS 6 00E 03 7.99E*C3 2.72E C3 1.06E*C3 1.14E*04 0.00E+ " 2.7EE-C3 TOTAL fcut PATWAT 7.4tt.C3 1.14E 02 3.77E 03

. 1.46E-C3 1.3t.E C3 0.00E+0; 4.522 03 26 Revision 7 Date 2I A-45 Approved By: _/U w ,

L

- - ~ ~ ' - ~

TABLE 3 LANDEPREADlWO IkCESTICW PATHWATS (CMILD)

(2.33 UCI TOTAL ACTIVITT.)

(MREM)

PA?WAT 80WE LIVIA r!DWET LIJWO Cl*LL1 TKTR0!0 VHDLE $37 INKALATION SS M 0.00Z+00 3.17E*06 7.41E 07 1.17E 04 1.69E 06 0.00E+00 7.03E*07 60 cc 0.00E+00 2.40E*05 0.00E+00 1.29E 02 1.76E*M 0.00E+00 4.15E 05 134 Cs 5.54E 05 8.63E*05 2.81E*05 1.03E 05 3.27E 07 0.00E+00 1.91E 05 137 CS , 2.03E 04 1.85E*04 4.32E 05 2.33E 05 8.10E*07 0.00C+00 2.87E 05 TOTAL fot PATWAT 2.58E 04 2.98E*04 9.20E 05 1.31E 02 1.79C 04 0.00E+00 9.00E*05 e *

$70 RED YECETA8tts 56 M 0.00E+00 7.25E 04 2.03E 04 0.00E+00 6.08E 04 0.00E+00 1.93E M 60 CD 0.00E+00 4.40E 03 0.00E+00 0.00E+00 2.44E*02 0.00E+00 134 CS 1.30E 02 8.42E*03 1.38E.02 4.28E.03 1.54E 03 7.45E 05 0.00E+00 2.91E*03 137 CS 3.80E*02 3.63E*02 1.18E*02 4.26E 03 2.27E*04 0.00E+00 5.36E 03 107AL fot PAT WAT _4.64E*02 c.14E 02 1.63E*02 5.80E*03 2.53E 02 0.00E+00 2.14E 02 LEATT WCETMLES

$4 M

0.00E+00 4.13E-05 1.16E*05 0.00E+00 3.47E 05 0.00t+00 60 EO 1.1M-05 0.00E*00 2.25E-04 0.00E+00 0.00E+00 934 CS '1.24E 03 0.00C+00 6.62E.04 4.43E 04 7.30E 04 2.26E*04 8.11E 05 13? Cs 3.93E 06 0.00E+00 1.54E 04 1.90E 03 1.82E 03 5.94E*04 2.14E*04 1.14E 05 0.00E+00 2.69E 04 707AL f0R PATWAT 2.35E 03 2.82E 03 8.32E*04 2.95E*04 1.29E-C3 0.00E+00 1.10E 03 COW MILE 56 M 0.00E+00 5.95E 06 457E06 0.00E+00 4.99E*06 0.00E*00 1.58E 06 (4 CD 0.00E+00 1.40E 04 0.00C+00 0.00E*00 7.77E 04 0.00E+00 4.132 04 134 CS 3.25f-C3 5.33E 03 > 1.65E C3 5.93E*04 2.87E*05 0.00E+00 1.12E 03 13? CS 1.45E 02 1.38E.C? * .51E*03 . 'E*03 8.67E*05 0.00E+00 2.04E 03 V WAL FOR PATHWAT 1.77E 02 1.93E 02 4.1TE 03 2.22E*03

. 8.97E*04 0.00E+00 3.51E C3 e

27 Revision 7 Date A-46 Approved By://h gaJ ,

/

Q.:w ~

_ ____ __ - - - - - ~

TARLE 4 LAwosrREMIND IWCEST10W PATMTS

(!kfANT)

(2.33 L'CI TOTAL ACTIVITT)

(MREM)

PAI M T SONE LIVER K!DNET LUWO Cl LLI TNTR010 WOLE BOOT IkkALATION

$4 M 0.00E+00 1.E7E 06 3.69E 07 T.39E 05 5.2iE-07 0.00E+00 60 ED 0.00E+00 3.69E 07 1.47E 05 0.00E+00 8.25E 03 5.84E 05 134 Cs 0.00E 00 2.16E 05 3.37E 05 5.98E 05 1.62E 05 6.78E 06 137 Cr 1.1&E 07 0.00E 40 6.34E 06 1.23E 06 1.37E 04 3.t$E 05 1.59E 05 2.99E 07 0.00E+00 1.02E 05 TOTAL to PAT M T 1.57E 04 2.13E 04 5.51E 05 8.35E 03 5.94E'05 0.00!+00 3.84E 05 o .

STORta VECETABLES

56 M 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 60 CO 0-00E*00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E*00 0.00E+00 0.00E+00 136 Cs 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 137 C3 0.00E+00 0.00E+00 0,00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 207L'. FOR PAT M T 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 l LTAFT VEDETAALES

! 5% M 0.00E*00 0.00E+00 0.00E+00 0.00E+0 3 0.00E+00 60 CD 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 134 Cs 0.00E*00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 13? C8 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E*00 0.00E*00 0.00E+0C 0.00E*00 0.00E+00 0.00E+00 TOTAL. TOR FAT M T C.00E+00 0.00E*00 0.00E*00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Cou KILX 54 M 0.00E+00 1.11E-05 2.45E 06 0.00E+00 4.06E-06 60 CD 0.00EMH3 2.51E-06 c 00E+00 2.86E 04 0.00E+00 0.00E+06 6.81E 04* ,0.00E+00 134 Cs 5.23E 03 9.76E 03 6.76E C4 2.51E 03 1.03E 03 2.65E 05

13? Cs 0.00E+00 9.tSE 04 2.31E 02 2.7DE*02 7.25E 03 2.94E-03 8.&5E 05 0.00E+00 1.92E 03 TOTAL FQ3 FAT M T 2.83E 02 3.71E 02 9.77E 03 3.97E 03 7.96E 04

. .. 0.00E+00 3.5SE-03 28 Revision 7 Date IO

' A.47 Approved By: //# .,

L

1 -

3 -

g ;._ ..

l- SOIL ABSORPTIONl s' i LEACH FIELD i SEPTIC i

!?

TANK LIQUID

~

= '

~

(PRETREATMENT) TREATMENT-

=

SCUM * *

~

YANKEE PLANT r

=-

LIQUID e

,. WASTE 1NAl t-R ! I 1

SLUDGE

i l i i,i SEPTAGE !

i ON-SITE PROCESS l <sco Noto) :

! (LIQUIDS) l l l WASTEWATER i 5 OFF-SITE PROCESS =!

, TREATMENT !

, (SEPTAGE) j . FACILITY . !

