ML041270286

Off-Site Dose Calculation Manual, Revision 26

ML041270286
Person / Time
Site:	Seabrook
Issue date:	02/13/2004
From:	Robinson D Florida Power & Light Energy Seabrook
To:	Office of Nuclear Reactor Regulation
References
Download: ML041270286 (411)

Text

{{#Wiki_filter:.Ty AI PROGRAM MANUAL Offsite Dose Calculation Manual SORC Review: 04-006 Date: 2/11/04 Effective Date: A-13yOL INR*BiA~TIClN ONLY ODCM Manual Owner: Rev. 26 D. A. Robinson

ABSTRACT The Offsite Dose Calculation Manual (ODCM) contains details to implement the requirements of Technical Specifications 6.7.6g and 6.7.6h. The Offsite Dose Calculation Manual (ODCM) is divided into two parts: (1) the Radioactive Effluent Controls Program for both in-plant radiological effluent monitoring of liquids and gases, along with the Radiological Environmental Monitoring Program (REMP) (Part A); and (2) approved methods to determine effluent monitor setpoint values and estimates of doses and radionuclide concentrations occurring beyond the boundaries of Seabrook Station resulting from normal Station operation (Part B). The sampling and analysis requirements of the Radioactive Effluent Controls Program, specified in Part A, provide the inputs for the models of Part B in order to calculate offsite doses and radionuclide concentrations necessary to determine compliance with the dose and concentration requirements of the Station Technical Specification 6.7.6g. The REIVIP required by Technical Specification 6.7.6h, and as specified within this manual, provides the means to determine that measurable concentrations of radioactive materials released as a result of the operation of Seabrook Station are not significantly higher than expected. Page 1 of 1 ODCM Rev. 24

OFFSITE DOSE CALCULATION MANUAL (ODCM) TABLE OF CONTENTS CONTENT PAGE PART A: RADIOLOGICAL EFFLUENT CONTROL AND ENVIRONMENTAL MONITORING PROGRAMS

1.0 INTRODUCTION

A.l-1 2.0 RESPONSIBILITIES (PART A) A.2-1 3.0 DEFITIONS A.3-1 4.0 CONTROL AND APPLICABILITY. A.4-1 5.0 RADIOACTIVE EFFLUENT MONITORING INSTRUMENTATION A.5-1 5.1 Liquids A.5-1 5.2 Radioactive Gaseous Effluent Monitoring Instrumentation A.5-9 6.0 RADIOACTIVE LIQUID EFFLUENTS A.6-1 6.1 Concentration A.6-l 6.2 Dose A.6-10 6.3 Liquid Radwaste Treatment System A.6-12 7.0 RADIOACTIVE GASEOUS EFFLUENTS A.7-1 7.1 Dose Rate A.7-1 7.2 Dose - Noble Gases A.7-7 7.3 Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form A.7-9 7.4 Gaseous Radwaste Treatment System A.7-11 8.0 TOTAL DOSE A.8-1 9.0 RADIOLOGICAL ENVIRONMENTAL-MONITORING A.9-1 9.1 Monitoring Program A.9-1 9.2 Land Use Census A.9-1 Page 1 ODCM Rev. 26

I I CONTENT PAGE 9.3 Interlaboratory Comparison Program A.9-13 10.0 REPORTS A.10-1 10.1 Annual Radiological Environmental Operating Report A.10-1 10.2 Annual Radioactive Effluent Release Report A.10-2 PART B: RADIOLOGICAL CALCULATIONAL METHODS AND PARAMETERS

1.0 INTRODUCTION

B.1-1 1.1 Responsibilities for Part B B.1-1 1.2 Summary of Methods, Dose Factors, Limits, Constants, Variables and Definitions B.1-2 2.0 METHOD TO CALCULATE OFF-SITE LIQUID CONCENTRATIONS B.2-1 2.1 Method to Determine F1 ENG and C NO B.2-1 2.2 Method to Determine Radionuclide Concentration for Each Liquid Effluent Source B.2-2 2.2.1 Waste Test Tanks B.2-2 2.2.2 Turbine Building Sump B.2-3 2.2.3 Steam Generator Blowdown Flash Tank B.2-3 2.2.4 Primary Component Cooling Water (PCCW) System B.2-3 3.0 OFF-SITE DOSE CALCULATION METHODS B.3-1 3.1 Introductory Concepts B.3-2 3.2 Method to Calculate the Total Body Dose from Liquid Releases B.3-4 3.2.1 Method I B.3-4 3.2.2 Method II B.3-5 3.3 Method to Calculate Maximum Organ Dose from Liquid Releases B.3-6 3.3.1 Method I B.3-6 3.3.2 Method II B.3-7 3.4 Method to Calculate the Total Body Dose Rate from Noble Gases B.3-8 3.4.1 Method I B.3-8 3.4.2 Method II B.3-10 Page 2 ODCM Rev. 26

CONTENT PAGE. 3.5 Method to Calculate the Skin Dose Rate from Noble Gases B.3-11 3.5.1 Method I B.3-11 3.5.2 Method I B.3-14 3.6 Method to Calculate the Critical Organ Dose Rate from Iodines, Tritium and Particulates with T,Tn Greater than 8 Days. *B.3-15 3.6.1 Method I B.3-15 3.6.2 Method I B.3-18 3.7 Method to Calculate the Gamma Air Dose from Noble Gases B.3-19 3.7.1 Method I B.3-19 3.7.2 Method I  ; B.3-21 3.8 Method to Calculate the Beta Air Dose from Noble Gases B.3-22 3.8.1 Method I B.3-22 3.8.2 Method II B.3-24 3.9 Method to Calculate the Critical Organ Dose from N odines, Tritium and Particulates B.3-25 3.9.1 Method I B.3-25 3.9.2 Method 1I B.3-27 3.10 Method to Calculate Direct Dose from Plant Operation 13.3-28 3.10.1 Method B.3-28 3.11 Dose Projections B.3-29 3.1 1.1 Liquid Dose Projections B.3-29 3.11.2 Gaseous Dose Projections B.3-30 3.12 Method to Calculate Total Dose from Plant Operations ,B.3-34 3.12.1 Method B.3-34 4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM B.4-1 5.0 SETPOINT DETERMINATIONS -B.5-1 5.1 Liquid Effluent Instrumentation Setpoints B.5-1 5.1.1 Liquid Waste Test Tank Monitor (RM-6509) B.5-1 Page 3 ODCM Rev. 26

I I CONTENT PAGE 5.1.1.1 Method to Determine the Setpoint of the Liquid Waste Test Tank Monitor (RM-6509) B.5-1 5.1.1.2 Liquid Waste Test Tank Monitor Setpoint Example B.5-5 5.1.2 Turbine Building Drains Liquid Effluent Monitor (RM-6521) B.5-6 5.1.3 Steam Generator Blowdown Liquid Sample Monitor (RM-6519) B.5-7 5.1.4 PCCW Head Tank Rate-of-Change Alarm Setpoint B.5-8 5.1.5 PCCW Radiation Monitor B.5-9 5.2 Gaseous Effluent Instrumentation Setpoints B.5-10 5.2.1 Plant Vent Wide-Range Gas Monitors (RM-6528-1, 2 and 3) B.5-10 5.2.1.1 Method to Determine the Setpoint of the Plant Vent Wide Range Gas Monitors (RM-6528-1, 2 and 3) B.5-10 5.2.1.2 Plant Vent Wide Range Gas Monitor Setpoint Example for Limiting Case B.5-12 5.2.2 Waste Gas System Monitors (RM-6504 and RM-6503) B.5-15 5.2.3 Main Condenser Air Evacuation Monitor (RM-6505) B.5-15 6.0 LIQUID AND GASEOUS EFFLUENT STREAMS, RADIATION MONITORS AND RADWASTE TREATMENT SYSTEMS B.6-1 7.0 BASES FOR DOSE CALCULATION METHODS B.7-1 7.1 Liquid Release Dose Calculations B.7-1 7.1.1 Dose to the Total Body B.74 7.1.2 Dose to the Critical Organ B.7-4 7.2 Gaseous Release Dose Calculations B.7-7 7.2.1 Total Body Dose Rate from Noble Gases B.7-7 7.2.2 Skin Dose Rate from Noble Gases B.7-8 7.2.3 Critical Organ Dose Rate from lodines, Tritium and Particulates with Half-Lives Greater Than Eight Days B.7-11 7.2.4 Gamma Dose to Air from Noble Gases B.7-13 7.2.5 Beta Dose to Air from Noble Gases B.7-14 7.2.6 Dose to Critical Organ from Iodines, Tritium and Particulates with Half-Lives Greater Than Eight Days B.7-16 7.2.7 Special Receptor Gaseous Release Dose Calculations B.7-18 7.2.7.1 Total Body Dose Rate from Noble Gases B.7-18 7.2.7.2 Skin Dose Rate from Noble Gases B.7-20 Page 4 ODCM Rev. 26

CONTENT PAGE 7.2.7.3 Critical Organ Dose Rate from Iodines, Tritium and Particulates with Half-Lives Greater Than Eight Days B.7-21 7.2.7.4 Gamma Dose'to Air from Noble Gases B.7-22 7.2.7.5 Beta Dose to Air from Noble Gases B.7-23 7.2.7.6 Critical Organ Dose from lodines, Tritium and Particulates with Half-Lives Greater Than Eight Days B.7-25 7.3 Receptor Points and Average Atmospheric Dispersion Factors for Important Exposure Pathways B.7-30 7.3.1 Receptor Locations B.7-30 7.3.2 Seabrook Station Atmospheric Dispersion Model B.7-30 7.3.3 Average Atmospheric Dispersion Factors for Receptors B.7-31 8.0 BASES FOR LIQUID AND GASEOUS MONITOR SETPOINTS B.8-1 8.1 Basis for the Liquid Waste Test Tank Monitor Setpoint B.8.1 8.2 Basis for the Plant Vent Wide Ranige Gas Monitor Setpoints B.8-5 8.3 Basis for PCCW Head Tank Ratewf-Change'Alarm Setpoint B.8-10' 8.4 Basis for Waste Gas Processing System Monitors (RM-6504 and RM-6503) B.8-1l 8.5 Basis for the Main Condenser Air Evacuation Monitor Setpoint (RM-6505) B.8-13 8.5.1 Limiting Example for the Air Evacuation Monitor Setpoint During Normal Operations,, , B.8-13 8.5.2 Example for the Air Evacuation Monitor Setpoint During Start Up (Hoggig Mode) B.8-15 References R-I

                                         ,        .             1'. , -..

Appendix A: Dose Conversion Factors A-I Appendix B: Annual Average Effluent Concentration Limits Taken From 10 CFR 20, Appendix B B-I Appendix C: EMS Software Documentation, C-1 Page 5 1 ODCM Rev. 26

Il CONTENT PAGE LIST OF TABLES AND FIGURES PART A TABLES A.5.1-1 Radioactive Liquid Effluent Monitoring Instrumentation A.5-5 A.5.1-2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements A.5-7 A.5.2-1 Radioactive Gaseous Effluent Monitoring Instrumentation A.5-15 A.5.2-2 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements A.5-17 A.6.1-1 Radioactive Liquid Waste Sampling and Analysis Program A.6-3 A.7.1-1 Radioactive Gaseous Waste Sampling and Analysis Program A.7-3 A.9.1-1 Radiological Environmental Monitoring Program A.9-3 A.9.1-2 Detection Capabilities for Environmental Sample Analysis A.9-7 A.9.1-3 Reporting Levels for Radioactivity Concentrations in Environmental Samples A.9-10 PART B TABLES B.l-1 Summary of Radiological Effluent Part A Controls and Implementing Equations B.1-3 B.1-2 Summary of Method I Equations to Calculate Unrestricted Area Liquid Concentrations B.1-6 B.1-3 Summary of Method I Equations to Calculate Off-Site Doses from Liquid Releases B. 1-7 B.1-4 Summary of Method I Equations to Calculate Dose Rates B.1-8 B.1-5 Summary of Method I Equations to Calculate Doses to Air from Noble Gases B.1-11 B.1-6 Summary of Method I Equations to Calculate Dose to an Individual from Tritium, Iodine and Particulates B.1-13 B. 1-7 Summary of Methods for Setpoint Determinations B.1-14 B.1-8 Summary of Variables B.1-15 B.1-9 Definition of Terms B.1-22 Page 6 ODCM Rev. 26

CONTENT PAGE B.1-10 Dose Factors Specific for Seabrook Station for Noble Gas Releases I B.1-23 B.1-11 Dose Factors Specific for Seabrook Station for Liquid Releases B.1-24 B.1-12 Dose and Dose Rate Factors Specific for Seabrook Station for Iodines, Tritium and Particulate Releases B.1-25 B.1-13 Combined Skin Dose Rate Factors Specific for Seabrook Station B Special Receptors for Noble Gas Release B.1-26 B.1-14 Dose and Dose Rate Factors Specific for the Science and Nature Center for Iodine, Tritium, and Particulate Releases B.1-27 B.1-15 Dose and Dose Rate Factors Specific for the "Rocks" for Iodine, Tritium, and Particulate Releases B.1-28 B.4-1 Radiological Environmental Monitoring Stations B.4-2 B.7-1 Usage Factors for Various Liquid Pathways at Seabrook Station B.7-6 B.7-2 Environmental Parameters for Gaseous Effluents at Seabrook Station B.7-27 B.7-3 Usage Factors for Various Gaseous Pathways at Seabrook Station B.7-29 B.74 Seabrook Station Long-Term Average Dispersion Factors* Primary Vent Stack B.7-33 B.7-5 Seabrook Station Long-Term Average Dispersion Factors for Special (On-Site) Receptors Primary Vent Stack B.7-34 B.7-6 Seabrook Station Long-Term Atmospheric Diffusion and Deposition Factors Ground-Level Release Pathway B.7-35 PART B FIGURES B.4-1 Radiological Environmental Monitoring Locations within 4 Kilometers of Seabrook Station B.4-5 B.4-2 Radiological Environmental Monitoring Locations Between 4 Kilometers and 12 Kilometers from Seabrook Station B.4-6 B.4-3 Radiological Environmental Monitoring Locations Outside 12 Kilometers of Seabrook Station B.4-7 B.44 Direct Radiation Monitoring Locations within 4 Kilometers of Seabrook Station B.4-8 B.4-5 Direct Radiation Monitoring Locations Between 4 Kilometers and 12 Kilometers from Seabrook Station B.4-9 Page 7 ODCM Rev. 26

I I CONTENT PAGE B.4-6 Direct Radiation Monitoring Locations Outside 12 Kilometers of Seabrook Station B.4-10 B.6-1 Liquid Effluent Streams, Radiation Monitors, and Radwaste Treatment System at Seabrook Station B.6-2 B.6-2 Gaseous Effluent Streams, Radiation Monitors, and Radwaste Treatment System at Seabrook Station B.6-3 Page 8 ODCM Rev. 26

LIST OF EFFECTIVE PAGES PAGE REV. PAGE REV. PAGE REV. Cover 26 B.6-3 25 Abstract 24 B.7-1 thru B.7-35 24 TOC 1 -8 26 B.8-1 thru B.8-18 25 LOEP I 26 R-1 21 A.l-l 24 A-1 thurA-15 21 A.2-1 24 B-1 thru B-52 25 A.3-1 24 C-1 22 A.4-1 thru A.4-4 25 C-2 22 C-3 16 A.5-1 thru A.5-17 25 C-4 16 C-5 16 A.6-1 thru A.6-13 25 C-6 22 A.7-1 thru A.7-12 25 A.8-1 24 A.8-2 24 A.9-1 thru A.9-13 26 A.10-1 24 A.10-2 24 A.10-3 24 B.1-0 thru B.1-28 25 B.2-1 25 B.2-2 25 B.2-3 25 B.3-1 thru B.3-34 25 B.4-1 thruB.4-11 26 B.5-1 thru B.5-18 25 B.6-1 25 B.6-2 25 Page 1 ODCM Rev. 26

PART A RADIOLOGICAL EFFLUENT CONTROL AND ENVIRONMENTAL MONITORING PROGRAMS

1.0 INTRODUCTION

The Offsite Dose Calculation Manual (ODCM) contains details to implement the Radioactive Effluent Controls and Environmental Monitoring Programs" of Technical Specifications 6.7.6g and 6.7.6h. The purpose of this manual is to contain details for the implementation of the Radioactive Effluent Technical Requirement Program (RETRP) and the Radiological Environmental Monitoring Program (REMP). These programs are required by Technical Specifications 6.7.6g and 6.7.6h. Part A of this manual defines specific concentrations, sampling regimes and frequencies for both the RETRP and the REMP. These activities are the defined surveillances for radiological releases. Part A also defines specific sampling locations for the RETRP. The information contained in Part A is used as input into the models that are used in Part B. The Part B models identify the calculational methods for determining radiation monitor setpoints, offsite doses and effluent concentrations of radionuclides. Part B also defines sampling locations for the REMP. The data resulting from the surveillance and monitoring programs described in Part A provide a means to confirm that concentrations of radioactive material released, as a result of routine Seabrook Station operations, do not contribute to effluent dose significantly different than as postulated in Part B. A.1-1 ODCM Rev. 24