Not0: Septago Hau!orfrankTruck Pipo Uno

~

f% YANKEE PLANT SANITARY WASTE DISPOSAL PROCESS

( ,

namm , -

Appendix B 1

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

)

Revision 11

' anure B-1 O

_ _ - - - _ _ _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - ~ - - - - - - - - - - - - - - - - - - --- - ~ ' ~ - ~ '

e Appendix B APPENDtX D TO fi20.1-20.602 --CONCENTRATIONS IN AIR AND WATER ABOvE NATURAL

- BACKGROUND a (see sootnotes at end of Appensa el isotopea Te64e 1 Tatde k -

EW (ad fiumber) g34 M2 Col. 2 gng, y4 4Wm0 gymn 4Wm0 gymn ActirAsm (09) Ac 227 S 2 x 10* " 6 x10" 6 x 10*" 2 x to-*

I 3 x10*" 9 x 10*

9 x 10*"

Ac229 S 3 x10**

8 x10's 3xgg s 3 x10" 0 x 10**

I 2 x10's 4 x10-s 6 x 10* "

Amerkman (95) Am 241 S 9 x10*

6 x10*is 1 x10" 2x10*" 4 x10*

l 1x10*

8 x10" 4 x10*"

Am 242m S 3 x10-8 6 x10*" 1 x10" 2 x10*" 4 x10**

i 3 x10*" 3x10's 9 x10 *"

Am 242 9x10*s S 4x10" 4 x10** 1 x10*

1 x10**

I 5x10-8 4 x10's Am 243 - S 2 x 10.*

1x10" 6 x10*'8 1 x10-* 2 x 10* ** 4 x10**

1 1x10"8 8x10" 4x10*" 3 x10" Am 244 S 4 x10** 1 xt0- 1 x10*' 5 x 10-'

i 2 x10** 1 x10*

6 x10*'

-. An5 mony Sb tt? 5x10-8 OV - S 8

2 x10*'

1 x10*'

8 x10**

6x10" 6x10" 3 x10-s 3x10-s Sb t24 S 2x10*' 7x10" 5 x10** 2 x10*

I 2x10-8 7x10" 7x10*"

Sb 125 S 2 x10**

' 5x10-8 3 x10*

2x10-e gxto-a Argort (18) l 3 x10-' 3x10" 9x10-"

A 37 1 x 10-8 Sub8 6 x10-s 4 x10**

A 41 Sub 2 x10**

hserac (33) As 73 4 x10**

S 2x10's 4 x10-8 7x10** 5 x10" 1 4 x10*' 1 x 10*

1 x10" 5 x10-*

As 74 S 3 x10" 2x10** 1 x10** 6 x10-8 1 1 x10*' 2x10's 4 x10** 5x10-8 As 76 8 1 x10*'

- 6 x10** 4 x10** 2x10-s l 1 x10':' 6x10" As 77 8 3x10-' 2x10**

5 x10-' 2x10** 2 x10-' 8 x10-s I 4 x10" Ex10-e 1x10-8 Astatine (es) At211 6 x10-8 S 7x10-* 5 x10-* 2x10*" 2x10**

Barksn (56) Ba 131 1 3 x10** 2x10-' 1 x10** 7x10-'

S 1 x10** 5x10-8 4x10" 2x10**

I 4 x10-' 5x10-8 1 x10** .

Ba 140 S 2 x10**

1 x10** 8 x10** 4 x10** 3x10 8 Se4eEuni(97) Bk 249 8 4 x 10*

7x10" 1 x10-' '2x10-8

S 9 x 10-" 2x10" 3 x 10-" 6x10" I 1 x10" 2 x10**

Bk250 S 4 x10" 6 x10**

1 x 10*' 6x10-s 5 x10** 2 x10*

l 1x10** 6x 10-s 4 x10" Beryhmt (4) Be 7- S 2x10"

6 x10** 5 x10*

2x10*' 2x10" 1 1 x10-* 5 x10-a 4 x10" 2x10" Beunutti (83) - Bi206 S 2 x10"' 1x10" 6 x10" 4 x t o-*

B1207 1 1 x 10** 1 x10" 5 x 10** 4 x 10-8 S 2 x 10" 2 x10" 6 x10" 6 x 10" 1 1 x 10-* 2 x 10*

6 x 10-" 6 x 10-'

Bi210- S 6 x 10" 1x10** 2 x 10 " 4 x 10**

. - t Revision 11 B-2 .

N ctor R:gul:t:ry C:mmissEn Pt. 20 [(( 20,1-20.602L App. B APPENDtx 0 TO $$ 20.1-20.602-CONCENTRATIONS IN Al1 AND WATE:1 ADOVE BACKGROUND-Continued (See tootnotes at end of Append.s B) 7 isotope '

Tatde i Table is Dement (atomic 'unber) Cot, 2-g34 Col.1 4 Col. 2 -

40/mQ [g*f,n 40/m0 Wa gg mQ I 6 x10-* 1 x 10-8 Bi tit 2 x 10* " 4 x 10**

S 1 x10" 1 x 10" 3 x 10" 4 x to-*

i 2 x 10*

1 x10*: 7 x 10" Bromne (35) Br 22 4 x 10" S 1 x10" 8 x 10'8 4 x 10" 3xto" l . 2x10" 1 x 10** 6 x 10" Cedmbm (4Q Cd 109 - S 4 x 10-8 5 x10" 5 x 10-8 2 x 10" 2 x10" -

I 7x10" 5 x10* 8 Cd 115m 3 x 10*' 2 x10-*

S- 4 x10" 7x10" 1x10" I 3 x to-*

4 x10" 7x10" 1 x 10

3 x10" e Cd 115 - S 2x 10-' 1 x 10-8 0 x 10" 3 x 10-8 Caldom (20) 1 2 x10" 1x10*8 6 x10" 4

' x 10-

Ce45 8 3 x10*

3 x10" 1 x 10 ' 9 x10-8

' 9 1 x10*' 5 x10-8 4x10" 2x10" Ca47 $ 2x 10-'

, 1 x 10*

6 x10" 5 x t0-8 Ca2fomium (98) 1 2 x10*' 1 x 10-8 6 x 10" 3 x10**

Q 249- S 2 x 10* ** 1 x10-* 6 x 10-" 4 x10-*

I 1 x10*" 7 x10**

Q 250 $ 3x 10"8 2 x10" 5 x107'8 4 x 10*

2x)O* ** 1 x10'8 O 251 S 8 1 x10-" 7x10" 3 x 10"8 3 x10**

2x10"8 1 x10** 6 x 10"* 4 x10-8 Q 252 -

1

1x10-" 0x10" 3 x10"* 3x10 s S 6x10* *8 2x10" I Px10-" 7 x t0**

3 x10* *' 2x10" 1 x10"8 7x 10**

O 253 - S 8 x10* " 4 x10*

3 x10-" 1 x10" l 8 x10*" 4 x10-8 3 x10-"

O 254- 1 x10**

7

) 1 S 6 x10*"

5x10**

4 x10-8 2 x10"8 t x to-'

CartxM (6) C 14 4 x ?0-* 2 x 10* ** 1 x t0" S 4 X10-4 2 x 10*'

(CCJ Stb 1 x10*' 8 x10**

Carbm (50) 6x10** 1 x10**

Ce 141 - S d x10-' 3 x10** 2x10" 0 xto-*

I 2x10*' 3 x 10*

Ce 143 S 6 x10-' 0x10-8 3 x10.-' l x10-8 9 x10" 4 x 10-*

Ce 144 1

2x10-* 1 x10" 7x10" 4 x10-8 S 1 x10-* 3 x10" 3 x 10'" t x10-8 Ceslum (SS) Cs 131-8 6x10" 3 x 10** 2x10 " 1 x10"

.___ S 1 x10-8 7x10-8 I

4x1V' 2 x t0*

3x10-8 3 x10-8 1x10" 9 x t0**

Cs 134m S 4 x10-* 2 x10" 1 x10** 6 x10" Cs 134- --

1 6x10** 3x10-8 2 x10*' 1 x10-s S 4 x10-* 3 x 10** 1 x10** 0x10-8 1

1x10** 1 x 10** 4 x10*" 4 x10-s Cs135 S E x10 3 x10-* 2x10" t x10"

l 9 x10-8 7x10'8 Cs 136 S 3 x10-' 2x10" 4x10% 2x10-8 1 r 0** 0 x10-8 4

2x10*' 2x10** 6 x s0** 6x10-s Cs137 S 6 x10-*

l 4 x10" 2x10**

2x10" t ,

1 x10-' 1 x10** E x10-"

CNorine (17) 0 36 S 4 x10**

4 x10**

2x10** 1 x10** 8 x10-8 0 38 4

2x10** 2x10-8 8 x10*" 6 xt0**

-. S 13x10-8 1 x10" I 0 x10-* 4 x10" Ctromium (24) 2x10-8 1x10-8 7x10 4x10" Cr 51 S 1 x10** 5x10-s 4 x to", ovton 1

- 2x10-4 E x 10-' 8 x 10** L ,10-8 Cobatt (27) CoS7 S 3 x10** 2 x t0** 1x10 ' 5 x10" I 2x10**

CoS8m S 1 x10* 8 6 x10" 4 x10" I

2 x 10-8 8 x10" 6 x 10-' 3 x 10*

9 x 10** 6 x10-8 3 x10" 2 x10" Co S8 S 8 x10-' 4 x 10-s oxion t xio" Co 60 1

5 x 10-* 3 x10" 2 x 10

9 x10" S 3 x10" 1 x 10" 1 x 10-* 5x10" l

9 x10*

1 x 10-8 3 x 10-"

C p (29, _

Cu 64 S 2 x10" 1x10" 3 x 10" 7 x 10" 3 x 10" 1

1 x 10-* 6 x 10*

4 x 10" 2 x 10*

Cur'A (96) . Cm 242 -- S 1 x 10* " 7 x 10*

4 x t0* n 2 x 10-s Revision 1I B-3

1 i

..; .- ' Pt. 20 ($ 20.1-20.602], App. 8 L

. . . 10 CFR Ch. I (1 193 Edidsn)

-(

- APPENotX B TO $$ 20,1-20.602-CONCENTRATIONS IN AIR AND WATER AeovE NATURAL..

BACKGROUND-Continued -'

, tsee sooerwi.e et ens of M el lootope

i Tetde I Talde li h (stomic stumt aq Col. 2-Col 1--Air Col 1--Ak Col. 2-4W50 9$fm0 4Ci/nq . y*g I- 2x10* " 7x 10**

Cm 243 S 6 x 10*" - 2 x10*

6 x10*" 1 x10" 2x10ns 5 x10**

1 1 x10*" 7 x 10" ,3 x 10"8 2 x 10* *

. Cm 244 - S 8x10*" 2x10" - 3 x 10*" 7x10**

1 1 x 10-" 8x 10** 3 x 10* " 3 x 10*

Cm245 S 5 x10*" 1x10" 2 x 10*" 4 x10** - -

I- ~ 1 x10*" 8x10" 4 x 10*" 3 x10" Cm 246 : S 5 x10*" 1 x10" 2 x10* " 4 x10-8 I i 1 x 10*" O x10" 4 x 10*"

L+

Cm 247 8 3x 10** '

-5x10ns 4 x10" 2 x10*" 4 x10-*

t 1x10** 4 x10" 4 x 10*

Cm248 S 2x10 s 6x10-" 1x10" 2x 10-" 4 x10" +

1. 1x10*" 4 x10*

Cm 249 4 x10-u 1 x10*8 S 1 x10-' 6x10*8 4 x 10" 2x10*:

Dyepreshan (88) 1 1x10-s 6x10-s 4xiO-' 2 x10's Dy 106 S 3x10** 1x10" 9 x 10-* 4 x10-4 I _ 2x10** 1x10's yxio-s Dy 106 S 4 x10* * '

2x10*' 1x10-s 8 x10-' 4 x10** ;

I 2x10*' 1 x10** 7x10**

Einsteinhan (se) Es 253 S 4x10" '

8 x10-# - 7x10" 3x10*H 2x10" 1 6x10*" '7 x10" 2x10 " 2 x10**

Es 254e= S 5x10*' 6x10** 2 x10* " 2x10-8

i. 6x10-* 8x10" 2x10*" 2 x10**

Es 254 S 2x10-"

- 4x10" 4x10"' 1 x10-* '

l' 1 1 x1*

4 x10" 4 x10-" 1x10

Es 355 S E x10*" e x10** 2x10-u 3xio-s

~ _

Estphan (GO) l 4 x10*

8x10" _1 x 10*" 3x10-s

- Er100 S 6x10*' 3x10" i 2 x10** 9x10**

4 x10*' 3 x10-8 1 x10** 9 x10**

Er 171 - S 7x10*' 3 x10-8 2x10" 1 x10**

I 6x10-' 3x10-s 2x10**

Ewoplwn (83) Eu152 S-1x 10-*

4 x10** 2 x10** 1 x10** 4x10-*

(T/2=92 tus)~ l 3x10-' 2x10-8 1x10 8 -.6 x10*

Eu 152 S ,1 x 10** 2x10-s (U2=13 yrs) 4 xio-= ex10-s I 2x10** 2x10-8 Eu 154 3 6 x10* " - 4x10"

' - 4 x10-' ex t0" - 1 x 10*" 2 x10**

I 7x10-' 6x10** 2x10*"

Eu 155 . 2x10**

$ 9x10** 4x10-s 3xio-o 2x10" l 7x10** 6 x10* * - 3 x 10** - 2x10" leermlun (100)* Fm 354 S 6x10**

, 4x10-s 2x10-* 1 x10-*

I 7x10** 4 x10-s - 2x10-' 1x10**

Fm 355 S

2x10**

1-1 x10-8 6x10*" 3 x 10-*

1x10-* 1 x10** 4 x10-" 3x10**

Fm 256 S 3x10** 3x10** 1 x10*" 9 x 10*'

I

2x10** 3x10-8 6x10*"

Noeine (s) F 16 S. 9x104 Ex10-* 2x10-' 2x10-' e x10-*

t 3x10**

GedoEnlym (64} Gd 183 S

1 x10** 9x10" . 5 x10**

2x10*' - 6x10-8 8x10" 2x10" I 8 x10** 6x10-s Gd 150 S 3 x10-' 2x10** -

- 5x10-' 2 x10** 2 x10** 8 x10"'

t . 4x10-7 2x10-'

Geuium ('31) Ge 72 .1 x10-8 - 8 x 10**

S 2x10" 1 x10** ' ext 0-' 4 x1C*

1 2x10*' 1x10's 4 x10-' 4 x10*

Germannan (32) Ge 71 S 1x10** . 5x10** 4 x10-' 2 x 10*

i 6 x10** 5x 10** - 2x10**

.L Gold h9) Au 188 ' S 2x 10-a ;

1x10" 5 x10** 4 x10" 2x10" l 6x10" 4 x10-8 2 x 10** 1 x10" Au'198 S 3 x10-' - 2x10*s ixion sxto-s t 2 x10" 1 x10-s Ao 199 S e x t0** 5 x 10-*

1 x 10-' 5 x10*

4 x 10** 2 x10"

.i* . I 8 x10" 4 x 10*:

Hafman (723 Hf181 S 3 x t0** 2 x 10"