2.0 RESPONSIBILITIES (PART A) All changes to the ODCM shall be reviewed by the Station Operation Review Committee (SORC), approved by the Station Director, and documented per Administrative Control 6.13 of the Technical Specifications. The change process is controlled by the Applicability Determination Process as controlled by the 10 CFR 50.59 Resource Manual (5059RM). Changes made to Part A shall be submitted to the NRC for its information in the Annual Radioactive Effluent Release Report for the period in which the change(s) was made effective, pursuant to T.S. 6.13. It shall be the responsibility of the Station Director to ensure that the ODCM is used in the performance of the Radioactive Effluent Control and Environmental Monitoring Program implementation requirements, as identified under Administrative Controls 6.7.6g and 6.7.6h of the Technical Specifications. A.2-1 ODCM Rev. 24

13.0 DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout these Controls. Terms used in these Controls and not defined herein have the same definition as listed in the Technical Specifications. A.3-1 ODCM Rev. 24

4.0 CONTROL AND APPLICABILITY This section provides a summary listing of the Controls and Applicability requirements of the ODCM. The RECP conforms with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable. The: REMP provides for monitoring the radiation and radionuclides in the environs of the plant.' The specific implemeniation details for the RECP and REMP are located in the OFFSITE DOSE CALCULATION MANUAL (ODCM). Contained within the ODCM are the following CONTROLS: C.5.1 - RADIOACTIVE EFFLUENT MONITORING INSTRUMENTATION - LIQUIDS CONTROL - At All Times The radioactive liquid effluent monitoring instrumentation channels shown in Table A.5.1-1 shall be OPERABLE with their AlarmfTrip Setpoints set to ensure that the limits of Control C.6.1.1 are not, ' exceeded. The Alarm/Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM), Part B. C.5.2 - RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION CONTROL- As Shown on ODCM Table A:5.2-1: ' The radioactive gaseous effluent monitoring instrumentation channels shown in Table A.5.2-1 shall be OPERABLE with their Alarni/Trip Setpoints set to ensure that the limits of Control C.7.1.1 are not exceeded. The Alarm/Trip Setpoints of these channels meeting Control C.7.1.1 shall be determined and adjusted in accordance with the methodology and parameters in the ODCM (Part B). - C.6.1.1 - RADIOACTIVE LIQUID EFFLUENTS - CONCENTRATION CONTROL'- At All Times

                        ' o-     '   ,        . .A    ;

The concentration of radioactive material released in liquid effluents at the point of discharge from the multiport diffuser (see Technical Specifications Figure 5.1-3) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 X 104 microCurie/ml total activity. AA4 ODCM Rev. 25

II C.6.2.1 - DOSE CONTROL - At All Times The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (see Technical Specification Figure 5.1-3) shall be limited

• During any calendar quarter to less than or equal to 1.5 mrems to the whole body and to less than or equal to 5 mrems to any organ, and
• During any calendar year to less than or equal to 3 mrems to the whole body and to less than or equal to 10 mrems to any organ.

C.6.3.1 - LIQUID RADWASTE TREATMENT SYSTEM CONTROL - At All Times The Liquid Radwaste Treatment System shall be OPERABLE and appropriate portions of the system shall be used to reduce releases of radioactivity when the projected doses due to the liquid effluent to UNRESTRICTED AREAS (see Technical Specification Figure 5.1-3) would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 3 1-day period. C.7.1.1 - RADIOACTIVE GASEOUS EFFLUENTS - DOSE RATE CONTROL - At All Times The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Technical Specification Figure 5.1-1) shall be limited to the following:

• For noble gases: Less than or equal to 500 mrems/yr to the whole body and less than or equal to 3000 mrems/yr to the skin, and
• For Iodine-131, for Iodine-133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrems/yr to any organ.

A.4-2 ODCM Rev. 25

C.7.2.1 - DOSE - NOBLE GASES CONTROL - At All Times The air dose due to noble gases released in gaseous effluents to areas at and beyond the SITE BOUNDARY (see Technical Specification Figure 5. 1-1) shall be limited to the following:

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

   ' to 20 mrads for beta radiation.                 -:   - .                      .;

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

• During any calendar quarter: Less than or equal to 7.5 mrems to any organ, and
• During any calendar year Less than or equal to 15 mrems to any organ.

C.7.4.1 - GASEOUS RADWASTE TREATMENT SYSTEM CONTROL - At All Times The VENTILATION EXHAUST TREATMENT SYSTEM and the GASEOUS RADWASTE TREATMENT SYSTEM shall be OPERABLE and appropriate portions of these system shall be used to, reduce releases of radioactivity when the projected doses in 31 days due to gaseous effluent releases to areas at and beyond the SITE BOUNDARY (see Technical Specification Figure 5.1-1) would exceed

• 0.2 mrad to air from gamma radiation, or'-
• 0.4 mrad to air from beta radiation, or 1'; '- .;-r' i
• 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.

A.4-3 ODCM Rev. 25

I I C.8.1.1 - TOTAL DOSE CONTROL- At All Times The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. C.9.1.1 - RADIOLOGICAL ENVIRONMENTAL MONITORING - MONITORING PROGRAM CONTROL - At All Times The Radiological Environmental Monitoring Program (REMP) shall be conducted as specified in Table A.9.1-l. C.9.2.1 - LAND USE CENSUS CONTROL - At All Times A Land Use Census shall be conducted and shall identify within a distance of 8 kin (5 miles) the location in each of the 16 meteorological sectors of the nearest milk animal, the nearest residence, and the nearest garden** of greater than 50 m2 (500 ft2) producing broad leaf vegetation. C.9.3.1 - INTERLABORATORY COMPARISON PROGRAM CONTROL - At All Times In accordance with Technical Specification 6.7.6.h.3, analyses shall be performed on all radioactive materials supplied as part of an Interlaboratory Comparison Program, that has been approved by the Commission, that correspond to samples required by REMP.

• • Broad leaf vegetation sampling of at least three different kinds of vegetation may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted relative deposition values (D/Qs) in lieu of the garden census. Specifications for broad leaf vegetation sampling in the REMP shall be followed, including analysis of control samples.

A.4-4 ODCM Rev. 25

MONITORING INSTRUMENTATION 5.0 RADIOACTIVE EFFLUENT MONITORING INSTRUMENTATION 5.1 Liquids CONTROLS C.5.1 The radioactive liquid effluent monitoring instrumentation channels shown in Table A.5.1-1 shall be OPERABLE with their Alarm/Trip Setpoints set to ensure that the limits of Control C.6.1.1 are not exceeded. The Alarm/Trip Setpoints of these chaninels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM), Part B. APPLICABILITY: At all times. ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel Alarm/Trip" Setpoint less conservative than required by the above specification, immediately

               ; -suspend the release of radioactive liquid effluents monitored by'the affected channel,'

or declare the channel inoperable. -

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

Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain in the next Annual Radioactive Effluent Release Report pursuant to Technical Specification 6.8.1.4 and Part A, Section 10.2, of the ODCM,

                  - why this inoperability was not corrected in a timely manner.

SURVEILLANCE REQUIREMENTS S.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 OPERATIONAL TEST at the frequencies shown in Table A.5.1-2. BASES ' The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm/Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alanm/trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. A.5-1 ODCM Rev. 25

I I Table A.5.1-2 Item 3a of Control C.5.1 requires that a Channel Operational Test be performed on the radioactivity monitors (RM-R-6515 and RM-R-6516) for the PCCW System. This channel operational test is a digital channel operational test and requires that it shall demonstrate automatic isolation of the pathway and control room alarm annunciation. For Seabrook Station, these two radioactivity monitoring channels provide control room annunciation, but do not provide automatic isolation of the release pathway. This particular item was discussed in detail with the NRC staff reviewers. For this particular reason, the words "But Not Termination of Release" were added to Item 3 of Table A.5.1-2. The purpose of adding the above words to Item 3 was to preclude the addition of another Table Notation to Table A.5.1-2. Therefore, the channel operational test for these monitors only requires that they provide control room alarm annunciation. The CHANNEL CHECK for Flow Rate Measurement Devices (Table A.5.1-2, items 2.a. and 2.b. ) is required "at least once per 24 hours on days when continuous, periodic, or batch releases are made." Additionally, ACTION 31 of Table A.5.I-1 is only applicable during actual releases. Based on the above requirements, these instruments are only required to be OPERABLE during actual releases. Therefore, the CHANNEL CHECK is only required during periods when continuous, periodic, or batch releases are being made. The Primary Component Cooling Water (PCCW) System is monitored by radiation monitors, which are l required by Technical Specifications 3.3.3.1 and ODCM C.5.1 to be OPERABLE, or sampling of the PCCW and Service Water (SW) Systems is required. Clarification of this requirement needs to be made for certain PCCW System conditions. Below is a list of 3 conditions and their corresponding requirements.

1) If the PCCW System is shut down but not drained, grab samples shall be taken of PCCW and SW, as required in Technical Specification Table 3.3-6, Items 6a and 6b (Action 28).

2) During transition times when the PCCW system is in the process of being drained, grab samples, as required by Technical Specification Table 3.3-6 and ODCM C.5. 1,shall be taken until such time as sampling of PCCW is no longer possible. At this time neither PCCW nor SW need to be sampled. During transition times when the PCCW system is being filled, the taking of grab samples shall commence as soon as physically possible and continue in accordance with the requirements of Technical Specifications 3.3.3.1 and ODCM C.5.1 until PCCW is in service, the pumps are operating, and monitors are operable.

3) When PCCW is drained, there are no sampling requirements.

The above statements are consistent with the Technical Specification definition of OPERABILITY and with the Bases for Technical Specification 3.3.3.1. A.5-2 ODCM Rev. 25

The following actions are required when the Service Waterside of the Primary Component Cooling Water (PCCW) Heat Exchanger is drained and grab samples of the Service Water System are required:

a. Grab samples from the Service Water System will be obtained at the frequencies specified in Technical Specification 3.3.3.1 and ODCM C.5.1 as the Service Water System is being drained until obtaining these samplei is not physically possible.

b. Grab samples are not required once the Service Water System is drained such that it is not physically possible'to obtain the samples.

c. When refilling the Service Water System, grab samples shall resume as soon as physically possible,' at the intervals specified in the aforementioned sources, and continue until the PCCW radiation monitors (1-RM-6515 and 1-RM-6516) are OPERABLE.

Sampling of the PCCW siystem with the Service Water system drained and the PCCW system in operation shall continue per the requirements of Techfiical Specification 3.3.3.1 and this Control. The purpose of the plant radiation monitors is to sense radiation levels in selected plant systems and. locations and determine whether or not predetermined limits are being exceeded. In the case of the Primary Component Cooling Water (PCCW) loops, the-radiation monitors (I-RM-6515 and 1-RM-6516) sense radiation inthe PCCW system which could leak into the Service Water System and be discharged to the environment via the multiport diffuser. Per Control C.6.1.1, the concentration of radioactive material-released in liquid effluents at the point of discharge from the multiport diffuser must be within specified limits. This limitation provides assurance that the levels of radioactive materials in unrestricted areas will not pose a threat to the health and safety of the public. Based on the importance of maintaining radioactive effluent releases within limits that guarantee the health and safety of the public will not be at risk, the PCCW radiation monitors are required to be in operation at all times. When a radiation monitor is inoperable, grab samples from the PCCW and Service Water systems must be obtained and analyzed as a compensatory measure in accordance with Technical Specification 3.3.3.1, Table 3.3-6 Action 28 and this Control. If the service water system is drained, there is no potential for inadvertent radioactive liquid effluent release through the service water system to the environment via the multiport diffuser. Thus, when the system is drained there is no need to obtain the grab sample. However, when the system is being filled, grab samples must be obtained as soon as possible to ensure that the water discharged to the environment is in compliance with Control C.6.1.1. The purpose of the PCCW monitors is to detect radioactivity indicative of a leak from the Reactor Coolant System or from one of the other radioactive systems which exchange with the PCCW System. These monitors are required to be operable at all times. Grab samples of PCCW are required when the PCCW monitors are not operable. Since the purpose of obtaining the PCCW samples is to provide an indication of a leak of radioactive liquid into the PCCW systemn, draining of the Service Water system does not remove the reason for obtaining the PCCW grab samples. These samples shall be obtained as specified in Technical Specification 3.3.3.1 and this Control. This determination is consistent with the Bases for Technical Specification 3.3.3.1. The temporary lowering of an RDMS channel setpoint, byRDMS data base manipulation to verify alarm/trip functions, does not prevent the channel from continuously monitoring radiation levels (except WRGM). Additionally, when the setpoint is lowered below background radiation levels the associated trip functions will actuate equipment in their required operating mode as if a high radiation condition exists. The channel remains OPERABLE because monitoring and associated trip functions are not inhibited. Refer to TS-1 42 for further details. A.5-3 ODCM Rev. 25

II When the SGBD demineralizers are being rinsed to the ocean using SGBD water, the SGBD flash tank radiation monitor (RM-6519) may become inoperable in this alignment from decreased backpressure to run the monitor sample pump. If this happens, the sampling requirements of Table A.5.1-l ACTION 30 must be performed. RM-6509, although in the flowpath of the SGBD demineralizer rinse, cannot perform the function of RM-6519 because it cannot achieve the same sensitivity to radiation. However, RM-6509 shall have its setpoints established per plant procedures since the discharge flow path is through the SGBD demineralizers (where a potential to acquire radioactivity exists), but after RM-6519. If RM-6509 is inoperable, then in addition to the periodic sampling requirements of Table A.5.I-1 ACTION 30 for RM-6519, the batch sample and lineup verification of ACTION 29 would also have to be complied with, for RM-6509. It should be noted that, during a SGBD demineralizer rinse to the discharge transition structure with SB liquid, SB-FE-1918 is not in the flow path. It is acceptable to use a flow monitoring device in the final flow path (such as WL-FIT-1458) so that Table A.5.1-1 ACTION 31 does NOT have to be entered. The Note which corresponds to Table A.5.1-1 "**" states that pump performance curves generated in place "should" be used to estimate flow. Hence, there is no requirement to use the pump curves as described in these tables. A.5-4 ODCM Rev. 25

TABLE A.5.1-1 RADIOACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM ACTION CHANNELS INSTRUMENT - OPERABLE 1.Radioactivity Monitors Providing Alarmn and Automatic Termination of Release

a. Liquid Radwaste Test Tank Discharge- I 29

b. Steam Generator Blowdown Flash Tank Drain J1* -30

c. Turbine Building Sump Effluent Line .. I 30 V.

2. Flow Rate Measurement Devicesi  ;

                                                                                                               . I. II

a. Liquid Radwaste Test Tank Discharge .. I 31 I I .

b. Steam Generator Blo.vdown Flash Tank Drain 31

c. Circulating Water Discharge 1** N.A.

3. Radioactivity Monitors Providing Alarm but Not Termination of Release

a. Primary Component Cooling Water System (in lieu of I 32 service water monitors)

4. Rate of Change Monitor

a. Primary Component Cooling Water System Head Tank I 33 (in lieu of service water monitors)

 *Only applicable when steam generator blowdown is directed to the discharge transition structure without intermediate collection. The required radiation monitoring channel is RM-65 19. The flow path must include a flow indicator which can be used to provide total flow discharged during period of interest.
**Puip performance curves generated in place should be used to estimate flow.

A.5-5 ODCM Rev. 25

II TABLE A.5.1-1 (Continued) ACTION STATEMENTS ACTION 29 - With the number of channels OPERABLE less than the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that prior to initiating a release

a. At least two independent samples are analyzed in accordance with Surveillance S.6. 1.1, and

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

Otherwise, suspend release of radioactive effluents via this pathway. ACTION 30 - With the number of channels OPERABLE less than the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are analyzed for radioactivity at a lower limit of detection of no more than 10 7microCurie/ml

a. At least once per 12 hours when the specific activity of the secondary coolant is greater than 0.01 microCurie/gram DOSE EQUIVALENT I-13 1, or

b. At least once per 24 hours when the specific activity of the secondary coolant is less than or equal to 0.01 microCurie/gram DOSE EQUIVALENT 1-131.

ACTION 31 - With the number of channels OPERABLE less than the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours during actual releases. Pump performance curves generated in place may be used to estimate flow. ACTION 32 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, collect grab samples daily from the Primary Component Cooling Water System and the Service Water System and analyze the radioactivity until the inoperable channel(s) is restored to OPERABLE status. ACTION 33 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the radioactivity level is determined at least once per 12 hours during actual releases. A.5-6 ODCM Rev. 25

TABLE A.5.1-2 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL sOPERATIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST

1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release

a. Liquid Radwaste Test Tank Discharge D P - R(2) ' P(1)

b. Steam Generator Blowdown Flash Tank Drain D M R(2) Q(l)

c. Turbine Building Sumps Effluent Line D R(2) Q(l)

2. Flow Rate Measurement Devices

a. Liquid Radwaste Test Tank Discharge* D(3) N.A. R N.A.

b. Steam Generator Blowdown Flash Tank Drain*** D(3) N.A. R N.A.

c. Circulating Water Discharge ** N.A. N.A. N.A.

3. Radioactivity Monitor Providing Alarm but Not Termination of Release

a. Primary Component Cooling Water System (in lieu D M R(2) Q(1) of service water monitors)

4. Rate of Change Monitor

a. Primary Component Cooling Water System (in lieu D(4) N.A. R N.A of service water monitors)

Isolation of the flow path is accomplished by the Waste Test Tank Discharge Pump Trip Circuitry.

• • Pump curves may be used to estimate flow. .
    • Applies to the flow indicator used in the discharge path when steam generator blowdown is directed to the discharge transition structure without intermediate collection.

A.5-7 ODCM Rev. 25

                                                                      - -I                           I TABLE A.5.1-2 (Continued)

TABLE NOTATIONS (1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and Control Room alarm annunciation occurs if the instrument indicates measured levels above the normal or Surveillance test Alarm/Trip Setpoint. (2) The initial channel calibration for radioactivity measurement instrumentation shall include the use of a known (traceable to National Institute for Standards and Technology) liquid radioactive source positioned in a reproducible geometry with respect to the sensor. These standards shall permit calibrating the system over its normal operating range of energy and rate. For subsequent channel calibrations, sources that have been related to the initial calibration shall be used. (3) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours on days on which continuous, periodic, or batch releases are made. (4) CHANNEL CHECK shall consist of verifying indication of tank level during periods of release. CHANNEL CHECK shall be made at least once per 24 hours. A.5-8 ODCM Rev. 25

5.2 Radioactive Gaseous Effluent Monitoring Instrumentation

                                                    - ~     ~ le.  . .:

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

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

b. With the number of OPERABLE radioactive gaseous effluent monitoring instrumentation channels less than the Minimum Channels OPERABLE, take the ACTION shown in Table A.5.2-1. Restore the inoperable instrumentation to.

OPERABLE status within 30 days or, if unsuccessful, explain in the next Annual Radioactive Effluent Release Report pursuant to Technical Specification 6.8.1.4 and Part A, Section 10.2, of the ODCM, why this inoperability was not corrected in a timely manner. SURVEILLANCE REQUIREMENTS  ;- S.5.2 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE byperformance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL OPERATIONAL TEST at the frequencies shown in Table A.5.2-2. I BASES I . . .. I. I The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The Alarm/Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters'in the ODCM (Part B) to iensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumejntation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. .The sensitivity of any noble gas activity monitors used to show compliance with the gaseous effluent release requirements of Control C.7.2.1 shall be such that c6ncentrations as low as 1 X I04 pCi/cc are measurable. A-.5-9. ODCM Rev. 25

I I The main condenser air evacuation radiation monitor, RM-6505, is included with the Turbine Gland Seal Condenser Exhaust in Tables A.5.2-1 and A.5.2-2. Table A.5.2-1 defines the minimum channels operable and the required actions for the radioactive gaseous effluent monitoring instrumentation. Table A.5.2-2 lists the surveillance requirements for this instrumentation. It is recommended that the out-of-service time for RM-6505 be tracked for reporting per C.5.2. (Reference ACR 96-197). The Plant Vent Wide Range Gas Monitor (WRGM) design includes three ranges of noble gas monitors and two ranges of iodine and particulate sampling filters. The noble gas monitor, the equipment necessary to provide flow through three ranges of the noble gas monitors, and the iodine and particulate sample filters all affect the operability of the WRGM. The various combinations of out-of-service components are addressed in this clarification. The WRGM noble gas activity monitor has three overlapping detector ranges: low, mid, and high. UFSAR Table 12.3-15 lists the following ranges for the WRGM: Low Range I 7 - 10 pCi/cc Mid Range 10i3 - I0 3 High Range 101- IO The minimum number of operable channels for the noble gas activity monitor, the flow rate monitors and the iodine sampler and particulate sampler is one, respectively. The Controls do not list the specific WRGM noble gas activity monitor, the iodine/particulate sampler or the flow rate monitor channels separately by an instrumentation identification tag number. Heat tracing of the sample lines, from the plant vent to the WRGM, is not listed as a specific requirement for WRGM operability. However, these circuits are necessary to ensure that the particulate and iodine concentration of the sample reaching the WRGM is representative of the effluent. The purpose of heat tracing is to ensure that the sample lines are free of moisture due to condensation. The low temperature alarm setpoint is variable based on outside ambient air temperature, and ensures that the sample line tubing metal temperature is high enough to prevent the moisture in the air from condensing inside of the sample line. The ability to detect of noble gases is not affected by the operational status of the heat tracing circuits. The heat tracing on the sample lines within the PAB (CP 433, circuit 55) is not required for WRGM operation. (Engineering Evaluation, SS-EV-960017) The following equipment normally defines an operational WRGM: During routine releases,

                  -Sample flow through one of the particulate and iodine (P&I) filters F-156-1,2,3 and channel 1 (low range) noble gas (NG) detectors using pump P-240-2, and
                  -Sample flow through P&I filters F-156-7,8 using pump RM-P-391.

or in the event the noble gas activity is in the mid/high range,

                  -Sample flow through one of the particulate/iodine (P&D filters F-156-4,5,6 and channels 2 or 3 (mid/high range) NG detectors using pump P-240-1, and
                 -Sample flow bypassing P&I filters F-156-7,8 using pump RM-P-391.

A.5-10 ODCM Rev. 25

At all times,

                -Heat tracing (HT) on the sample lines from the plant vent to the WRGM.

Note: Dewpoint measurements maybe used if heat tracing is out of service. (See the following table)

                -Vent stack flow rate monitor.
                -WRGM sample flow rate for the channel(s) in service.

The table below lists the action required in the event that a WRGM component is out of service. Out of Service Component Action Low range NG detector Enter Action 33. Perform grab sampling as required.' High range NG detector Enter A 'ction33.' The actions required by Action 33 are satisfied provided the Low range NG detector provides continuous indication of the'effluent concentrations, grab sampling not required. In the

                                        -unlikelyevent that elevated effluent concentrations above the capability of the low range detector are present, then grab sampling or backup monitoring will/may be required.

Mid range NG detector No action required, detection capability met by the overlapping ranges of the low and high NG'detectors. (May need to ensure that the high range pump'[RM-P-240-] starts on increasing activity.) RM-P-391 Enter Action 35. The mid and high range particulate and iodine sampling capability is lost. If a low range P&I filter F-156-1,2 or3'. is in service then no further action is required. If the low range P&I filtersare out of service then complywith Action 35 within one hour. - ' P-240-l (High range pump) -  :; EnterAction 33 and 35. (Action 33 is satisfied provided the low - range NG detector provides continuous indication of the effluent;

                                       *concentrations, grab sampling is not required. Action 35 is satisfied "if P-240-2 and filters 1F-156-1,-2, or -3 are in service. If these P&I
                                       'filters are out of service and the NG activity is in the low range, then ensure compliance with Action 35 within one hour of identifying the out of service'conditionl 'In the Unlikely event that elevated effluent concentrations above the capability of the low range detector are present,' then, with P-391 operating, install a portable sample pump
                                      ' across valves V28 and V29 to facilitate P&I grab sampling using filters F-156-4,-5, or -6, and noble gas sampling using the medium and highorange detectors.

P-240-2 (Low range pump) Enter Action 33 and 35. Action' 33 is satisfied by performing grab

                                       'samples. -Action 35 is satisfied by ensuring the operation of P-391
                                        .withfilters F-156-7 & 8 in service within one hour of identifying the out-of-service condition.'

A.5-11 ODCM Rev. 25

I I HT circuit: Enter Action 36. Action 36 is satisfied and the WRGM may CP-434 Ckt 28. remain OPERABLE with CP434 Ckt 28 out of (Sample line temperature service provided that CP-426 Ckt 46 is less than setpoint.) energized within I hr of the out-of-service condition. Flow rate monitor and/or Comply with Action 32. sampler flow rate monitor. Action Statement 35 provides no guidance with regard to time required to initiate auxiliary sampling upon failure of a monitor. A finite time is required to take the appropriate actions to initiate auxiliary sampling. An interval of 60 minutes is a reasonable period of time in which to accomplish these actions provided that no activity occurs during this period which could result in an increase in radiation release levels. Since the intent of Action 35 is to allow continued release of gaseous effluents provided an alternate means of continuous monitoring/collection capability is on-going during the release of radioactive gaseous effluents, the 60 minute time frame for auxiliary sampling to be established is still a reasonable period of time to complete the necessary manual actions to establish auxiliary sampling. If auxiliary sampling cannot be established within 60 minutes then the initial action of immediately suspending the release of radioactive gaseous effluents should be done, as specified in Action a. of C.5.2. It should be noted that for lack of specified criteria the 60-minute time period is solely based on prudent engineering judgment for completion of manual actions in order to satisfy the intent of Action 35. Operation beyond 60 minutes without auxiliary sampling service would need to be justified by engineering calculation to ensure continued compliance with 10 CFR Part 20 limits. On those occasions when a radiation monitor or any system/component must be rendered inoperable to perform a surveillance test, the Station Management Manual (SSMM) policy regarding "the use of ACTION requirements to perform maintenance or a test" applies. When a surveillance test must be performed on the WRGM, rendering it inoperable, Action 35 cannot be fully satisfied because of the nature of testing is incompatible with the Action 35 required installation of auxiliary sampling equipment. However, because the performance of the WRGM surveillance renders it inoperable for only a short period of time (e.g., less than one hour), it is reasonable to allow the surveillance test to be performed without the installation of the auxiliary sampling equipment. It should be noted that neither C.5.2 Action a. nor Action b. requires the immediate establishment of auxiliary sampling. However, if there is concern that the results of surveillance testing activities will identify the instrumentation as inoperable then it would be prudent to set up the auxiliary sampling equipment prior to surveillance testing. The prudent action would prevent the potential situation of continued release of gaseous effluents beyond 60 minutes without continuous monitoring/collection capability. The current procedural method of collecting the grab sample from the plant vent release pathway requires the shutdown of the compensatory sampling equipment pump (for pressure equilibrium purposes) whenever a grab sample is to be withdrawn into the sample bottle. Shutting down the pump raises the question as to whether this action contradicts the "continuous collection" requirement of Action 35. Action 35 allow effluent release to continue provided samples are continuously collected (as required in Table A.7. 1-1) with auxiliary equipment whenever the number of channels OPERABLE is less than the Minimum Channels OPERABLE requirement. Table A.7.1-1 requires that the sampling frequency be continuous for iodine and particulate and a monthly grab sample for noble gasses (Kr and Xe). The ODCM also requires that the ratio of the sample flow rate to the sampled stream flow rate be known/determined for the time period covered by each dose or dose rate calculation made in accordance with C.7. 1.1, C.7.2. 1, and C.7.3.1 (i.e., weekly and/or monthly). A.5-12 ODCM Rev. 25

It must be noted that Action 35 pertains to the iodine and particulate samplers. For noble gas collection, Action 33 is applicable which requires grab samples be taken once per 12 hours and analyzed for radioactivity within 24 hours. Action 33 does not specify that auxiliary sampling for noble gas must be continuous; therefore, the concern for "continuous" monitoring/collection is not applicable for auxiliary sampling of noble gas. Whenever the station is operating under the auspices of Action 35 the process of collecting grab samples by the auxiliary sampling method necessitates, on occasions, the temporary disablement of permanent and/or temporary equipment (e.g., installation, and disconnection of auxiliary sampling equipmient, pressure equalization, etc.) in order to achieve and comply with the requirements of Action 35. Therefore, actions required (e.g. temporarily shutting down the sample pump in order to install / remove / equalize sample bottles, thus interrupting continuous flow) to obtain a grab sample are not considered actions that are contrary in meeting the intent of Action 35. The temporary lowering of an RDMS channel setpoint, by RDMS data base manipulation to verify alami/trip functions, does not prevent the channel from continuously monitoring radiation levels (except WRGM). Additionally, when the setpoint is lowered below background radiation levels the associated trip functions will actuate equipment in their required operating mode as if a high radiation condition exists. The channel remains OPERABLE because monitoring and associated trip functions are not inhibited. Refer to TS-142 for further details. Therefore, during performance of a RDMS channel DCOT, the LCO remains satisfied. Entering an ACTiON statement is not appropriate nor required (except for WRGM DCOT). However, because the channel is in alarm status, increased operator vigilance is required to note any increase in radiation levels during the DCOT surveillance period and to take remedial actions if required. C.5.2 ACTION Statement #33 is applied if RM-6504 is inoperable. The intent of the last sentence is that RM-6503 may be used instead of taking a grab sample. It is not intended that RM-6503 be used in place of RM-6504 and ACTION Statement #33 not entered. RM-6504 monitors the radiation level of the gas stream at the outlet of the waste gas compressors. If a high radiation level is detected, RM-6504 automatically closes WG-FV-1 602. The closing of WG-FV-1602 isolates a potential radiological release path to the environment. RM-6503, located at the inlet to the waste gas compressor, provides alarm and monitoring functions only. It does not have the ability to terminate a radiological release. -Therefore, it cannot be used as a substitute for RM-6504. C.5.2 ACTION b. permits operations to continue for up to 30 days with an inoperable instrument channel. If the inoperable instrument is not returned to OPERABLE status within this time, a report must be submitted explaining why the inoperability was not corrected in a timely manner. If RM-6503 were considered a alternate for RM-6504 then operations could continue indefinitely without the ability to automatically terminate a radiological release. This is clearly not the intent of C.5.2 ACTION Statement #33. Table A.5.2-1, Radioactive Gaseous Effluent Monitoring Instrumentation, specifically lists RM-6504 as the instrument required to satisfythe Limiting Condition for operation. This table also states that the monitor provide the functions of alarm and automatic termination of release.

f -- -

A.5-13 :3ODCM Rev. 25

TABLE A.5.2-1 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION

1. (Not Used)

2. PLANT VENT-WIDE RANGE GAS MONITOR

a. Noble Gas Activity Monitor I

• 33

b. Iodine Sampler I

• 35

c. Particulate Sampler I

• 35

d. Flow Rate Monitor I

• 32

e. Sampler Flow Rate Monitor I

• 32, 35

f. Sample Line Temperature I

• 36

3. GASEOUS WASTE PROCESSING SYSTEM (Providing Alarm and Automatic Termination of Release - RM-6504)

a. Noble Gas Activity Monitor (Process) I

• 33

4. TURBINE GLAND SEAL CONDENSER EXHAUST

a. Iodine Sampler I 35

b. Particulate Sampler 1 35

c. Sampler Flow Rate Indicator 32, 35

d. Noble Gas Activity Monitor (RM 6505) I 34 At all times.

(Not Used.) When the gland seal exhauster is in operation. A.5-14 ODCM Rev. 25

TABLE A.5.2-1 (Continued) ACTION STATEMENTS ACTION 32 - With the number of channels OPERABLE less than the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours. ACTION 33 - 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 12 hours and these samples are analyzed for radioactivity within 24 hours. For RM-6504, RM-6503 may be used instead of taking grab samples (see Bases for reporting requirements). ACTION 34 - With RM-6505 INOPERABLE and the gland seal exhauster in operation, effluent releases via the turbine gland seal condenser exhaust may continue provided grab samples from condenser air evacuation pump effluent are taken at least once per

              '12 hours, and analyzed for'radioactivity within 24 hours.