- -- - 4 x.10" 2 x10" 1 x 10*

7 x 10*

t 7 x 10" 2 x 10**

Hosmium (67) Ho 166 - 3 x 10*

7 x 10*

S 2 x 10" 9 x t'J" 7 x 10-

3 x 10"

- Revision 11 - B-4

, , , , ., ,-.,w- = ~ '

______. _ -----_ _ e -

i

~ ~~~WI59tEAi-80402), App. B APPENDtX B TO $$ 20.1-20.602--CONCENTmJsONS IN AIR AND WATE7t ABOVE fJATU BACXCROUND-COnlinued

. (See toottetes et end of Apperda B)

I Isotope

Tetse i Tatde It Eiement (etomic nwice4 cet 1- n cota- c ,. , ,,,,;, cota-( Ci/mn k

4Cdmq [g*j'g' Hydrogen (t)- i 2x10" 9x10" H3 6 x 10" 3 x10" S E x10" l

1x10" 2x10" 3 x10" 6x10*8 1 x 10" 2 x 10*' 3 x 40* 8 j, indum (49) -

$4 2 x 10** 4 x 10" in 113rn S 8 x 10** 4 x 10*' 3 x 10*' 1 x10" I 7 x10** 4 x10's In 114m - S 2x 10*' t x t0** .

1x10" 5x 10" 4 x10" 2x t0" 1

i 2x10" 5 x 10*

D 115m S 2 x 10**

7x t0!" 2 x 10-*

1 x 10" 8 x10" 4 x10" i 2 x 10**

1n 115 - S 1 x 10*

6 x10" 4 x10**

2x 10" 3 x10*8 9 x 10" 9 x10-8 I 3 x10" 3x10 8 lodine ($3)_ . (125- S 1 x 10** 9 x t0" 6x10" 4 x 10-8 8 x10*" 2 x 10*'

(126 1

2x10-' 6 x 10*

6 x 10*' 2 x 10* *

,- S 8x10-'

' 6 x 10** 9 x10* " 3 x10" I 3 x10*' 3 x 10**

1129- S 1 x t0" 9 x10*:

2 x10" 1x10" 2 x 10*" 6 x10" I 7x10** 6x10" (131- S 2 x10** 2x10" 9 x10-' 6 x 10** 1x10*" 3 x10" l 3 x10** 2 x10" 1132 S 1x10" 6 x10-s 2 x10*' 2x10" 3 x10" 8 x t0**

I 9 x10*' 5 x10's gxto.e (133 S 2 x to" 3x10" 2 x 10** 4 x10-" 1 x to-*

I 2 x10*'

(134- 1 x10-8 7x10** 4 x10**

S 5x10" 4 x10** 6x10" 2 x 10**

I 3 x10** 2 x10-8 h ( 135 S 1 x t0*' 6 x t0**

I 1 x10*' 7x10" 1 x10** 4 x 10**

Iridium 971 1:190 4 x10*' 2 x10-' 1 x10** 7x10"

S 1 x10**

i 6x10" 4 x10" 2xto" tr192 4 x10*' 5x10-s gxto-e 2 x10*

S 1 x10** 1 x10-8 4 x10** 4 x10**

I 3 x10**

Ir 194 1 x10-8 9 x10-" 4 x10**

S 2x10-1 1 x 10-* 8 x10-' 3 x to-*

I 2 x10*'

tron (26) Fe 55 S 9x10" 5 x10*' 3x10" i

9 x10-* 2x10" 3 xt0*

8x10" 1x10-8 7x10*a 3x10*

Fe 59 S 2x10* *

. 1 x10-' 2 x10*

5 x10** 6 x 10-*

1 Krypton (36)

Kr 65m Sub 5 x10** 2x10" 2x t0*' 5 x 10**

Kr 8C 6 x10** 1 x10-8 Se 1 x 10**

Kr 67 S4 3 Xt0*'

Kr 88 1 x10** - 2x10-'

Sub 1 x10-*

Lanthanum (57) La 140 S 2x10-*

2 x10-' 7x10** 5x10 ' 2xto-*

Lead (62) Pb 203 1

1x10-' - 7x10-* 4 x10** 2 x10**

S 3 x10**

1x10*8 9 x10*

4 x10-*

I 2x10**

Pb 210 - 1 x10*

6 x t 0** 4 x10" S 1 x10*

I 4 x10" 4 xt0-# 1x10" Pb 212 2x10*" 5 x10-' 8 x t0* 88 2x10"

S 2x10-*

I 6 x10-* 6xt0*" 2xto-*

IJutetium pt)-

Lu 177 2x10-* 5x10" 7x10-" 2 x10**

S 6 x10-' 3x10". 2x10-8 1 x10**

Manganese (25) 1 E x10*' 3x10" 2 x10-'

' Mn52 S 2 x10-' 1 x10" 1x10-8 7 x10-' 3 x10"

) 1 x10-' 9 x10-*

Mn54 ~ S 5 x 10-' 3 x 10**

4 x10-' 4 x 10 1 x10*e l 4 x10** 4 x10" Mn56 - 3 x10-8 1 x 10-' 1x10" S 8 x10-' 4 x10-a Mercury CO - l 5 x10-' 3 x10" 3 x10-8 1 xt0" Hg 197m S 2x10" 1 x10" 7x 10-' 6 x10*

3xto" 2 x 10**

I 8x10-'

Mg 197__ 5 x10-8 3 x10" 2 x10**

-h . S 1 x10-8 9 x10-a i 4 x t0*

3x10" 1

3 x10** 1 x 10* 8 9 x t0-8 Hg 203- S 7x10" 5 x t0" 5 x t 0" 2 x 10" 2x t0**

1 1 x10" 3 x 10*

4 x 10-' 1xto" Resision 11- '

B-5

. .- Pt 20 ($$ 20.1-20.602), App. B 10 CFR Ch. I (1-1-93 Edition)

APPENDIX B TO $$ 20.1-20.602-CONCENTF1ATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND-Continued (See footnotes at end of /pperda B) isotope

Table i Tatde it Element (stomic numbe<l g,w Col.2- Col. 2-g 3_a

. ( Wmq -yk 4Cumn gg Mofytidenum (42) Mo 99 5 7x10

5 x10-8 0x 10" 2 x10" I 2x10-8 1 x 10* 8 7 x10" 4 x 10"

-" _ f, ;M. (60) Nd 144 S 8 x10-H

2 x10-8 3 x 10"8 7x 10" I 3 x10-" 2x10-s 1 x 10-" 8 x 10*

Nd147 $ 4 x10*' 2 x 10*

1 x 10" 6 x 10" I 2x10" 2 x 10-8 8x10" 6 x 10-*

Nd149 S 2 x 10** 8 x 10" 6 x10" 3x10"

1 1 x 10-* 8 x 10*

5 x 10" 3x10" Nepturnum (93) Np 237 $

' 4 x 10*" 9 x 10*

1xtous 3 g iou 4 1 x 10*" 9x10" 4 x10*" 3x10" Np 239 $ 8 x10-' 4x10 s 4 4 x10" 1 x10"

4 7x10-' 4 x10*

2 x 10*

1x10"

- Nicdel(28) Ni S9 S 5 x10-' 6x10-3 2x10" 2x10"

, t 8x10-' 6x10*8 3x10" 2 x10-8 N;63 S 6 x10** 8 x 10** 2 x10" 3 x10" l 3 x10*' 2x10 ' 1 x10" -7x10" Ni 65 S 9x10*' 4 x 10** 3 x10-* 1x 10" 1 5 x10-' 3 x10-8 2 x 10** 1 x 10" Notuum (Columbium) (41) Nb 93m S 1 x10** 1x10-8 4 x10-* 4 x10" I- 2x10*' 1x10- $ x10** 4 x 10"