ACTION 35 - With the number of channels OPERABLE less than the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may coitinue provided samples are continuously collected with auxiliary sampling equipment as required in this document. Auxiliary sampling must be initiated within 60 minutes. Additionally, the'auxiliary sampling equipment need not be installed during surveillance activities provided the surveillance testing is completed in less than one hour. Actions required (e.g., temporarily shutting down the sample pump in order to install / remove / equalize sample bottles, thus interrupting continuous flow) to obtain a grab sample are not considered actions that are contrary in meeting the intent of this Action. Auxiliary sample equipment includes sample flow monitoring to provide information used in the sample analysis. ACTION 36 - If, for any reason, the sample line temperature cannot be maintained greater than 200 above outside ambient air temperature, the WRGM may remain OPERABLE provided dewpoint measurements are obtained every 12 hours verifying that conditions do not exist for condensation in the sample line with the inservice' operating samnple pump. (CXO901.38) A.5-15 ODCM Rev. 25

TABLE A.5.2-2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REOUIREMENTS CHANNEL SOURCE CHANNEL CHANNEL MODES FOR WHICH INSTRUMENT CHECK CHECK CALIBRATION OPERATIONAL SURVEILLANCE IS TEST REQUIRED I (Not Used)

2. PLANT VENT-WIDE RANGE GAS MONITOR

a. Noble Gas Activity Monitor D M R(3) Q(2) *

b. Iodine Sampler W N.A N.A. N.A. *

c. Particulate Sampler W N.A. N.A. N.A. *

d. Flow Rate Monitor D N.A. R Q**** *

e. Sampler Flow Rate Monitor D N.A. R Q**** *

f. Sample Line Temperature N.A. N.A. R N.A. *

3. GASEOUS WASTE PROCESSING SYSTEM (Providing Alarm and Automatic Termination of Release)

a. Noble Gas Activity Monitor D N.A. R(5) Q(M) *

(Process)

4. TURBINE GLAND SEAL CONDENSER EXHAUST

a. Iodine Sampler W N.A. N.A. N.A.

b. Particulate Sampler W N.A. N.A. N.A.

c. Sampler Flow Rate Indicator D N.A. N.A. N.A.

d. Noble Gas Activity Monitor (RM 6505) D M R(3) Q(2)

A.5-16 ODCM Rev. 25

TABLE A.5.2-2 (Continued) TABLE NOTATIONS

• At all times.

**   (Not Used.)
*** When the gland seal exhauster is in operation.

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

(1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and Control Room alarm annunciation occurs if the instrument indicates measured levels above the normal or Surveillance test AlarmnTrip Setpoint. (2) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that Control Room alarm annunciation occurs if the instrument indicates measured levels above the normal or Surveillance test Alarm Setpoint. (3) The initial channel calibration for radioactivity measurement instrumentation shall include the use of a known (traceable to National Institute for Standards and Technology) radioactive source positioned in a reproducible geometry with respect to the sensor. These standards should permit calibrating the system over its normal operating range of rate capabilities. For subsequent channel calibrations, sources that have been related to the initial calibration shall be used. (4) (Not Used). (5) The CHANNEL CALIBRATION shall be performed using sources of various activities covering the measurement range of the monitor to verify that the response is linear. Sources shall be used to verify the monitor response only for the intended energy range. A.5-1 7 ODCM Rev. 25

6.0 RADIOACTIVE LIQUID EFFLUENTS 6.1 Concentration ' ' ' I-CONTROLS ' C.6.1.1 The concentration of radioactive material released in liquid effluents at the point of discharge from the multipbrt diffuser (see Technical Specifications Figure 5.1-3) shall be,. limited to not more than ten times the concentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2, for radionuclides other than dissolved or entrained noble gases. 'For dissolved or entrained 'noble gases, the concentration shall be limited to. 2 X 10-4 pCihid total activity. : APPLICABILITY: At all times.

                                                                                                  ..  .  . t. -;

ACTION: With the concentration of radioactive material released in liquid effluents at the point of discharge from the multipart diffuser exceeding the above limits, restore the concentration to within the above limits within 15 minutes. SURVEILLANCE REQUIREMENTS

                                        ..         .mln   . . dfdralzdcod
                                                                .d.                    to the S.6.1.1      .Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program specified in Table A.6.1-1.

S.6.1.2 The results of the radioactivity analyses shall be used in accordance with the methodology and parameters in Part B of the ODCM to assure that the concentrations at the point of release are maintained within the limits of Control C.6.1.1.' BASES This Control is provided to ensure that the concentration of radioactive materials released in liquid waste effluents at the point of discharge from the multiport diffuser will be less than the concentration levels specified in 10 CFR Part 20, Appendix B to 20, Table 2, Column 2 (most restrictive). This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within (1) the Section Kl.A design objectives of Appendix 1, 10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the limits of Appendix I, 10 CFR 20.1301 and 20.1302 to the population. Those values assure a continuous discharge at those concentrations (8760 hours per year). Pursuant to the requirements of 10 CFR 50.36a to maintain effluent concentrations as low as reasonably achievable (ALARA), Appendix I tolO CFR 50 specifies dose values that are a small percentage of the dose limits in 10 CFR 20.1301. Consistent with Appendix I tolO CFR 50, to allow operational flexibility, this specification in conjunction with the dose specification in Section C-6.2 permits an instantaneous concentration release rate up to a factor of ten

• times greater than specified in 10 CFR 20, Appendix B, Table 2, Column 2 while continuing to limit the total annual discharge to a small fraction of the allowable annual dose as specified in Appendix I.

ODCM Rev. 25

                                                                                                              -  l l The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2. For technical requirements associated with the release of liquid                ^

effluent, the method currently in use for controlling releases of dissolved and entrained noble gases is suitable for demonstrating conformance to the requirements of the "new" 10 CFR 20, Appendix B ECL concentration limits because the 2X10 4 gCi/ml criterion is based on the "old" MPC value for Xe-135 as the driving radionuclide. Controlling liquid effluent to within the MPC values based on an instantaneous release rate (i.e., no time averaging of effluent concentrations) is considered more conservative than the requirements of the new Part 20 which have limits stated as effluent concentrations averaged over a year. In other words, if discharged dissolved and entrained noble gas concentrations remain within the instantaneous concentration limit of 2X10O4 pCi/ml during the times that discharges actually take place, then there is reasonable confidence that the annual average limits established by the ECL values will also be met. This position is based on a June 30, 1993 letter from Thomas E. Murley (then Director, Office of Nuclear Reactor Regulation) to Thomas E. Tipton of NEI, in which the NRC responded to an industry inquiry on promulgation of a new Part 20. Controls C.6.1.1 and C.5.1 provide controls to ensure that the concentration of radioactive materials released in liquid waste effluents at the point of discharge from the multiport diffuser will be less than the concentration levels specified in 10 CER 20, Appendix B, Table 2, Column 2. As no LLD is specified for the compensatory samples taken for an inoperable PCCW Head Tank Rate of Change Monitor, the LLD for these samples must ensure that these limits are met. Although the periodic Service Water System sample is counted to an LLD of 5x10epCi/cc, the compensatory samples for inoperable SGBD Flash Tank and Turbine Building Sump Monitors are required to be counted to an LLD of lxlOepCi/cc. This more restrictive limit will ensure that the limits of 10 CFR 20 are met during periods of PCCW Head Tank Rate of Change Monitor inoperability, thereby ensuring compliance with the requirements of the respective Controls. Counting the required grab samples to an LLD of lxI0-7pCi/cc is therefore an acceptable method of complying with these requirements; it is not necessary to meet the LLD of Ixl0I OCi/cc specified as the equivalent sensitivity of the PCCW Head Tank Rate of Change Monitor. A.6-2 ODCM Rev. 25

TABLE A.6.1-1 RADIOACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit of Detection Minimum Analysis (LLD)(l. Liquid Release Type Sampling Frequency Frequency Type of Activity Analysis (PUCi/ml) A. Liquid Radwaste Test P P Principal Gamma 5x 10O7 Tanks Each Batch Each Batch Emitters() - 1-131 lxiO-6 (Batch Release))  ; P M Dissolved and Entrained lxlO-4 One Batch/M Gases (Gamma Emitters) P 4 H-3 Ix10'l-Each Batch Composite . .. Gross Alpha x10

                                   .                  Q(4)      Sr-89, Sr         -        5xif0 Each Batch          Composite Fe-55                           lxlO'     -

B. Turbine Building - W - -W Principal Gamma - 5x107 Sump Effluente'" Grab Sample Emitters(3) I-131 lxI O (Continuous W - M- Dissolved and Entrained lx1 5-Release)(5 Grab Sample Gases (Gamma Emitters) A.6-3 ODCM Rev. 25

TABLE A.6.1-1 RADIOACTIVE LIOUTD WASTE SAMPLING AND ANALYSIS PROGRAM (Continued) Lower Limit of Detection Minimum Analysis (LLD) (l) Liquid Release Type Sampling Frequency Frequency Type of Activity Analysis (11Ci/ml) B. (Continued) W M H-3 x10 5 Grab Sample Gross Alpha 1x10 7 W Q (9) Sr-89, Sr-90 5x10 ' Grab Sample Fe-55 IXI0.6 C. Steam Generator W W Principal Gamma 5x10 7 Blowdown 6 8 Flash Grab Sample Emitters(3 ) Tank( x ) 1-131 IxlIO (Continuous Release)(5) W M Dissolved and Entrained 1x10*5 Grab Sample Gases (Gamma Emitters) W M H-3 x1i O-Grab Sample Gross Alpha 1x1O7 W Q(9) Sr-89, Sr-90 5x10 8 Grab Sample Fe-55 lxI 06 A.6-4 ODCM Rev. 25 0 0

TABLE A.6.1-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM (Continued) Lower Limit Minimum of Detection Analysis Type of Activity (LLD) (1) Liquid Release Type Sampling Frequency Frequency Analysis (iLCi/ml) D. Service Water(7)( 0 )w W Principal Gamma 5x10 7 I Grab Sample Emitters(3 ) I-131 lxi0.6 W M Dissolved and Entrained - lxO 5 Grab Sample Gases (Gamma Emitters) W M H-3 lxiO0-Grab Sample Gross Alpha 1x10 7 W Q Sr-89, Sr-90 5x103 Grab Sample Fe-55 lxi06

                                                 . .. ,. I I A.6-5                                     ODCM Rev. 25

TABLE A.6.I-1 RADIOACTIVE LIOULD WASTE SAMPLING AND ANALYSIS PROGRAM (Continued) Lower Limit Minimum of Detection Analysis Type of Activity (LLD) (') Liquid Release Type Sampling Frequency Frequency Analysis (~LCi/ml) E. Subsurface M M Principal Gamma 5xl0' Dewatering Grab Sample Emitters (3) H-3 2x10 6 M M Gross Alpha 1x10 7 Grab Sample M Q Sr-89, Sr-90 5x10 8 Grab Sample Fe-55 lxI0.6 P - Prior to Discharge W - Weekly M - Monthly Q - Quarterly A.6-6 ODCM Rev. 25

TABLE A.6.1-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM (Continued) Notations (l) The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95 percent probability with only 5 percent probability of falsely concluding that a blank observation represents a "real" signal.' For a particular measurement system, which may include radiochemical separation: LLD =' 4.66 Sb

                                         -ExVx2.22xlo
                                             -            6 xYxexp(-AAt)

Where: - - LLD = the "a priori" lower limit of detection (microcurie per unit mass or,volume), 4.66 = a constant derived from the K.1pba and Kb,, values for the 95% confidence level;  : Sb = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute), EB' =\ the counting efficiency (counts per disintegration), V = the sample size (units of mass or volume), 2.22 x10 6 = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable, X = the radioactive decay constant for the particular radionuclide (s7!), and At = the elapsed time between the midpoint of sample collection and the time of counting(s) . >:¢ .- '-- Typical values of E, V, Y, and A t should be used in the calculation. It should be recognized that the LLD is defined as an priori (before-the fact) limit" representing the capability of a measurement systeman'd not asan a iosteriori (after the fact) limit for a particular measurement. - (2) A batch release is the discharge of liquid wastes of a discrete volume: Prior to sampling for

    - analyses, each batch shall be isolated, and then thoroughly mixed to assure representative sampling.

A.6-7? ODCM Rev. 25

I I TABLE A.6.1-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM (Continued) Notations (Continued) (3) The principal gamma emitters for which the LLD specification applies include the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radioactive Effluent Release Report in accordance with Technical Specification 6.8.1.4. Isotopes which are not detected should be reported as "not detected." Values determined to be below detectable levels are not used in dose calculations. (4) A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released. (5) A continuous release is the discharge of liquid wastes of a nondiscrete volume, e.g., from a volume of a system that has an input flow during the continuous release. (6) Sampling and analysis is only required when Steam Generator Blowdown is directed to the discharge transition structure. (7) Principal gamma emitters shall be analyzed weekly in Service Water. Sample and analysis requirements for dissolved and entrained gases, tritium, gross alpha, strontium 89 and 90, and Iron 55 shall only be required when analysis for principal gamma emitters exceeds the LLD. The following are additional sampling and analysis requirements: a PCCW sampled and analyzed weekly for principal gamma emitters.

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

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

d. With a confirmed PCCW/SWS leak and PCCW activity in excess of lx 4 [tC/cc, sample SWS every 12 hours for principal gamma emitters.

e. The setpoint on the PCCW head tank liquid rate-of-change alarm will be set to ensure that its sensitivity to detect a PCCW/SWS leak is equal to or greater than that of an SWS radiation monitor, located in the unit's combined SWS discharge, with an LLD of IxlO -10C/cc. If this sensitivity cannot be achieved, the SWS will be sampled once every 12 hours.

A.6-8 ODCM Rev. 25

TABLE A.6.1-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM (Continued) Notations (Continued) (8) If the Turbine Building Sump (Steam Generat6r Blowdown Flash Tank)'isolate'due to high concentration of radioactivity, that liquid stieam will be sampled and analyzed for Iodine-1 31 and principal gamma emitters prior to release. (9) Quarterly composite analysis requirements "shall only be required when analysis for principal gamma emitters indicate positive radioactivity.

0) A grab sample can be considered as a combination of aliquots taken from each SW train during the same collection cycle or as individual samples taken from each train in service.,

o Principal gamma emitters and tritium shall be analyzed monthly in subsurface dewatering samples. Sample analysis requirements for gross alpha, strontium 89 and 90, and Iron 55 shall only be required when analysis for principal gamma emitters exceeds the LLD. I . A.6-9 ODCM Rev. 25

_ lIl 6.2 Dose CONTROLS C.6.2. 1 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (see Technical Specification Figure 5.1-3) shall be limited

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

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

APPLICABILITY: At all times. ACTION: With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.8.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits. SURVEILLANCE REOUIREMENTS S.6.2.1 Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in Part B of the ODCM at least once per 31 days. A.6-10 ODCM Rev. 25

BASES This Control is provided to implement the requirements of Sections H.A, lf.A, and IV.A of Appendix I to 10 ,CFR Part 50. The Control implements ,the guides set forth in Section IHA of Appendix L The ACTION statements provide the required oPerating flexibility and at the same time impleiment the guides set forth in Section 1V.A of Appendix I to assure that the releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept as low as is reasonably achievable. The dose calculation methodology and parameters in the ODCM implement the requirements in Section fIE.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of implementing Appendix I," April 1977.

                                                                                     . .                 1 12 A.6-1 1                                    ODCM Rev. 25

I I 6.3 Liquid Radwaste Treatment System CONTROLS C.6.3.1 The Liquid Radwaste Treatment System shall be OPERABLE and appropriate portions of the system shall be used to reduce releases of radioactivity when the projected doses due to the liquid effluent to UNRESTRICTED AREAS (see Technical Specification Figure 5.1-3) would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 3 1-day period. APPLICABILITY: At all times. ACTION: With radioactive liquid waste being discharged without treatment and in excess of the above limits and any portion of the Liquid Radwaste Treatment System which could reduce the radioactive liquid waste discharged not in operation, prepare and submit to the Commission within 30 days, pursuant to Specification 6.8.2, a Special Report that includes the following information:

a. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability,

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

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

SURVEILLANCE REQUIREMENTS S.6.3.1 Doses due to liquid releases from each unit to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the methodology and parameters in Part B of the ODCM when Liquid Radwaste Treatment Systems are not being fully utilized. S.6.3.2 The installed Liquid Radwaste Treatment System shall be considered OPERABLE by meeting Controls C.6.1.1 and C.6.2.1. A.6-12 ODCM Rev. 25

BASES The OPERABILITY of the Liquid Radwaste Treatment System ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept as low as is reasonably achievable. This specification implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objective given in Section ll.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the Liquid Radwaste Treatment System were specified as a suitable fraction of the dose design objectives set forth in Section Ml.A of Appendix A to 10 CFR Part 50 for liquid effluents. A.6-1 3 ODCM Rev. 25

7.0 RADIOACTIVE GASEOUS EFFLUENTS 7.1 Dose Rate CONTROLS

                                       ! _ ^ ; 'e'i i'-'al* re'lae in g as zd-C.7.1.1      The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (seeTechnical Specification Figure'5.1-1) shall be limited to the following:- -

a. For noble gases: Less than or equal to 500 mrems/yr to the whole body aiid less than or equal to 3000 mrem's/yr to the skin, and

b. For Iodine-13 1, for Iodine-133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less 'than or equal to 1500 mrems/yr to any organ. -

APPLICABILHTY: At all times. ' - ' - ACTION: With the dose rate(s) exceeding the above limits, decrease the release rate within 15 minutes to within the above limit(s). SURVEILLANCE REQUIREMENTS S.7.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in Part B of the ODCM. S.7.1.2 The dose rate due to Iodine-131, Iodine-133, tritium, and all 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 methodology and parameters in the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table A.7. 1-1. A.7 ODCM Rev. 25

I I BASES This Control is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 to UNRESTRICTED AREAS. The annual dose limits are the doses associated with the concentrations of I10 CFR Part 20, Appendix B, Table 2, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual average concentrations exceeding the limits specified in Appendix B, Table 2 of 10 CFR Part 20 (10 CFR Part 20.1302[c]). For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of the MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in the ODCM. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrems/year to the whole body or to less than or equal to 3000 mrems/year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/year. A.7-2 ODCM Rev. 25

TABLE A.7.1-1 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Gaseous Release Type Sampling Frequency Minimum Analysis Type of Activity Lower Limit of Frequency .... Analysis Detection(') (LLD) (jxCi/cc)

1. Plant Vent M(3X4) M Principal Gamma 1x10 4

             .            Grab Sample                           -Emitters(2).
                                                                 .H-3                        lxio06 Continuous 5 )            WI6           -131                      lx10.12 Charcoal Sample Continuousl5                (6)        Principal Gamma            lxlO"1 Particulate Sample   Emitters(2)

Continuous(5) M Gross Alpha 1x10 1' Composite

                           . <Particulate                Sample Continuous 5               Q           Sr-89, Sr-90               lx10 "
                        .             .Composite Particulate Sample

2. Condenser Air M( . M(7) Principal Gamma - lxlO Removal Exhaust Grab Sample Noble Gases Emitters(2 )

H-3 1x1O6 A.7-3 . ODCM Rev. 25

TABLE A.7.1-1 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM (Continued) Gaseous Release Type Sampling Frequency Minimum Analysis Type of Activity Lower Limit of Frequency Analysis Detection(l) (LLD) (jICi/cc)

3. Gland Steam Packing Continuous W Principal Gamma lx 10" Exhauster Particulate Sample Emitters(2)

Continuous W 1-131 1x10 -12 Charcoal Sample Continuous M Gross Alpha lxi 0"- Composite Particulate Sample Continuous Q Sr-89, Sr-90 1x1O0 Composite Particulate Sample(8)

4. Containment Purge p(3) P Principal Gamma 1x10 4 Each Purge Grab Each Purge Emittcrs 2 )

Sample H-3 (oxide) 1x10 6 A.7-4 ODCM Rev. 25

TABLEA.7.1-1 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM' (Continued) Notations (I) The LLD is defined, for'purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95 percent probability with only 5 percent probability of falsely concluding that a blank observation represents a "real' signal.. - For a particular measurement system, which may include radiochemical separation: T.TD- ~4.66 Sb--; - .. . - ExVx2.22xiO 6 xY xexp(-At) Where: LLD the "a priori" lower limit of detection (microcurie per unit mass or volume), 4.66 = a constant derived from the Kaipha and Kb,. values for the 95% confidence level; Sb the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute), E = the counting efficiency (counts per disintegration), V = the sample size (units of mass or volume), 2.22 x 106 = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable, the radioactive decay constant for the particular radionuclide (sl), and

                                       .. . ,.    -              W.
                                                                -o.   ..            -   ,

At = the elapsed time between the midpoint of sample collection and the time of counting(s). ' Typical values of E, V, Y. and A t should be used in the calculation. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

              .    .                 '     1.

{! . ' ' ' - .. A.7-5 ODCM Rev. 25

__ _I_ I TABLE A.7.I -I RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM (Continued) Notations (Continued) (2) The principal gamma emitters for which the LLD specification applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, I-131, Cs-134, Cs-137, Ce-141 and Ce-144 in iodine and particulate releases. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radioactive Effluent Release Report in accordance with Technical Specification 6.8.1.4 and Part A, Section 10.2 of the ODCM. Isotopes which are not detected should be reported as "not detected." Values determined to be below detectable levels are not used in dose calculations. (3) Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER within a one hour period unless; 1) analysis shows that the DOSE EQUIVALENT I-131 concentrations in the primary coolant has not increased more than a factor of 3; 2) the noble gas activity monitor for the plant vent has not increased by more than a factor of 3. For containment purge, requirements apply only when purge is in operation. (4) Tritium grab samples shall be taken at least once per 24 hours when the refueling canal is flooded. (5) The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Controls C.7.1.1, C.7.2.1, and C.7.3.1. (6) Samples shall be changed at least once per seven (7) days and analyses shall be completed within 48 hours after changing, or after removal from sampler. Sampling shall also be performed at least once per 24 hours for at least seven (7) days following each shutdown, startup, or THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER within a one-hour period and analyses shall be completed within 48 hours of changing. When samples collected for 24 hours are analyzed, the corresponding LLDs may be increased by a factor of 10. This requirement does not apply if 1) analysis shows that the DOSE EQUIVALENT I-13 1 concentration in the reactor coolant has not increased more than a factor of 3; and 2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3. (7) Samples shall be taken prior to start-up of condenser air removal system when there have been indications of a primary to secondary leak. (8) Quarterly composite analysis requirements shall only be required when analysis for principal gamma emitters indicate positive radioactivity. A.7-6 ODCM Rev. 25

7.2 Dose - Noble Gases CONTROLS

       .. . .     ..                1              9 C.7.2.1         The air dose due to noble gases released in gaseous effluents to areas at and beyond the SlTE BOUNDARY (see Technical Specification Figure 5.1-1) shall be limited to the following:        -'                            -
                ....a. During any calendar quahe'r Less than or equal to 5 mrads for gamma radiation and less than or equal to' 10 mrads for beta radiation, and

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

APPLICABILITY: At all times. -. ACTION:

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

SURVEILLANCE REQUIREMENTS S.7.2.1 Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters in Part B of the ODCM at least once per 31 days. A.7-7 ODCM Rev. 25

_- ll BASES This Control is provided to implement the requirements of Sections II.B, III.A, and IV.A of Appendix I to 10 CFR Part 50. The Control implements the guides set forth in Section I.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I at the SITE BOUNDARY that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept as low as reasonably achievable. The Surveillance Requirements implement the requirements in Section HLI.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977, and Regulatory Guide 1.1 11, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977. The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions. A.7-8 ODCM Rev. 25

7.3 Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form CONTROLS C.7.3.1 The dose to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released to areas at and beyond the SITE BOUNDARY (see Technical Specification Figure

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

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

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

APPLICABILITY: At all times. ACTION: With the calculated dose from the release of Iodine-131, Iodine-133, tritium, and radionuclides in particulate formrwith half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.8.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits. SURVEILLANCE REOUIREMENTS S.7.3.1 Cumulative dose contributions for the current calendar quarter and current calendar year for Iodine-1 31, Iodine-I 33, tritiurnarid radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters in Part B of the ODCM at least once per 31 days. A.779 ODCM Rev. 25

I I BASES This Control is provided to implement the requirements of Sections II.C, L.A, and IV.A of Appendix I to 10 CFR Part 50. The Controls are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents at the SITE BOUNDARY will be kept as low as reasonably achievable. The ODCM calculation methods specified in the Surveillance Requirements implement the requirements in Section M.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977, and Regulatory Guide 1.1 11, "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 Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclide pathways to man in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of the calculations were (1) individual inhalation of airborne radionuclides, (2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition of radionuclides onto grassy areas where milk animals and meat-producing animals graze followed by human consumption of that milk and meat, and (4) deposition of radionuclides on the ground followed by subsequent human exposure. A.7-10 ODCM Rev. 25

7.4 Gaseous Radwaste Treatment System CONTROLS

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

a. '0.2 mrad to air from gamma radiation, or:

b. 0.4 mrad to air from beta radiation, or

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

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

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

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

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

SURVEILLANCE REQUIREMENTS S.7.4.1 Doses due to gaseous releases from each unit to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days in accordance with the methodology and parameters in Part B of the ODCM when Gaseous Radwaste Treatment Systems are not being fully utilized. S.7.4.2 The installed VENTILATION EXHAUST TREATMENT SYSTEM and GASEOUS RADWASTE TREATMENT SYSTEM shall be considered OPERABLE by meeting Controls C.7. 1.1, and C.7.2. 1, or C.7.3.1. A.7-1 1 ODCM Rev. 25

BASES The OPERABILITY of the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTELATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept as low as is reasonably achievable. This Control implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section ll.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections ll.B and Il.C of Appendix I to 10 CFR Part 50, for gaseous effluents. A.7-12 ODCM Rev. 25

l8.0- TOTAL DOSE (JUN I KUL C.8.1.1 'The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25.mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. APPLICABILITY: At all times. ACTION: With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding'twice the limits of Controls C.6.2.l.a, C.6.2.l.b, C.7.2.I.a, C.7.2.1.b, C.7.3.1.a, or C.7.3.l.b, calc'ulations shall be made including direct radiation contributions from the units 'arid from outside storage tanks to determine whether the aboove limits of Control C.8.1.1 have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to' Technical Specification 6.8.2,'a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent - recurrence of exceeding the above limits'and includes the schedule for achieving

            ' conformance with the above limits. This Special Report, as defined in 10 CFR 20.405(c);

shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose(s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete. SURVEILLANCE REOUIREMENTS S.8.1.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Surveillance Requirement S.6.2.1, S.7.2.1, and S.7.3.1, and in accordance with the methodology and parameters in Part B of the ODCM. S.8.1.2 Cumulative dose contributions from direct radiation from the units and from radwaste storage tanks shall be determined in accordance with the methodology and parameters in Part B of the ODCM. This requirement is applicable only under conditions set forth in ACTION a. of Control C.8.1.1. A.8-1 ODCM Rev. 24

BASES This Control is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46FR18525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses due to releases of radioactivity and to radiation from uranium fuel cycle sources exceed 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix L and if direct radiation doses from the units (including outside storage tanks, etc.) are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site are within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Controls C.6.1.1 and C.7.1.1. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle. A.8-2 ODCM Rev. 24

9.0 RADIOLOGICAL ENVIRONMENTAL MONITORING 9.1 ' Monitoring Program .' COUNIKUL - -,--- C.9.1.1 The Radiological Environmental Monitoring Program (REMP) shall be conducted as specified in Table A.9.1-l. - APPLICABILITY: At all times. ACTION: - --: -

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

b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table A.9.1-3 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days from receipt of the laboratory analyses, pursuant to Technical Specification 6.8.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines' the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose* to a MEMBER OF THE PUBLIC is less than the calendar year limitsof Control C.6.2;1, C.7.2.1, or C.7.3.1. When more than one of the radionuclides in the REMP are detected in the sampling medium, this report shall be submitted if

            -         concentration ()               'concentration (2) reporting level (1)    '             reporting level (2)   + *-->1.0 When radionuclides other than those listed in the REMP are detected and are the, result of plant effluents, this report shall be submitted if the potential annual dose*

to a MEMBER OF THE PUBLIC from all radionuclides is equal to or greater than the calendar year limits of Co-ntrol 'C.6.2.1, C.7.2.1, or C.7.3.1. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report required by Technical Specification 6.8.1.3 and Part A, Section 10.1 of the ODCM.

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

                                                         ,A.9-1                                ODCM Rev. 261

ACTION: (Continued) With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by the REMP, identify specific locations for obtaining replacement samples and add them within 30 days to the REMP given in the ODCM. The specific locations from which samples were unavailable may then be deleted from the monitoring program. Pursuant to Technical Specification 6.13, and Part A, Section 10.2, of the ODCM, submit in the next Annual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure(s) and table for the ODCM reflecting the new location(s) with supporting information identifying the cause of the unavailability of samples and justifying the selection of the new locations(s) for obtaining samples. SURVEILLANCE REQUIREMENTS S.9.1.1 The radiological environmental monitoring samples shall be collected pursuant to' Table A.9.1-1 from the specific locations given in the table and figure(s) in Part B of the ODCM, and shall be analyzed pursuant to the requirements of Table A.9.1-l and the detection capabilities required by Table A.9.1-2. BASES The REMP required by this Control provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting from the plant operation. This monitoring program implements Section IV.B.2 of Appendix I to 10 CFR Part 50, and thereby supplements the REMP by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1, November 1979. The initially specified monitoring program will be effective for at least the first 3 years of commercial operation. Following this period, program changes may be initiated based on operational experience. Detailed discussion of the LLD and other detection limits can be found in Currie, L.A., "Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements," NUREG/CR-4007 (September 1984). A.9-2 ODCM Rev. 261

TABLE A.9.1I-1

                           RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway and/or    Number of Representative Samples and         Sampling and Collection       Type and Frequency of Sample                        Sample Locations8                         Frequency                     Analysis I DIRECT RADIATIONb    40 routine monitoring stations with two or      Quarterly.                Gamma dose quarterly.

more dosimeters placed as follows: An inner ring of stations, one in each meteorological sector in the general area'of the SITE BOUNDARY; An outer ring of stations, one in each meteorological sector, generally in the 6 to 8-kmn range from the site; The balance of the stations to be placed in special interest areas, such as population centers, nearbyi residences, schools, and control locations.

2. AIRBORNE Radioiodine and Samples from five locatio'nsd: Continuous sampler Radioiodine Canister:

Particulates operation with sample Three samples from close to the three SITE collection weekly, or I-13 1 analysis weekly. BOUNDARY locations, in different sectors, more frequently if-- of high 'alculated~ long-term average required by dust loading. PriuaeSmlr groud-lvel IQ;Gross beta radioactivity One sample from the vicinity of a analysis following filter community having the highest calculated . change;. long-term average groun'd-level D/Q. . Gamma isotopic analysis' of composite (by'location), quarterly. A.9-3 A.9-3ODCM Rev. 26 1

TABLE A.9.1-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (Continued) Exposure Pathway and/or Number of Representative Samples and Sampling and Collection Type and Frequency of Sample Sample Locations' Frequency Analysis

2. (Continued) One sample from a control location, as for example 15-30 km distant and in the least prevalent wind direction.

3. WATERBORNE

a. Surface One sample in the discharge area. One Monthly grab sample. Gamma isotopic analysis' sample from a control location. monthly. Composite for tritium analysis quarterly.

b. Sediment from One sample from area with existing or Semiannually. Gamma isotopic analysise shoreline potential recreational value. semiannually.

4. INGESTION

a. Milk Samples from milking animals in three Semimonthly when Gamma isotopic' and I-13 1 locations within 5 km distance having the milking animals are on analysis on each sample.

highest dose potential. If there are none, pasture, monthly at other then, one sample from milking animals in times. each of three areas between 5 to 8 km distant where doses are calculated to be greater than 1 mrem per yrf One sample from milking animals at a control location, as for example, 15-30 km distant and in the least prevalent wind direction. A.9-4 ODCM Rev. 26 l

TABLE A.9.I-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (Continued) Exposure Pathway and/or Number of Representative Samples and Sampling and Collection. Type and Frequency of Sample Sample Locationsa Frequency Analysis

4. (Continued)' One sample of each of three commercially Sample in season, or Gamma isotopic analysise on and recreationally important species in semiannually if they are edible portions.

b. Fish v and vicinity of plant discharge area. not seasonal.