, Nb95- S 6x10-' . 3 x10-8 2x10-8 1 x10" I 1x10" 3 x10-8 3 x10" 1 x 10" Nb S7 S - 6x10-8 3x10 s 2 x 10-' 9x10" l 5x10** 3x10-8 2x10" 9x 10*

Osmum ps) Os 185 S 5 x10-' 2x10*: 2x10-8 7 x 10* *

, @ Os 191m I

S 6x10" 2x10" 2x10-8 7 x10-*

2x10" 6 x10" 7x10-8 3x10" l 9 x10-* 7x10-8 3 x10" 2x10-8 Os 191 S 1 x10-* 6x10-8 4 x 10-* 2 x10*

I 4 x10-' 5 x10*s 1x13" 2 x10" Os 193 S 4x10-8 2x10-8 1 x 10-8 6x 10*

I 3 x10-' 2x10-8 9x10" 5 x 10*

PaRadium (46) Pd 103 S 1 x10** 1 x10-8 - 5xto" 3 x to" l 7x10-' 8 x10-8 3 x10" 3 x10" Pd 109.,_ S 6 x10-' 3x10-8 2 x 10*

9x10-8 Phosphorus (15) -

1 4 x10" 2 x10-' 1x10-8 7x10-s

P 32 S 7x10-8 5x10-* 2x10" 2x10-8 I 8 x10-* 7x10" 3x10" 2 x10-*

Platinum (78) Pt 191

_ S I

8xtr? 4 x10-8 3 x10-8 1 x 10-*

6x10-1 3 x10-8 2x10-8 1x10" Pt 193m S 7x10" 3 x10-' 2x10-' 1 x10-s I 5x10** 3x10-8 2x10" 1 x10**

Pt 193 S 1 x10-* 3 x10*' 4 x10** 9 x to-*

I 3 x10-' 5x10-s 1x10-8 2x10" Pt197m S 6x10-e 3x10-8 2 x10-' 1 x10-8 .

I 6 x10** 3x10-s 2x10" 9x10" Pt 197 S 8x10" 4 x10-s 3 x10-* 1 x10" l 6x10" 3 x10-8 2x10-8 1 x10-#

Plutornum (94) Pu 238 S 2x to-88 1 x10-* 7x 10*" 5x10-e 1 3 x 10- 8 x10** 1 x 10-" 3 x10-8 Pu 239 S 2x10-"

> 1x10" 6 x 10-** 6 x10" l 4 x10-n 6 x10-* 1 x10*" 3 x10" Pu 240 S 2x10"*

l

1 x 10-* 6x10-" 5x10" 4 x10"' 8 x10" 1 x10*" 3 x 10-*

Pu 241 S 9x10"' 7x10-s axio-n 2x10" 8 4 x10** 4 x10*

1 x10** 1 x10" Pu 242 S 2x10-" 1 x10" 6 x 10"* 5 x 10-

1 4 x10* n 9 x10" 1 x 10"- 3 x 10" Pu 243 = S 2 x10** 1 x10-8 6 x 10" 3 x 10" )

, b Pu 244 I

S 2 x10*

1 x 10" 8 x 10-* 3 x 10" ;

2 x 10* " 1 x 10" 6 x 10"* 4 x 10"

. ," ~

8 3 x10-" 3 x 10** 1 x 10"' 1 x 10"

)

i Polonium (64) Po 210 S 5 x 10* " 2 x 10-8 2 x to' " 7 x 10" I 2 x 10"* - e x 19" 7 x to-" 3xt0" ]

3 I

Resision 11 B-6

_ _ _ _ - - _ - _ - - - - _ - -_ --- ~~

__ _7 PP.B APPENDfX B TO ff 20.1-20.602--CONCENTRATIONS IN AIR AND WATER ABOVE NATUR BACKGROUNO---Continued (See loopwtes tt end of Apperes 83 Isotope 8 - Tatde i Tatse u Enoment latomic numbe4 -_

g ,_u Col 2- Cd1-h Cas. 2-

_40/mn g[of,g 40/mn gyf%

Potassium (191 K42 S' 2 x 10" s x to'* 7 x 10" 3 x to" l 1 x 10" 6 x 10" j Preseodymans (59) -

Pr142 4 x 10" 2 x 10" !

S 2 x 10" 9 x 10" 7 x 10" 3 x 10" l 2 x 10** 9 x 10*

5 x t0** {

Pr14L. 3 x 10*

S 3 x 10" 1 x t0" )

1 x 10" 5 x 10* * '

Promesuum let) 1 2 x10" 1 x 10" ' 6 x t0" 5 x 10" Pm 147- S 6 x t0" 6 x 10*

2 x 10" - 2 x 10" l 1 x10" 6 x 10" 3 x 10" Pm 149 S 2 x 10" 3 x 10" l x 10" 1x10" 4 x10" i 2 x 10* ' l x 10**

Protoachraum (91) Pa 230 8 x 10** 4 x 10**

W S 2 x10" 7 x 10" 6 x 10"* 2 x10" l 8 x 10* " 7 x 10*

3 x 10*"

Pa 231- S 2 x 10*

1x10"8 3 x 10*

4 x 10*" 3 x10" l- 1 x 10*

Pa 233 8 x 10*

4 x 10"8 2 x 10* *

, S 6 x 10*' 4 x *0-8 t

2 x 10" t x t0" Redum (84) 2x 10*' 3 x10* 8 6 x 10" 1 x10" Ra223- S 2 x10*

2 x10" 4 x 10* " 7 x 10" l- 2 x10* "

Ra 224 -

1 x 10*

8 x 10*" 4 x 10**

S 5 x10" 7 x 10*

I 2 x 10*" 2 x10" 7 x 10* " 2 x10" 2 x 10*" 5 x t0**

Ra p26 - S 3 x 10* " 4 x10" 3 x 10*" 3 x t0**

I 5 x 10*" 9 x10**

Ra 228 2 x 10* " 3 x t0*

S 7 x 10* " 8 x 10*'

l 2x10"' 3 x t0" Redon (86) 4 x 10*" 7 x10*

1 x10*'8 An220 S 3 x 10*

3 x10*

1 x 10**

Rn 222

3 x10" Rhenbm (75! Re 183 - S 3 x10"

- 3 x 10** 2 x10" 9 x 10** 6 xto"

[

1 2x10*' 8 x10" 6x10" 3 x 10**

Re 186 S 6 x10*'

i 3x10*8 2x10" 9 x10"

{ l 2x10" 1 x10** 8 x10*'

Re 187 S 5 x 10*

9x10" 7 x10** 3 x10" 3 x 10*

I 5 x 10** 4 x 10*

Re 188 $ 2x 10** 2 x10*

4 x10" 2x10" 1 x10** 6 x 10*

I 2 x 10*' 9 x10" Rhodum (45) Rh 103m S 6 x 10" 3 x 10*

8x to-* 4 x 10*

3 x 10** 1 x10**

l 6 x10** 3x 10" Rts 105 - S 2x10" t x10"

, 8 x 10-' 4 x10" 3x10 8 1 x10" l 5 x10*' 3 x10" Rutmfum (37) Rb86 S 2x 10** 't x10**