1inve'rtebrates'§

               ;f;.               *2, . ;h One sample of similar species in areas not influenced by plant discharge.

c.' Food Pr6ducts Samples of three (if practical) different Monthly, when available. Gamma isotopic' and I-131

                          -  kinds of broad leafvegetation3 grown -:                               analysis.

nearest each of two different off-site locations of highest predicted long-term. average ground-level D/Q if milk sampli'ig

                       . is not performed.
                         .-  One sample of each ofthe similar broad leaf Monthly, when available. Gamma isotopice and I-131 .-

vegetation3 grown at a control lo'catlon, as analysis. for example 15-30 km distant in the least prevalent wind direction, if milk sampling is not performed. - A.9-5 ODCM Rev. 26 l

TABLE A.9.1-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (Continued) Table Notations

a. Specific parameters of distance and direction sector from the centerline of the Unit I reactor, and additional description where pertinent, shall be provided for each and every sample location in Table B.4-1 in the ODCM, Part B. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to circumstances such as hazardous conditions, seasonal unavailability and malfunction of automatic sampling equipment. If specimens are unobtainable due to sampling equipment malfunction, effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report as specified in Part A, Section 10.1. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the Radiological Environmental Monitoring Program. Identify the cause of the unavailability of samples for that pathway and identify the new location(s), if available, for obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report as specified in Part A, Section 10.2 and also include in the report a revised figure(s) and table for the ODCM reflecting the new location(s).

b. A thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters.

c. Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than ten times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.

d. Optimal air sampling locations are based not only on D/Q but on factors such as population in the area, year-round access to the site, and availability of power.

e. Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.

f. The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the ODCM, Part B. With no calculated dose rates greater than I mrem, the highest calculated dose rate shall be used.

g. If broad leaf vegetation is unavailable, other vegetation will be sampled.

A.9-6 ODCM Rev. 26

TABLE A.9.1-2 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS3 f Lower Limit of Detection (LLD)b Fish and Water Airborne Particulate or Invertebrates Milk Food Products Sediment Analysis (pCi/lkg) Gas (pCi/kg, wet) (pCi/kg, wet) (pCi/kg) (pCi/kg, wet) (pCi/kg, dry) Gross Beta 4 0.01 H-3 3,000 Mn-54 15 130 Fe-59 30 . 260 Co-58, 60 15 30 130 15 Zn-65 30 260 IS.O Zr-Nb-95 15C 18. I-131 15 0.07 I 60e 60 Cs-134 15 0.05 130 15 60 - Cs-137 0.06 150 18 80

          . 5~d:' .

I Ba-La-140 15 cd 1I¢, I A.9-7 ODCM Rev. 26 l

                                                                                      .   -LL   _    -

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

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

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

For a particular measurement system, which may include radiochemical separation: LLD = 4.66 Sb ExVx 2.22 x10 6 x Yxexp(-Att) Where: LLD is the "a priori" lower limit of detection as defined above, as picocuries per unit mass or volume; 4.66 is a constant derived from the KaIpha and Kbt. values for the 95% confidence level; sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate, as counts per minute; E is the counting efficiency, as counts per disintegration; V is the sample size in units of mass or volume; 2.22 is the number of disintegrations per minute per picocurie; Y is the fractional radiochemical yield, when applicable; X is the radioactive decay constant for the particular radionuclide as per second; and A t for environmental samples is the elapsed time between sample collection and time of counting, as seconds. Typical values of E, V, Y, and A t should be used in the calculation. In calculating the LLD for a radionuclide determined by gamma ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples (e.g., Potassium-40 in milk samples). A.9-8 ODCM Rev. 26 l

TABLE A.9.1-2 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS (Continued) Table Notations (Continued) It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a - particular measurement. This does not preclude the calculation of an a posteriori LLD for a particular measurement based upon the actual parameters for the sample in question and appropriate decay correction parameters such as decay while sampling and during analysis. Analyses shall be performed in such a manner that the stated LLDs will be achieved under. routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LL.Ds unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report per Part A, Section 10.1.

c. Parent only. -

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

e. Broad leaf vegetation only.

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

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

A.9-9 ODCM Rev. 26 l

TABLE A.9.1-3 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Analysis Water Airborne Particulate or Gas Fish and Milk Food Products (pCi/kg) (pCi/kg, wet) Invertebrates (pCi/kg) (pCi/kg, wet) (pCi/kg, wet) H-3 30,000*** Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000 Zn-65 300 20,000 Zr-Nb-95 400* I-131 100 0.9 3 100** Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-La-140 200* 300* $ Parent only.

• • Broad leaf vegetation only.
    • Plant dewatering reporting level = 20,000 pCi/kg (2E-05 piCi/ml)

A.9-10 ODCM Rev. 26 l

9.2 - Land Use Census CONTROL C.9.2.1 A Land Use Census shall be' conducted and shall identify within a distance' of 8 km (5 miles) the location in each ofthe 16 meteorological sectors of the nearest milk animal,' the nearest residence, and the nearest garden** of greater than 50 m2 (500 ft2) producing broad leaf vegetation. APPLICABILlTY: At all times.. ACTION

a. With a Land Use Census identifying a location(s) that yields a calculated'dose or dose commitment greater than the values currently being calculated in Surveillance S.7.3.1 pursuant to Technical Specification 6.8.1.4 and Part A, Section 10.2, of the ODCM, identify the new location(s) in the next Annual Radioactive Effluent Release Report.

b. With a Land Use Census identifying a location(s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordance with Control C.9. 1.1,add the new location(s) within 30 days to the REMP given in the ODCM, if permission from the owner to collect samples can be obtained and sufficient sample volume is available. The sampling location(s), excluding the Control station location, having the lowest calculated dose or dose commitment(s),

via the same exposure pathway, may be deleted from this monitoring program after October 31 of the year in which this Land Use Census was conducted. Pursuant to Technical Specification 6.13 and Part A, Section 10.2 of the ODCM, submit in the next Annual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure(s) and table(s) for the ODCM reflecting the new location(s) with information supporting the change in sampling locations. SURVEILLANCE REQUIREMENTS S.9.2.1 The Land Use Census shall be conducted during the growing season at least once per 12 months using a method such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities, as described in the ODCM. The results of the Land Use Census shall be included in the Annual Radiological Environmental Operating Report pursuant to Part A, Section 10.1 of the ODCM.

• • Broad leaf vegetation sampling of at least three different kinds of vegetation may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted relative deposition values (D/Qs) in lieu of the garden census. Specifications for broad leaf vegetation sampling in the REMP shall be followed, including analysis of control samples.

A.9-11 ODCM Rev. 26 l

I i BASES This specification is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the REMP given in the ODCM are made if 0 required by the results of this census. Information from methods such as the door-to-door survey, from aerial survey, of from consulting with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 50 m2 provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored, since a garden of this size is the minimum required to produce the quantity (26 kg/year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were made: (1) 20% of the garden was used for growing broad-leaf vegetation (i.e., similar to lettuce and cabbage), and (2) there was a vegetation yield of 2 kg/m 2 . A.9-12 ODCM Rev. 26 l

9.3 Interlaboratory Comparison Program CONTROL C.9.3.1 In accordance with Technical Specification 6.7.6h.3, analyses shall be performed on all radioactive materials supplied as part of an Interlaboratory Comparison Program, that has been approved by the Commission, that correspond to samples required by REMP. APPLICABILITY: At all times. ACTION: With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to Part A, Section 10.1 of the ODCM. SURVEILLANCE REOUIREMENTS S.9.3.1 The Interlaboratory Comparison Program shall be identified in Part B of the ODCM. A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to Part A, Section 10.1 of the ODCM. BASES The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the Quality Assurance Program for environmental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10 .CFR Part 50. A.9-13 ODCM Rev. 26 l

 . 10.0   REPORTS
  *10.1   Annual Radiological Environmental Operating Report Routine Annual Radiological Environmental Operating Reports covering the operation of the station during the previous calendar year shall be submitted prior to May I of each year pursuant to Technical Specification 6.8.1.3.'

The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, aid an analysis of trends of the results of the radiological environmental Surveillance activities for the report period, including a comparison with preoperational studies, with opeiational Controls, as appropriate, and with previous environmental Surveillance reports, and an assessment of the observed impacts of the plant operation on the environment.' The reports shall also include the results of the Land Use Census required by Control C.9.2.1. The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples 'ad of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in Part B of the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report. The reports shall also include the following: a summary description of the Radiological Environmental Monitoring Program; at least two legible maps**** covering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor; the results of licensee participation in the Interlaboratory Comparison Program and the corrective action taken if the specified program is not being performed as required by Control C.9.3.1; reason for not'conducting the Radiological Environmental Monitoring Program as required by Control C.9.1.1, aid discussion of all deviations from the sampling schedule; discussion of environmental sample measurements that exceed the reporting levels but are not the result of plant effluents, pursuant to ACTION b. of Control C.9.1.1; and discussion of all analyses in which the LLD required was not achievable.; . ****One map shall cover locations near the SITE BOUNDARY; the more distant locations shall be covered by one or more additional maps. A.10-1 ODCM Rev. 24

10.2 Annual Radioactive Effluent Release Report A routine Annual Radioactive Effluent Release Report covering the operation of the station during the previous calendar year of operation shall be submitted by May 1 of each year, pursuant to Technical Specification 6.8.1.4. The Annual Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the station as outlined in Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof. For solid wastes, the format for Table 3 in Appendix B shall be supplemented with three additional categories: class of solid wastes (as defined by 10 CFR Part 61), type of container (e.g., LSA, Type A, Type B, Large Quantity) and SOLIDIFICATION agent or absorbent (e.g., cement). The Annual Radioactive Effluent Release Report shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form ofjoint frequency distributions of wind speed, wind direction, and atmospheric stability.***** This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY Technical Specification (Figure 5.1-3) during the report period. All assumptions used in making these assessments, i.e., specific activity, exposure time, and location, shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM). The Annual Radioactive Effluent Release Report shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190, "Environmental Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, October 1977.

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

A.10-2 ODCM Rev. 24

The Annual Radioactive Effluent Release Report shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period. The Annual Radioactive Effluent Release Report shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM and the ODCM, pursuant to Technical Specifications 6.12 and 6.13, respectively, as well as any major change to Liquid, Gaseous, or Solid Radwaste Treatment Systems pursuant to Control 11.0. It shall also include a listing of new locations for dose calculations and/or environmental monitoring identified by the Land Use Census pursuant to Control C.9.2. The Annual Radioactive Effluent Release Report shall also include the following: an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Control C.5.1 or C.5.2, respectively-, and description of the events leading to liquid holdup tanks or gas storage tanks exceeding the limits of Technical Specification 3.11.1.4. A.10-3 ODCM Rev. 24

             -,   .        I  '       ;
.-    . SEABROOK STATION ODCM PART B RADIOLOGICAL CALCULATIONAL METHODS AND PARAMETERS II I i
                . I . ..   . - , : 1; I -., .

I

                                                     . ,  . , I.

I, . I B.1 - ODCM Rev. 25

1.0 INTRODUCTION

The Offsite Dose Calculation Manual (ODCM) contains details to implement Radioactive Effluent Controls and Environmental Monitoring Program as required by Technical Specifications 6.7.6g and 6.7.6h. Part B of the ODCM provides formal and approved methods for the calculation of off-site concentration, off-site doses and effluent monitor setpoints, and indicates the locations of environmental monitoring stations in order to comply with the Seabrook Station Radioactive Effluent Controls Program (RECP), and Radiological Environmental Monitoring Program (REMP) detailed in Part A of the manual. The ODCM forms the basis for station procedures which document the off-site doses due to station operation which are used to show compliance with the numerical guides for design objectives of Section II of Appendix I to I OCFR Part 50. The methods contained herein follow accepted NRC guidance, unless otherwise noted in the text. The references to 10 CFR Part 20 in Part B of the ODCM refer to revisions of 10 CFR Part 20 published prior to I January 1993. The decision to continue the use of the "old" version of 10 CFR Part 20 is based on an NRC letter dated June 30, 1993, from Thomas E. Murley to Thomas E. Tipton. For the convenience of the plant staff a copy of 10 CFR Part 20 (Rev. 1, January 1992) has been included in Appendix B. 1.1 Responsibilities for Part B All changes to the ODCM shall be reviewed by the Station Operation Review Committee (SORC), approved by the Station Director, and documented in accordance with Technical Specification 6.13. The change process is controlled by the Applicability Determination Process as controlled by the 10 CFR 50.59 Resource Manual (5059RM). Changes made to Part B shall be submitted to the Commission for their information in the Annual Radioactive Effluent Release Report for the period in which the change(s) was made effective. It shall be the responsibility of the Station Director to ensure that the ODCM is used in the performance of surveillance requirements and administrative controls in accordance with Technical Specifications 6.7.6g and 6.7.6h, and Effluent Control Program and Radiological Environmental Monitoring Program detailed in Part A of the manual. In addition to off-site dose calculations for the demonstration of compliance with Technical Specification dose limits at and beyond the site boundary, 10 CFR 20.1302 requires that compliance with the dose limits for individual members of the public (100 mrem/yr total effective dose equivalent) be demonstrated in controlled areas on-site. Demonstration of compliance with the dose limits to members of the public in controlled areas is implemented per Health Physics Department Procedures, and is outside the scope of the ODCM. However, calculations performed in accordance with the ODCM can be used as one indicator of the need to perform an assessment of exposure to members of the public within the site boundary. Since external direct exposure pathways are already subject to routine exposure rate surveys and measurements, only the inhalation pathway need be assessed. The accumulated critical organ dose at the site boundary, as calculated per ODCM Part B Sections 3.9 and 3.1 1, can be used as an indicator of when additional assessments of on-site exposure to members of the public is advisable (see Section 3.11.2). Off-site critical organ doses from station effluents should not, however, be the only indicator of potential on-site doses. B.1-1 ODCM Rev. 25

1.2 Summary of Methods, Dose Factors, Limits, Constants, Variables and Definitions This section summarizes the Method I dose equations which are used as the primary means of demonstrating compliance with RECP. The concentration and setpoint methods are identified in Table B.1-2 through Table B.1-7. Appendix C provides documentation for an alternate computerized option, designated as Method IA in the ODCM, for calculating doses necessary to demonstrate compliance with RECP. The Effluent Management System (EMS) software package used for this purpose is provided by Canberra Industries, Inc. Where more refined dose calculations are needed, the use of Method II dose determinations are described in Sections 3.2 through 3.9 and 3.11. The dose factors used in the equations are in Tables B.1-lO through B.1-14 and the Regulatory limits are summarized in Table B.1-1. The variables and special definitions used in this ODCM, Part B, are in Tables B.1-8 and B.A-9. ( .. B.I-~2 ODCM Rev. 25

TABLE B.1-1

SUMMARY

OF RADIOLOGICAL EFFLUENT PART A CONTROLS AND IMPLEMENTING EOUATIONS Part A Control Category Method I0) Limit I C.6. 1.1 Liquid Effluent Total Fraction of ECL Eq.2-1 <1.0 Concentration Excluding Noble Gases Total Noble Gas Concentration Eq. 2-2 < 2 x 10 4 Ci/ml C.6.2.1 Liquid Effluent Dose Total Body Dose Eq. 3-1 < 1.5 mrem in a qtr.

                                                                                       <3.0 mrem in a yr.

Organ Dose Eq. 3-2 < 5 mrem in a qtr.

                                                                                       < 10 mrem in a yr.

C.6.3. 1 Liquid Radwaste Total Body Dose Eq. 3-1 < 0.06 mrem in a mo. Treatment Operability Organ Dose Eq. 3-2 <0.2 mrem in a mo. C.7.1.1 Gaseous Effluents Total Body Dose Rate from Eq. 3-3 < 500 mrem/yr. Dose Rate Noble Gases Skin Dose Rate from Noble Eq. 3-4 < 3000 mrem/yr. Gases Organ Dose Rate from I-13 1, Eq. 3-5 < 1500 mrem/yr. 1-133, Tritium and Particulates with T112> 8 Days B. 1-3 ODCM Rev. 25

TABLE B.1-l

SUMMARY

OF RADIOLOGICAL EFFLUENT PART A CONTROLS AND IMPLEMENTING EOUATIONS (Continued) Part A Control Category Method p Limit C.7.2.1 Gaseous Effluents Gamma Air Dose from Noble Eq. 3-6 < 5 mrad in a qtr.

                                                                                        .. . Il  ,I Dose fromn Noble
         .Ga.                       Gases --

Gases -10 mrmd in a yr. Beta Air Dose from Noble Eq. 3-7 < I0 mrad in a qtr. Gases

                                                                                        < 20 mrad in a yr.

C.7.3.1 Gaseous Effluents Organ Dose from Iodines, Eq. 3-8 < 7.5 mrem in a qtr. Dose from 1-131, Tritium and Particulates with 1-133, Tritium, and Tin > 8 Days <15 mrem in a yr. Particulates C.7.4.1 Ventilation Organ Dose Eq. 3-8 < 0.3 mrem in a mo. Exhaust Treatment C.8.1.1 Total Dose (from All Total Body Dose Footnote (2). < 25 mrem in a yr. Sources) Organ Dose < 25 mrem in a yr. z .. I " .. Thyroid Dose <75 mrem in a yr. C.5.1 Liquid Effluent Monitor Setpoint Liquid Waste Test Alarm Setpoint Eq. 5-1 Control C.6.1.1 Tank Monitor. ;- - -

                            ~~~. ..        . . .1  :       . .  '   ' . I B.1-4                                  ODCM Rev. 25

TABLE B.1-l

SUMMARY

OF RADIOLOGICAL EFFLUENT PART A CONTROLS AND IMPLEMENTING EQUATIONS (Continued) Part A Controls Category Method i(') Limit C.5.2 Gaseous Effluent Monitor Sctpoint Plant Vent Wide Alarm/Trip Setpoint For Total Eq. 5-9 Control C.7.l.Ia Range Gas Monitors Body Dose Rate (Total Body) Alarm/Trip Setpoint for Skin Eq. 5-10 Control C.7. 1.1 a Dose Rate (Skin) (1) More accurate methods may be available (see subsequent chapters). (2) Part A Control C.8.1.Ia requires this evaluation only if twice the limit of equations 3-1, 3-2, 3-12, 3-15 or 3-18 is reached. If this occurs a Method IIcalculation, using actual release point parameters with annual average or concurrent meteorology and identified pathways for a real individual, shall be made. B. 1-5 ODCM Rev. 25

TABLE B.1-2

SUMMARY

OF METHOD I EQUATIONS TO CALCULATE UNRESTRICTED AREA LIQUID CONCENTRATIONS Equation Number Category Equation 2-1 Total Fraction of ECL in FENG=Z ~P 1 Liquids, Except Noble Gases p i ECLi 2-2 Total Activity of Dissolved and Entrained Noble Gases from all ( pNiG N Station Sources I ~.ml )-' 2E- 04 B.1-6 ODCM Rev. 25

Il TABLE B.1-3

SUMMARY

OF METHOD I EQUATIONS TO CALCULATE OFF-SITE DOSES FROM LIQUID RELEASES Equation Number Category Equation 3-1 Total Body D tb (mrem) = k E Q.i DFL itb Dose i 3-2 Maximum D MO(mrem) = k YQ.iDFLMO Organ Dose i B.1-7 ODCM Rev. 25

0 TABLE B. 1-4

SUMMARY

OF METHOD I EQUATIONS TO CALCULATE DOSE RATES Categorv Equation Receptor Release Number Location' HeightP Eguation

        Total BodyD'ose         3-3a                OS               E

• DFB 1)

  . - 'Rate From Noble                                                           D)         085  *      (

• DFBi)

Gases 3-3b Os G Dtbg) = 3.4 *

• DFB1) 3-3c EC E DtbE(e) 0.0015 *
• DFBi) 3-3d EC G

                                                                                  ),bEfg  =  0.0074 *Doi

• DFBi) 3-3e R E DtbR(C) = 0.038 *
• DFBi) 3-3f R G DtbR(g) 0.2
• DOI
• DFB)

I "OS -Off-Site, EC =Science & Nature Center, forrnerly the Education Center, R =The "Rocks" bE Elevated, G = Ground B.1-8 ODCM Rev. 25

TABLE B.1-4

SUMMARY

OF METHOD I EOUATIONS TO CALCULATE DOSE RATES (Continued) Categorv Equation Receptor Release Number Location' HeighPb Eauation Skin Dose Rate 3-4a Os E From Noble Gases Dskm) = (0j

• DF' Ke))

3-4b OS G D kinl,) 5 = (Q

• DF' (j))

3-4c EC E DskinE(e) = 0.0014 * (j* DF' iE()) 3-4d EC G DskinEg) = 0.0014* (,

• DF'jExg))

3-4e R E DbkinR,¢) = 0.0076

• DQi
• DF' iR(e))

3-4f R G Dkijmg) = 0.0076 * *(iD iR(g)) 'OS = Off-Site, EC = Science & Nature Center, formerly the Education Center, R = The "Rocks" bE = Elevated, G = Ground B.1-9 ODCM Rev. 25 0

TABLE B. 1-4

SUMMARY

OF METHOD I EOUATIONS TO CALCULATE DOSE RATES (Continued) Categorv Equation Receptor Release Number Location' Heightb Equation Critical Organ Dose 3-5a OS. *E Rate From I-131, Dcv(e) , j (0j

• DFG' iv¢))

1-133, H-3, and Particulate With Tl/ >8 Days 3-5b OS G Deo(z) = ( j* DFG' ;.d) 3-5c EC E DfOEe) = 0.0014

• Z(Q
• DFG' icoEe))

3-5d EC G DCOE(s) = 0.0014 * *DFG' icoE(g)) 3-5e R E DcoR(e) = 0.0076

• 2 (Q.* DFG' koR(e))

3-5f R G i 'OS = Off-Site, EC = Science & Nature Center, formerly the Education Center, R = The "Rocks" bE = Elevated, G = Ground B.1-10 ODCM Rev. 25

TABLE B.G-5

SUMMARY

OF METHOD I EQUATIONS TO CALCULATE DOSES TO AIR FROM NOBLE GASES Category Equation Receptor Release Number Locationa Heightb Equation Gamma Dose to Air 3-6a OS E From Noble Gases Dyr() = 3.2 E-07

• t 0 275
• Z(Q;
• DFOD 3-6b Os G Dliqg) = 1.6E-06
• t'0*293
• Z(Qj
• DFI) 3-6c EC E D&E(.) = 4.9 E-10
• t0252 * (Qj
• DFD) 3-6d EC G D&L t) = 4.4E-09
• t4'`32
• EQR
• DFr) 3-6e R E D)&IR(,) = 5.1E-09
• t 0 '155* Z(Q
• DFO 3-6f R G DarlR(g) = 4.1 E- 08
• t-0.204 * (Q;
• D~R) aOs = Off-Site, EC = Science & Nature Center, formerly the Education Center, R = The "Rocks" bE = Elevated, G = Ground B.1-1l ODCM Rev. 25 0

TABLE B.1-5

SUMMARY

OF METHOD I EQUATIONS TO CALCULATE DOSES TO AIR FROM NOBLE GASES (Continued) Equation Receptor Release Number Locationa Heightb Equation Beta Dose to Air 3-7a OS E D!,l() = 4.1 E- 07

• t03
• j(Qj
• DFf)

From Noble Gases 3-7b OS G DITg) = 6.OE-06

• t'3"'
• F(Qi
• DFe) 3-7c EC E Die) = 1.8E-09
• t-035
• 1(Q1
• DFf')

3-7d EC G D.rtg) = 2.4 E-08

• t1037 * (Q;
• DFf')

3-7e R E DLrR(e) = 3.9E-08

• t4244
• Z(Q
• DFW) 3-7f R G D&rR(g) = 4.6 E- 07
• t -0 267
• E(Q
• DFf) aoS = Off-Site, EC = Science & Nature Center, formerly the Education Center, R = The "Rocks" bE = Elevated, G= Ground B.1-12 ODCM Rev. 25

TABLE B. 1-6

SUMMARY

OF METHOD I EOUATIONS TO CALCULATE DOSE TO AN INDIVIDUAL FROM TRITIUM. IODINE AND PARTICULATES Categorv Equation Receptor Release Number Locationa Heightb Equation Dose to Critical 3-8a Os E Do(e) = 14.8

• t40.297 *E(Q
• DFGk ())

Organ From Iodines, Tritium, and Particulates 3-8b OS G D.g) = 17.7

• t4-"' 6 *x(Qi
• DFGos 8 ))

3-8c EC E DcoE(e) = 3.3E- 02

• l4J49 * (Q
• DFGioE(,))

3-8d EC G D.E($) = 3.3 E- 02

• t 0.347
• E(Qi
• DFGicOE(s))

3-8e R E D,.R(O)= 7.3 E- 02

• t40*248 *2:(Q. *DFGioR(t))

3-8f R G DCOR(g) = 8.6 E- 02

• t26 * (Qj
• DFGIoR(8))

aos = Off-Site, EC = Science & Nature Center, formerly the Education Center, R = The "Rocks" bE = Elevated, G = Ground B.1-13 ODCM Rev. 25 9

TABLE B.1-7

SUMMARY

OF METHODS FOR SETPOINT DETERMINATIONS Equation Number Category Equation 5-1 Liquid Effluents:. IFd E CY Liquid Waste Test Rsetpointml I Fm X DFmin Tank Monitor (RM-6509) 5-23 PCCW Rate-of-Change RC 1 AlamRC PCc Gaseous Effluents: Plant Vent Wide - Range Gas Monitors (RM-6528-1, 2, 3) 5-5 Total Body

                                          -R, - (/ICi/sec)) = 588    1 DFBr   G 5-6       Skin.
                                        - '=3000 I       :                    -1 f,
                                             ,k (puCilsec)=D -

B.1-14 ODCM Rev. 25

I I TABLE B.1-8

SUMMARY

OF VARIABLES Variable Definition Units CG = Concentration at point of discharge and entrained noble ACi/mi gas "i" in liquid pathways from all station sources Cli coNG Total activity of all dissolved and entrained noble gases PCi/ml in liquid pathways from all station sources Cdi Concentration of radionuclide "i" at the point of liquid PCi/mi discharge Ci Concentration of radionuclide "i" PCi/mI Cpo Concentration, exclusive of noble gases, of pCi/ml radionuclide "i" from tank "p" at point of discharge Cy i Concentration of radionuclide "i" in mixture at the pCi/ml monitor DIr(e) Off-site beta dose to air due to noble gases in elevated mrad release Off-site beta dose to air due to noble gas in ground mrad level release D~fll() Beta dose to air at Science & Nature Center due to mrad noble gases in elevated release Beta dose to air at Science & Nature Center due to mrad noble gases in ground level release Beta dose to air at "Rocks" due to noble gases in mrad elevated release Beta dose to air at "Rocks" due to noble gases in mrad ground level release DArRde) Off-site gamma dose to air due to noble gases in mrad elevated release Off-site gamma dose to air due to noble gases in mrad ground level release DYirE(c) Gamma dose to air at Science & Nature Center due to mrad D."Hq) noble gases in elevated release Gamma dose to air at Science & Nature Center due to mrad noble gases in ground level release Darir~t) Gamma dose to air at "Rocks" due to noble gases in mrad elevated release B.l-15 ODCM Rev. 25

TABLE B.1-8

SUMMARY

OF VARIABLES (Continued) Variable Units DRi) = Gamma dose to air at "Rocks" due to noble gases in mrad ground level release Dco(e) = Critical organ dose from an elevated release to an mrem off-site receptor

              =  Critical organ dose from a ground level release to an       mrem off-site receptor DcOE(C)      =  Critical organ dose from an elevated release to a           mrem receptor at the Science & Nature Center DcoE<g       =  Critical organ dose from a ground level release to a        mrem receptor at the Science & Nature Center DcoR(e)      =  Critical organ dose from an elevated release to a           mrem DCOR(C) receptor at the "Rocks" Dcoftg)         Critical organ dose from a ground level release to a        mrem receptor at the "Rocks" Dd           = Direct dose                                                  mrem Mfiniml      = Gamma dose to air, corrected for finite cloud                 mnrad
              = Dose to the maximum organ                                    mnrem Ds           = Dose to skin from beta and gamma                             mrem Dtb          = Dose to the total body                                       mrem DFmin        = Minimum required dilution factor based on all (beta -         ratio emitting and gamma - emitting) radionuclides
              = Minimum required dilution factor necessary to ensure that the sum of the ratios for the concentration of each I                gamma-emitting radionuclide to the respective ECL value is not greater than 1 (dimensionless).

DF'i = Composite skin dose factor for off-site receptor mrem-sec/pCi-yr DF'iE = Composite skin dose factor for Science & Nature mrem-sec/tCi-yr Center DF'iR = Composite skin dose factor for the "Rocks" mrem-sec/pCi-yr DFBi = Total body gamma dose factor for nuclide "i" (Table mrem3 B.1-10) pCi-yr 1 . ; B.1-16 ODCM Rev. 25

I I TABLE B.1-8

SUMMARY

OF VARIABLES (Continued) 0 Variable L-,-QIJIXLAWz, Units DFBC Composite total body dose factor mrem 3 pCi-yr DFL~t Site-specific, total body dose factor for a liquid release mremlpCil of nuclide "i" (Table B.I-I 1) DFL1,= Site-specific, maximum organ dose factor for a liquid rnrernm/Ci release of nuclide "i" (Table B.I-l1) DFBicc~c) Site-specific, critical organ dose factor for an elevated mremn/tLCi gaseous release of nuclide "i" (Table B.1-12) DFGico~) Site-specific critical organ dose factor for a ground level mrem/ipCi release of nuclide "i" (Table B.1-12) DFGicoE(c) Science & Nature Center-specific critical organ dose mrem/pCi factor for an elevated release of nuclide "i" (Table B.1-14) DFGicoft) Science & Nature Center-specific critical organ dose mrem/jiCi factor for a ground level release of nuclide "i" (Table B.1-14) DFGi,,Re) The "Rocks"-specific critical organ dose factor for an mrem.lCi elevated release of nuclide "i" (Table B.1-15) DFGicoRU The "Rocks"-specific critical dose factor for a ground mrem/pCi level release of nuclide "i" (Table B.1-15) DFG Pcoe) Site-specific critical organ dose rate factor for an mrem-sec/gCi-yr elevated gaseous release of nuclide "i" (Table B. 1-12) DFG'o~g) Site-specific critical organ dose rate factor for a ground mrem-sec/pCi-yr level release of nuclide "i" (Table B.1-12) DFG'icoE(e) Science & Nature Center-specific critical organ dose mrem-sec/gCi-yr rate factor for an elevated release of nuclide "i" (Table B.1-14) DFG 'co E(z) Science & Nature Center-specific critical organ dose mrem-sec/tiCi-yr rate factor for a ground level release of nuclide "i" (Table B.1-14) DFG' sOR(e) The "Rocks"-speciflc critical organ dose rate factor for nmrem-sec/tiCi-yr an elevated release of nuclide "i" (Table B.1-15) DFG' icoRfg) The "Rocks"-specific critical organ dose rate factor for a mrern-sec/pCi-yr ground level release of nuclide "i" (Table B.1-15) B.1-17 ODCM Rev. 25

TABLE B.1-8

SUMMARY

OF VARIABLES (Continued) Variable Definition Units DFS ' Beta skin dose factor for nuclide "i" (Table B.1-10) mrem-m3 p Ci- yr DFj = Combined skin dose factor for nuclide "i" mrem- sec/,Ci-yr (TableB.1-10) DFr Gamma air dose factor for nuclide "i" (Table B.1-10) mrad- m3 pCi-yr DFf = Beta air dose factor for nuclide "i" (Table B.1-1 0) mrad- m3 pCi-yr Dco(e) = Critical organ dose rate to an off-site receptor due to mrem elevated release of iodines, tritium, and particulates yr Dco-E) Critical organ dos6 rate to an off-site receptor due to - mrem ground level release of iodines, tritium, and particulates DcoE(e) Critical organ dose rate to a receptor at the Science & mrern Nature Center due'to an elevated release of iodines, T tritium, and particulates 1E) = Critical organ dose rate to a receptor at the Science & mrem Nature Center due to a ground level release of iodines, yr tritium, and particulates DcoR(c) Critical organ dose rate to a receptor at the "Rocks" due mrem to an elevated release of iodines, tritium, and particulates

                =  Critical organ dose rate to a receptor at the "Rocks" due        mrem to a ground level release of iodines, tritium, and                yr particulates DSkWnC)          =  Skin dose rate to an off-site receptor due to noble gases        mrem in an elevated release                                            yr skin(s)       =  Skin dose rate to an off-site receptor due to noble gases        mrem in a ground level release                                         yr DskinE(e)           Skin dose rate to a receptor at the Science & Nature             mrem Center due to noble gases in an elevated release                  yr b,,E~r)=            Skin dose rate to a receptor at the Science & Nature             mrem Center due to noble gases in a ground level release               yr B.1-18                             ODCM Rev. 25

I I TABLE B.1-8

SUMMARY

OF VARIABLES (Continued) Variable Definition Units DskinRWe) = Skin dose rate to a receptor at the "Rocks" due to noble mrem gases in an elevated release yr