3 x 10* *' 2x10-a 1x10" 7x10*

Rb 87 8

7x10-* 7x10-* 2 x10-' 2xt0**

8- 5 x10*' 3 x10-s I 2 x10-* 1 x10**

Ruthenium (44) - Ru S7 7x10-' 5 x10** 2 x10-' 2x10" S- ~ 2x10** 1 x10**

I 8 x10*

4 x t0" 2x10-* 1 x 10** 6 x10** 3 xt0" Rut 03_. S 5 x 10** 2x10** 2 x10-* 6 x10" l 8 x 10*

2x10-8 Ru t05 - S 3 x10 ' 8 x10**

7x 10-' 3 x10-8 2x10-8 1 x10-*

1 E x10-' 3 x 10** 2 x10-* 1 x10-*

Ru t06 S 8 x 10**

. 4x10" 3 x10** 1 x t0-*

I 6 x10-' 3x 10**

Samedum (62) Sm,147 S 2 x10*" .1 x t0*

7 x10* 8' 2x 10** 2x10'u I 6 x to-*

3 x10*" 2 x10" e x t o-" 7 x 10*

Sm 151 S 6 x10-*

1 x10** 2 x10-* 4xto"

'

i 1 x10" 't x10**

Sm 153 5 x10*

4 x 10**

S 5 x10*' 2 x 10-8 2 x 10*

8xto*

1 4 x10" Scandium (21) 2 x 10** 1 x10"* 8 x10*

Sc46 S 2x10" 1 x 10** 8 x 10** 4xto

l ,2 x 10-' 1 x 10" 8 x 10* "

Sc 47 S 4 x 10*

6 x 10" 3 x 10" 2x t0" ' 9 x 10-*

I 5 x 10** 3 x10* 8 2 x 10*

9 x10" 4 Se de. ~. S 2x 10-' 8 x 10" 6 x 10" 3 x t0-*

t 1 x 10* ' ex t0** 5 x 10*

Seierem 341 -

Se 7 5_._ S 3 x t0**

_. 1 x 10" 9xto" 4 x 10" 3 x t0" g

t x 10-' 8 x 10*

4 x t0** 3xto" Revision 11 B-7 4

U.

Pt 20 [$ 20,1-20.602), App. B 10 CFR Ch.1 (1-1-93 Editi:n)

APPENoix 0 T3 fl 20.1-20.602--CONCENTRATIONS IN AIR AND WATER A:0VE NATURAL BACKGROUNO--Continued (See footnotes at end of Appendia B)

Isotope

Table a f abie Is C*nent (alomic numbef) gotg_g Col 2- Col 2-Col.1--h 4C./mi) W*%

g 4C/M g, SW (14) Tw 31 S 6 x 10-4 3x10. 2 x 10" 9 x 10" I 1 x 10-* 6 x 10*

3 x 10" 2 x 10" Saver (47) Ag 105 S 6 x 10*' 3 x10" 2 x 10" 1 x 10" I 8 x 10" 3 x 10-8 3 x 10-'l 1x10" Ag 110m S 2 x 10" 9 x 10-

7 x 10" l 3 x t0" i 1x10" 9x10" 3 x 10-" 3 x 10" Ag 111 S 3 x 10*' 1 x 10-8 1 x 10" 4 x10*

I 2 x 10*' 1 x 10" 8x10" 4 x10*

Sodum (11) Na 22 S, 2x10" 1 x 10* 8 6 x 10" 4x10" I 9 x 10** 9 x 10*

3 x 10* " 3 x10*

Na 24 S 1 x 10*

6x10" 4x10" 2 x 10" I 1 x 10-' 8 x10** 5x10" 3 x 10*

Strontium (38) Sr85m S 4 x 10" 2 x 10" 1 x 10-8 7 x10*

t ,

3 x10-* 2 x10" 1 x 10** 7 x10*

Sr 85 S 2 x10" 3 x10* 8 8 x 10-' 1 x 10" i 1 x 10*' 5 x 10-8 4 x 10*' 2 x 10" Sr89 $ 3 x10" 3x10" 3 x 10*" 3 x 10-*

I 4 x10" 8 x 10-* 1 x 10" 3 x10"

$r 90 S 1 x 10** 1 x 10-8 3 x10-u 2 x10" l 5 x 10-' 1 x10-8 2x10 " J r 0**

Sr 91 S 4 x10-' 2 x 10-8 2 x 10*

7 v' 0" 1 3 x10*' 1 x 10-8 9 x10" 5 x 10-*

Sr st S 4 x10*' 2 x10*

2x10" 7 x 10-*

I 3 x 10*' 2 x10-s 4x10" 6 x10-*

i Suttur (16) S 35 S 3 x 10*' 2 x t o-* 9 x t o-' 6 x 10-*

I 3 x10-' o x10** 9 x10-8 3 x 10* *

,, Tarttalum (73) Ta 182 S 4 x 10*

1 x 10** 1 x10" 4 x 10**

I

'

2x10" 1 x10" 7x 10*" 4 x10**

TechneGum (43) Tc96m S 8 x10*

4x10" 3 x10" 1 x10*

I 3 x10** 3 x 10*

1 x10-* 1 x10-8 TcD6 S 6 x10*' 3x10 s 2 x10-8 1 x10*

I 2 x10-' 1 x10* 8 x10" 5 x10" l Tc 07m S 2x10** 1 x10*

8 x10" 4 x 10" i 2 x10*' 5 x10*

5 x10** 2 x 10**

Tc S7 5 1 x 10** 5 x 10-' 4 x 10* ' 2 x 10*

I 3 x10" 2x10-8 1 x 10-8 8 x10" Tc 99m S 4 x 10** 2 x10'* 1 x10-* 6 x10" 1 1 x10** 8 x10*

5x10" 3 x10-s Tc 99 S 2 x10-8 t x10** 7 x10-8 3 x10"

, t 6 x10-8 5 x10-a 2x10** 2 x10**

Tenudum (52)'

?* 125tr S 4 x10-' 5x10 a 1 x to-* 2x10" 1 1 x10*' 3 x 10** 4 x10** 1 x10**

Te 127tr S 1 x10-' 2x10" 5 x10-* 6x10-8 I 4 x10** 2 x10** 1 x10-* 5 x10**

To 127 S 2 x10** 8 x10** 6x10-8 3 x 10*

I 9 x10*' 5 x10-a 3 x10" 2x10" To 129er _

S 8 x10** 1 x10" 3 x10" 3 x to" 1 3 x10-' 6 x10** 1 x10-* 2x10"

Te 129 S 5 x10** 2 x10*

2 x 10-'

, 8 x 10**

I 4 x10** 2x10-8 1 x 10*

8x10" Te 131re S 4 x10* ' 2 x10-8 1 x10** 6 x10*

1 2x10" 1 x10" 6 x10" '4 x10-8 -

Te 132 S 2x10*' 9x10" 7x10" 3 x10*

1 1 x 10*' 6 x10" 4 x 10" 2 x10-*

Terbium (65) Tb 160 S 1xto" 1 x t o-* 3xto" 4 x10" l 3 x 10-' 1 x10-a i xion 4 x 10-s Thallum (61) il 200 S 3 x10** 1 x 10-' 9 x 10-

4 x 10" I 1 x 10'* 7x10*a 4xio-a 2 x10" Tl201 S 2 x 10*

9 x 10-s yxto-. 3 x 10" t 9 x 10-' 5 x 10-' 3 x 10-* 2 x 10"

.- T1202 S 8 x 10" 4 x 10" 3 x109 1 x 10" 1 2x10*' 2 x 10* 8 8 x 10" 7 x 10"

(* "\

/ T1204 S 6 x 10*

3 x 10-a 2 x10" 1 x 10" 1 3 x 10*

2 x 10- 8 9 x 10* " 6 x 10" Revision 11 B-8 I

1

.-----, - - Mdn Pt. 20 [@@ 20.1-20.602), App. B APPENO x 0 TO ff 20.1-20.602--CONCENTRATIONS IN AIR AND WATER AeOvE fJATURAL BACKGROUND-Continued (See tooux>tes at end of Appenda 0]

tauope

Table 1 Table it Dement (atomic numbed g q ,,,y Col 2- g 9,,_g Col 2-4M (p h 4W (pd[mt)

Trerium (90) Th 227- S 3 x 10*" 5 x 10" 1 x 10* " 2 x to" Th 226 --

1 2 x10* " 5 x to" 6x10"' 2x t0**

. . _ S 9x10"" 2 x 10*

3x t0*" 7 x 10-*

I 6 x 10* " 4 x 10*

Th 230 2 x 10*" t x t0" S 2 x 10"' 5x 10-* 8 x 10*" 2xt0" t t x 10-" 9 x 10*

Th 23t 3 x 10"8 3 x10" S 1 x 10*

7x10" 5 x 10-* 2 x t0" I 1 x10" 7x10 s Th 232 _ 4 x 10" 2 x 10*

S 3 x10*" 5 x 10-*

t 1 x10"* 2 x t0"

' 3 x10-" 1x10" 1 x 10-" 4 x t0**

Th naturat -. S 6 x 10-*' 6 x t0** 2 x 10* " 2 x 10-*

I 6 x 10* " 6 x 10" Th 234 S 2x 10* " 2xt0" 6 x10" 5 x10" 2 x10" 2 x t0" l 3 x10" 5 x 10*

C 1huGum (69) Tm 170 S 1x10" 2 x 10**

4x10" 1 x to'8 1 x to" 5 x to" I 3 x10** 1 x10" Tm 171 - S 1x10" 5 x10*8 1 x10" 1x 10" 4 x10" 5 x10*

I 2x 10-' 1 x 10*

Tu' t (50) Sn tt3 S 8 x10" 5 x t0**

4 x10** 2x t0-8 1 x10" ext 0" l 5 x10** dx 10**

Sn 125 S 2 x10" 8 x 10*

1 x10-' 5 x10-* 4 x10*

2 x10*

I 8 x10** 5x10" Tungsten (Woerrang g4) W 181 S 3 x10" 2 x 10-*

2 x10-* 1 x10-' 8xt0" 4 x10-*

1 1 x 10-' 1x10-8 4 x 10-'

W 185 S 3 x10**

8 x10-' 4 x 10-* 3 x10-* t x t0-*

I W 187 --

t x10-' 3 x10" 4 x10" 1 x 10**

S 4 x10" t

2 x 10** 2x10" 7x10" thanum (9M 3 x10-' 2 x 10-' 1 x10" 6 x10-8 U 230 S

,

3x10-" 1 x10-* 1 x10-" 5x10" UW 1

1x10*" 1 x10-* 4 x10-'8 5 x10-*

S 1 x10-"

I 8x10" 3 x 10-" 3 x 10* *

. 3 x10-" 8 x10*

0x10-u 3 x10-e U 233 S 5 x10-" 9 x10** 2x10 " 3 x t0-8 U 234-1 1 x10-" 9 x 10*

4 x 10* " 3 x 10**

S* 6 x10*" 9x10-* 2x10*" 3 x t0**

U 234 1 1 x10-" 9 x 10*

4 x10"' 3x10 s S* 5x10*" 8x10

2x t0*" 3 x10-8 U 236 -

8 1 x10-" 8 x10" 4 x10* *8 3 x10-s S 6 x10*" 1 x10** '

t 2 x10*" 3x10" 1 x10*" 1x10-s 4 x10-u 3 x10**

U 238 Sa 7x10-** 1x10-8 3 x10-" 4 x10**,

1 1 x10*" 1x10-8 5 x10*"

U 240 S

4 x10-8 2x10*' 1 x10*

8 x t0** 3 x10**

I 2 x 10*'

Umatural . S* 1x10-8 6x10" 3x10-8 1 x10-" 1 x10" E x t o-*8 3xto-s

Vanadum (23) 1 1 x 10-" 1 x10-a 5x10"*

V 48 S 3 x10-s 2 x10*' 9 x 10-* 6x10" 3 x t0-*

I 6x10-8

Xeon (54) Ke 131m S-h 8 x10" 2x10-* 3 x10**

2x10-8 4 x10*!

Xe 133 Sub 1 x10** 3xt0-'

Xe 133m Sub 1 x 10** 3 x t 0-'

Yttertaium p0)-

Xe 135_ - Sub 4 x10**

Yb 175-- S 1 x10*'

, 7 x 10*' 3 x 10 2x10" t 1 x t0**

Yttrium (39) Y 90 6x t0-' 3 x10" 2 x 10" 1 x 10-*

S 1 x10-'

1 6 x 10" 4 x10-' 2 x 10**

1 x10*' 6 x10" 3 x 10*

Y 91% S 2 x10-* 1 x10" 2 x 10*

l 8 x 10" 3 x 10" 2 x 10** 1 x 10" 6 x 10* '

Y 91 - S 3 x 10-8 4 x 10" 8 x 10" t x t0" 1

3 x 10*

3xt0" Y 92 8 x 10"

1 x 10-* 3xt0" S 4 x10" 2 x 10" 1 x to-* 6 x 10**

t 3 x10" 2 x 10-a Y s3 S 2x10" 1 x 10 *

6 x 10" 8x to** 6 x t0" 3x10" i 1 x 10" 8x to" 5 x 10" 3 x 10" Resision 11' B-9

, Pt. 20 [$$ 20,1-20.602), App. B 10 CFR Cit. I (1-1-93 Edificn)

APPEN')lX B TO fl 20.1-20.602-CONCENTRtTIONS IN AIR AND WATER AeovE NATURAt.

BACKGROUNO- Continued (See footnotes at end d Appenda B) anotopoe Table i Tabie II Eiernent (atomic nurecert Cd 1-4. I Col. 2- Cd 1--Ai, Col.2-N'"9 6 m0 b "O 4C m0 Ise (30) 2n 65 S 1x10*' 3xiO** 4xty' 1 x 10" 1 6 x 10*

5 x 10*

2 x 1C 2 x10" 2n 69m S 4x10" 2x10 a 1 x 19" 7 x 10" 1 3 x 10* ' 2 x10-s 1 x 10*

6 x10" A 2n 69 S 7 x10-* 5 x10-e 2x10" 2x10 a i 9 x 10-8 5 x 10** 3 x 10" Zircorsa (40) Zr 93 2 x10'8

$ 1 x 10*' 2x10" 4 x to" 8 x10"

( 3 x 10*' 2 x 10*

1 x 10'* 8 x10"

Zs 95 S 1 x 10* ' 2 x 30" 4 x 10" 6 x 10"

, t 3 x 10** 2x10" 1x10" 6 x 10*

Zr 97 S 1x10 5 x 10* ' 4 x 10" 2 x10" Ary sirgie redonuclide not Esied abow l 9 x10." 5 x10" 3 x10" 2 x 10* *

Sub 1 x10** 3 x 10**

wth decay mode tener tien alpha esimesion or spontaneous sesion and with redoectree hen 4fe less than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . I Any singie radiertuclide not lisied abtve 3 x10** 9 x 10** 1 x 1C"' 3 x 10'*

tudIh decay mode other then alpos ornission or sporteneous fission ersd w4th radoecew hen 48e greater than 2 tons.

Any single radionuclide not listed above, whidi decays tg alpha emie.

6xtona 4 x10" 2 x 10*" 3 x 10*

sian or spontaneous esehn -

A i 'Soldie (S); basea (1).

S4 means that welues sien are ser 44mersbn in a semisphericalinGnita caoud of astxxne metetet .

'These endon concentracons are approiwtate tor protecton from redorW222 comtaned wah as shor14=d daugws.

cornbhation of shor14et' redor>222 daughie s, poloruurN!18. lead-214. txamuth 214 arwi po wtqhout regard to sie degree of equlEbnunt that wet result h the uttnate emission of 1Jx10

MeV cf alpha particle eneeg.)

The Table It weiue may be sopiaced by one thirtseth (%e) of a *worWng. novel" The Emit on raden 222 concentrations in restricsed areas may be bened on an arruel avem0e.

. *For vh.soluble L. .,minneros of U-2,8. U-234 and 0,235 in air chenucal toxscity may be the Emitino factor, if the percent ty of U-235 is less than 5. the cor centratiori watue for a 4 pour werkweek. Table't is 0.2 treligrams uruniuri per cubic meter of air avera0s, For any enrichment, the prodJct of the average concetration and tune of exposure d.Jn'no s 40 hour4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months workweek shes act exceed 8x10** SA plMv/od, where SA is the specif.c acGwity of the uranium hhaled. The concentration welue for Tacio it is 0.007 mluigrarna nium per cut >ic meter of air. The specific a ' for natural uranium is 6J7x10-' curies per gram U. The SA=3.6x10"cunes/grarn u specnic scevity for other mixtures of U-238 tA235 and U.,234. i. ed knomt shalt be:

U<iepieted

SA=(OA+0.38 E+0A034 E') 10** Et0J2 ,

where E is the percentage by weight of U-235, expressed s.s percent.

, this Appendbt should be daiermined as'lonows:Nott in any case where there is a mixture in air or water et more than one redo ,

1. If the idenuty and concentration of each redonuclide in the oixture are known, the Emiting values shoulci be derived as totlows: Detennine for each redonuedide h the mixture, the ratio netween the quantity present h the mixture and the Emit otherwise estatdehad in Appendst 8 for the spoofsc radionucEde vn.en rW h a mixture. The sum of such ratios for aR the

radorwedae in the mixture may not exceed "1- (i.e.

unity") .

and MPC and MPCe i=;--L-Q. then the concentratione shaR b4Exas.sPLc tt r6dionucdidos A. B and C are present h kmhed so that the tohowing relationship esists:

(C,mPC )+(C MPC.)+(CeMPCc) 51 of 2. If olther Appendix B shanthe klentity be: or the concentration of any radonuciade in the mixt te is not known, the EmitinD values tsr pr.rposes *

a. For purposes of Table t. Col.1 -6 x 10"*

b. For purposes of Tabie I. Cot. 2-4 x10"

c. For purposes of Table ti. Col.1-2 x 10'"

6. For purposes of labie it. Col 2-3 x 10**

3. If any specif*ed at paragraph of the cordtions 2 above.specisied below are met, the correspondng values sps%ed below may be used in Geu of those

a. If the isentity of each radionuclde in we mixture is known but the concentratior of one or more of the radonvet. des in

.. me mixture is not kneer. the concentraten temit for the mirture is the limit specir.ed in r opendix B for the rad.onues.de in the mixtur. hawna m iowesi conceniraton imi. or

{s* . b. It the identity of each radenuclide in*the mixture is not known, but it is known Piat certain radeonuctedes speeded in .

in Appendur s for any radenuci.de which is not known to be absent from the mixture: ci% -r are not present in tn Revision H B - 10

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

4 Nuclear Regulatory Commission Pt. 20 [sQ 20.1-20.602], App. C Tate f Table si

c. Eternent tatomic number) and isotope Cod -

Col. 2.- Col.1.- Col. 2--

g gj Water As(pO/ Water gng (*0/m4 rn0 4G/m0 W 8 is k that Se so, t 125, i 126. I 129. I 131 (I 133. Tabee N onid. PD

210. Po 210. At 211. Ra 223. As 224. Ra 2N. Ac 227. As 228. Th 230, Pa

231. Th 232. Theat. Cm 244 Q 254. and Fm 256 are not presene d W e is known that Sr to, t 125. I 126. I 128 (l 131. I 133. Table li only). Pb

~9 x 10*

3 x 10*

w 210. Po 210. Ra 223 Ra 226. He 228. Pa 231. Theat. Cm 248, Q 254 and Fm 256 are not spesent

- 6 x10" 2 x 10* *

. If a is knoum that Sr 90, i 129 (i 125. I 126. I 131. Table si oned. Pb 210. Ra 226. Ra 2N.Cm 2de, and Cr 254 are not proeect 2 x 10*

6 x 10*'

it a la knoom tint (1129. TabH N only). Ra 226. and Re 228 are not present - 3 x 10**

9 Of a la knoum mist aiphagminers and Sr 90, t 129. PD 210. 4 227. Ra 228 1 x 10*'

Pa 230. Pu 241, and Sk 249 are not present .

3 x10" 1 x 10* "

If k is known that alphaeminers and PD 210. Ac 227. Re 228 and Pu 241 ase cot preosc8

' If k b knoom tot alphaemitters and M 227 are not preseret 3 x t0* " 1 x 10* " - - -

3 x 10* " 1 x 10"8

W k is knoum that Ac 227. Th 230. Pe 231. Pu 234. Pu 239. Pu 240. Pu 242 Pu 244.Cm 248. Q 249 and Q 251 are not presear 3 x10*" 1 x 10"*

-peregraphs tem the 1. 2. ore.

or 3 above, Wie waiwes specired belour may be used for uranium and ks daughte

a. For purposes of Table f. Col 1-1x10*"

micrograms per cubic eneter of air natural uranium. pO/mi gross alpha aclMey: or 5x10*" pc/tre natural urarium or 75 h.- b.

,or.F.or.h.urposes.er o< o,f Tabte m .,.iu,a = it. Col 1--a x10*"pO/mi groes alpha acervity. 2x10"8pCi/ent natural urarwm; or 3 m A. For purposes of this note, a radionuchde may be cons.dered as not present h a eniature if (a) the ratio of the concentuhon of that rudonuchde in the mixture (C.) to the concentration lirnit for that radionuckle specir.ed h Table 8 of 6 Appendiz conaldered as *11" (A8%)h does not present not esosed the mixture Me. (i.e.

does not exceed % Lt.C./MM.51/10) and (b) the sum of such ratios for as the radionuc6 des (C./MM.+ G,/MMu + 5 %).

L e

[

+

9.

e Rev..ision 11 B - 11 s.;4 .. . e er:

.- . . _ . . . _ ,-

ML20198G638

Text

1.0 INTRODUCTION

Reference:

Dear Mr. Papanic:

SUBJECT:

1.0 INTRODUCTION

SUMMARY

Subject:

Dear Sirs:

1.0 INTRODUCTION

SUMMARY

7.0 REFERENCES

7.0 REFERENCES

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