                -       Skin dose rate to a receptor at the "Rocks" due to noble      mrem gases in a ground level release                                 yr
                -bt,)  Total body dose rate to an off-site receptor due to noble     mrem gases in an elevated release                                    yr
               -bas    Total body dose rate to an off-site receptor due to noble     mrem gases in a ground level release                                 yr DbbEe)          =       Total body dose rate to a receptor at the Science &           mrem Nature Center due to noble gases in an elevated release         yr blbEW,          =       Total body dose rate to a receptor at the Science &           mrem Nature Center due to noble gases in a ground level              yr release D)Re)           =       Total body dose rate to a receptor at the "Rocks" due to      mrem noble gases in an elevated release                              yr lbRW)           -       Total body dose rate to a receptor at the "Rocks" due to      mrem noble gases in a ground level release                            yr D/Q             =       Deposition factor for dry deposition of elemental                1 radioiodines and other particulates                              M2 ECL,           =       Effluent concentration limit (ECL) for radionuclide "i" (excluding dissolved and entrained noble gas) as               AiCilml specified in 10 CFR 20, Appendix B, Table 2.
              -       The fraction of the offsite limiting total body dose rate   Dimensionless administratively assigned to the plant vent release Fd             =       Actual or estimated flow rate out of discharge tunnel       gpm or ft3 tsec Fm           .         Flow rate past liquid waste test tank monitor                   gpm F=             -       Maximum allowable discharge flow rate from liquid               gpm test tanks based on all (beta - emitting and gamma -

emitting) radionuclides F-r = Maximum allowable discharge flow rate from the test gpm, tank past the monitor which would equate to the control concentration limit for the gamma radioactivity mixture determined to be in the test tank B.l-19 ODCM Rev. 25

TABLE B.1-8

SUMMARY

OF VARIABLES (Continued) Variable Definition Units fg = The fraction of the offsite limiting total body dose rate Dimensionless administratively assigned to monitored ground level release F Flow rate past plant vent monitor cc sec fglan d = Release reduction factor to be administratively assigned Dimensionless to account for potential unmonitored contributions from the Turbine Gland Seal Exhaust I f1;:f2; f3; = Fraction of total ECL associated with Paths 1,2, 3, Dimensionless and 4 I F,EFN,G = Total fraction of ECL in liquid pathways (excluding Dimensionless noble gases)

      ',         =   Maximum permissible concentration for radionuclide               pCi/cc I                     "i" ("old" 10 CFR 20, Appendix B, Table II, Column 2)

Qi Release to the environment for radionuclide "i" curies, or [L curies Qi Release rate to the environment for radionuclide "i" pCi/sec point Liquid monitor response for the limiting concentration jiCi/ml at the point of discharge in = Response of the noble gas monitor to limiting total body cpm, or pCi/sec dose rate Rtb Response of the noble gas monitor to limiting total body cpm, or pCi/sec dose rate SF Shielding factor Dimensionless Sg Detector counting efficiency from the gas monitor cpm mR/ hr

                                                                                     -- or calibration                                                pCi-cc         p Ci/ cc Sgi                Detector counting efficiency for noble gas "i"               cpm omR/ hr Pci-cc         PCi/Cc Si                 Detector counting efficiency from the liquid monitor           cps/pCi/ml calibration Sli             =  Detector counting efficiency for radionuclide "i"              cps/ACi/ml X/Q             =  Average long-term undepleted atmospheric dispersion                 sec factor (Tables B.74, B.7-5, and B.7-6)                              m3 B.1-20                             ODCM Rev. 25

I I TABLE B. 1-8

SUMMARY

OF VARIABLES (Continued) Variable Definition Units [X/Q]y - Effective long-term average gamma atmospheric sec dispersion factor (Tables B.7-4, B.7-5, and B.7-6) M3 SWF = Service Water System flow rate gph PCC = Primary component cooling water measured (decay ACi/ml corrected) gross radioactivity concentration t-a = Unitless factor which adjusts the value of atmospheric Dimensionless dispersion factors for elevated or ground-level releases with a total release duration of t hours B. 1-21 ODCM Rev. 25

TABLE B.1 -9 DEFINITION OF TERMS Critical Receptor - A hypothetical or real individual whose location and behavior cause him or her to receive a dose greater than aniy other possible real individual. Dose - As used in Regulatory Guide 1.109, the term "dose," when applied to individuals, is used instead of the more precise term "dose equivalent," as defined by the International Commission on Radiological Units and Measurements (ICRU). When applied to the evaluation of internal deposition or radioactivity, the term "dose," as used here, includes the prospective dose component arising from retention in the body beyond the period of environmental exposure, i.e., the dose commitment. The dose commitment is evaluated over a period of 50 years. The dose is measured in mrem to tissue or mrad to air. Dose Rate - The rate for a specific averaging time (i.e., exposure period) of dose accumulation. Liquid Radwaste Treatment System - The components or subsystems which comprise the available treatment system as shown in Figure B.6-1. n , . B.1-22 ODCM Rev. 25

TABLE B.1-10 DOSE FACTORS SPECIFIC FOR SEABROOK STATION FOR NOBLE GAS RELEASES Gamma Total Beta Skin Dose Combined Skin Dose Combined Skin Dose Beta Air Dose Gamma Air Dose Body Dose Factor Factor Factor for Elevated Factor for Ground Factor Factor Radio- Release Points Level Release Points nuclide mrem- m3 Imren- m3 DFi(e) (mre. se) nirem- sec DF irad C- y mrad- m3 DF'i(g)( Ci ) pCi -yr DF( i r) DFBi ( ) pCi- yr 4uCi- yr pu Ci- yr 1pCi- yr pCi- yr Ar-41 8.84E-03 2.69E-03 1.09E-02 6.20E-02 3.28E-03 9.30E-03 Kr-83m 7.56E-08 1.81E-05 7.28E-05 2.88E-04 1.93E-05 Kr-85m 1.17E-03 1.46E-03 2.35E-03 1.92E-02 1.97E-03 1.23E-03 Kr-85 1.61E-05 1.34E-03 1.1 E-03 1.35E-02 1.95E-03 1.72E-05 Kr-87 5.92E-03 9.73E-03 1.38E-02 1.21E-01 1.03E-02 6.17E-03 Kr-88 1.47E-02 2.37E-03 1.62E-02 8.1OE-02 2.93E-03 1.52E-02 Kr-89 1.66E-02 .OlE-02 2.45E-02 1.66E-01 1.06E-02 1.73E-02 Kr-90 1.56E-02 7.29E-03 2.13E-02 1.34E-01 7.83E-03 1.63E-02 Xe-131m 9.15E-05 4.76E-04 5.37E-04 5.35E-03 1.11 E-03 1.56E-04 Xe-133m 2.51 E-04 9.94E-04 1.12E-03 1.12E-02 1.48E-03 3.27E-04 Xe-133 2.94E-04 3.06E-04 5.83E-04 4.39E-03 1.05E-03 3.53E-04 Xe-135m 3.12E-03 7.1 IE-04 3.74E-03 1.98E-02 7.39E-04 3.36E-03 Xe-135 1.81E-03 1.86E-03 3.33E-03 2.58E-02 2.46E-03 1.92E-03 Xe-137 1.42E-03 1.22E-02 1.14E-02 1.28E-01 1.27E-02 1.51E-03 Xe-138 8.83E-03 4.13E-03 1.20E-02 7.60E-02 4.75E-03 - 9.21E-03 8.84E-03 = 8.84 x 10 3 B.1-23 ODCM Rev. 25 9

TABLE B.1-1 DOSE FACTORS SPECIFIC FOR SEABROOK STATION FOR

                          !    ,     LIOUID RELEASES ! ;

Total Body Maximum Organ

                                        -          Dose Factor                  Dose Factor mrem Radionuclide                            DFL-h (i)                          D 1 uLw(-;Ci pUCi H-3                                   3.02E-13                      3.02E-13 Na-24                                    1.38E-10                     1.42E-10 Cr-51                                  1.83E-II                      lA8E-09 Mn-54                                   5.15E-09                      2.68E-08 Fe-55                                   1.26E-08                     7.67E-08 Fe-59                                  8.74E-08                      6.66E-07 Co-58                                   2.46E-09                      1.40E-08 Co-60                                   6.15E-08                      9.22E-08 Zn-65                                   2.73E-07                      5.49E-07 Br-83                                  1.30E-14                      1.89E-14 Rb-86                                   4.18E-10                      6.96E-10 Sr-89                                  2.17E-10                      7.59E-09 Sr-90                                  3.22E-08                      1.31E-07 Nb-95                                   5.25E-10                      1.58E-06 Mo-99                                   3.72E-1 1                    2.67E-10 Tc-99m                                   5.22E-13                      l.95E-12 Ag-110m                                   1.01E-08                      6.40E-07 Sb-124                                  1.71E-09                      9.89E-09 Sb-125                                  6.28E-09                      8.31 E-09 Te-127m                                   7.07E-08                     .1.81E-06 Te-127                                   3.53E-10                      9.54E-08 Te-129m                                   1.54E-07                      3.46E-06 Te-129                                   7.02E-14                      1.05E-13 Te-131m                                   3.16E-08                     2.94E-06 Te-132                                   9.06E-08                     3.80E-06 1-130                                  2.75E-1 1                    3.17E-09 1-131                                  2.30E-10                      1.00E-07 1-132                                  6.28E-1 1                    6.36E-1 1 1-133                                  3.85E-1 I                     1.1513-08 I-134                                  1.19E-12                      1.41E-12 I-135                                  5.33E-1 I                    4.69E-10 Cs-134                                   3.24E-08                     3.5613-08 Cs-136                                   2.47E-09                     3.27E-09 Cs-137                                   3.58E-08                     4.03E-08 Ba-140                                   1.70E-10                     3.49E-09 La-140                                   1.07E-10                     4.14E-08 Ce-141                                   3.85E-1 1                    9.31 E-09 Ce-144                                   1.96E-10                     6A6E-08 Other*                                  3.12E-08*                     1.58E-06*

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

B.1 ODCM Rev. 25

I I TABLE B.I-12 DOSE AND DOSE RATE FACTORS SPECIFIC FOR SEABROOK STATION FOR IODINES. TRlTIUM AND PARTICULATE RELEASES Critical Organ Critical Organ Dose Critical Organ Dose Rate Critical Organ Dose Dose Factor Factor for Ground Factor for Elevated Rate Factor for Ground for Elevated Level Release Point Release Point Level Release Point Release Point mrem mrem mrem-sec mrem-sec Radio- DFGico.e) ( ) DFGiw(z) (C) DFG'ic(.) ( ) DFG'ic Ci) nuclide piPci yr-puCi H-3 3.08E-10 3.76E-09 9.71E-03 1.19E-01 Cr-51 8.28E-09 2.89E-08 2.91E-01 1.01E+00 Mn-54 1.1lE-06 3.79E-06 4.38E+01 1.50E+02 Fe-59 1.06E-06 3.65E-06 3.53E+01 1.21E+02 Co-58 5.56E-07 1.91E-06 2.OOE+01 6.88E+01 Co-60 1.21E-05 4.12E-05 5.42E+02 1.85E+03 Zn-65 2.33E-06 7.93E-06 7.82E+01 2.66E+02 Sr-89 1.98E-05 6.73E-05 6.24E+02 2.12E+03 Sr-90 7.21 E-04 2.47E-03 2.27E+04 7.79E+04 Zr-95 1.IOE-06 3.77E-06 3.63E+01 1.24E+02 Nb-95 2.01E-06 6.86E-06 6.40E+01 2.20E+02 Mo-99 1.63E-08 1.1OE-07 5.39E-01 3.56E+00 Ru-103 3.03E-06 1.04E-05 9.62E+01 3.31E+02 Ag-i 10m 5.02E-06 1.72E-05 1.80E+02 6.15E+02 Sb-124 1.83E-06 6.28E-06 6.15E+01 2.1 l E+02 1-131 1.47E-04 5.04E-04 4.64E+03 l.59E+04 1-133 1.45E-06 5.72E-06 4.57E+01 1.80E+02 Cs-134 5.62E-05 l.91E-04 1.81E+03 6.18E+03 Cs-137 5.47E-05 1.86E-04 1.79E+03 6.09E+03 Ba-140 l.55E-07 6.39E-07 5.01E+00 2.06E+01 Ce-141 2.65E-07 9.28E-07 8.45E+00 2.96E+01 Ce-144 6.09E-06 2.09E-05 1.93E+02 6.62E+02 Other* 4.09E-06 1.39E-05 1.29E+02 4.38E+02

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

B.1l-25 ODCM Rev. 25

TABLE B.1-13 COMBINED SKIN DOSE RATE FACTORS SPECIFIC FOR SEABROOK STATION SPECIAL RECEPTORS(') FOR NOBLE GAS RELEASE Science & Nature Science & Nature ,The "Rocks" The "Roc ks" Center Center Combined Skin Combined Skin Combined Skin Combined Skin Dose Rate Factor for Dose Rate Faictor for Dose Rate Factor for Dose Rate Factor for Elevated Release Point Ground Level Release Elevated Release Point Ground Level Point Release Point Radio. moremse mrem- sec mrem-sec , mreim-sec nuclidi DF' see) DF ( ) DF'iR(e) ( pCi-yr ) DF' iR(g) ( pC i- yr

                       ,Ci-vr                 pCi-yr p

Ar-41 1.57E-02 1.17E-01 9.73E-02 6.99E-01 Kr-83m 2.35E-05 1.13E-04 1.07E-04 5.57E-04 Kr-85m 3.84E-03 4.08E-02 3.16E-02 2.69E-01 Kr-85 2.16E-03 3.09E-02 2.29E-02 2.15E-01 Kr-87 2.3 1E-02 2.60E-01 2.OOE-01 1.73E+00 Kr-88 2.23E-02 1.44E-01 1.25E-01 8.18E-01 Kr-89 3.73E-02 3.34E-01 2.68E-01 2.12E+00 Kr-90 3.15E-02 2.64E-01 2.14E-01 1.64E+00 Xe-131m 9.52E-04 1.1 9E-02 8.96E-03 8.07E-02 Xe-133m l.99E-03 2.48E-02 1.87E-02 1.68E-01 Xe-133 9.20E-04 9.1 1E-03 7.16E-03 5.91E-02 Xe-135m 5.24E-03 3.61E-02 3.07E-02 2.1 lE-01 Xe-135 5.32E-03 5.41E-02 4.23E-02 3.53E-01 Xe-137 2.14E-02 2.89E-01 2.16E-01 2.OOE+00 Xe-138 1.78E-02 1.49E-01 1.21E-01 9.27E-01 (1) See Seabrook Station Technical Specification Figure 5.1-1. B.1-26' ODCM Rev. 25

I I TABLE B.1-14 DOSE AND DOSE RATE FACTORS SPECIFIC FOR THE SCIENCE & NATURE CENTER FOR IODINE, TRlTIUJM AND PARTICULATE RELEASES Critical Organ Dose Critical Organ Dose Critical Organ Dose Rate Critical Organ Dose Rate Factor for Elevated Factor for Ground Factor for Elevated Factor for Ground Level Release Point Level Release Release Point Release Point Point Radio- mrem mrenm DFG'.E() ( rem- sec inren- sec nuclide DFGi.o(e) ( -. ) DFQtcoEwg pCi- yr) DFG' )(m esc) p Ci-yr H-3 6.45E-1 I 9.27E-10 2.03E-03 2.92E-02 Cr-51 4.98E-09 2.88E-08 2.12E-01 1.1 E+00 Mn-54 1.39E-06 5.71E-06 6.24E+01 2.39E+02 Fe-59 3.09E-07 1.89E-06 1.29E+01 7.16E+0 1 Co-58 3.89E-07 2.IOE-06 1.72E+01 8.26E+01 Co-60 2.17E-05 8.03E-05 9.78E+02 3.63E+03 Zn-65 7.34E-07 3.19E-06 3.31 E+0I 1.33E+02 Sr-89 1.15E-07 1.61E-06 3.63E+00 5.08E+01 Sr-90 5.14E-06 7.19E-05 1.62E+02 2.27E+03 Zr-95 3.38E-07 2.57E-06 1.35E+01 9.15E+01 Nb-95 1.53E-07 9.35E-07 6.43E+00 3.53E+01 Mo-99 1.62E-08 1.92E-07 5.58E-01 6.21E+00 Ru-103 1.30E-07 8.64E-07 5.33E+00 3.19E+01 Ag-110in 3.43E-06 1.54E-05 1.55E+02 6.34E+02 Sb-124 6.96E-07 4.46E-06 2.89E+01 1.67E+02 1-131 7.79E-07 1.08E-05 2.47E+01 3.41E+02 I-133 1.84E-07 2.56E-06 5.83E+00 8.1 1E+01 Cs-134 6.83E-06 2.53E-05 3.08E+02 1.14E+03 Cs-137 1.03E-05 3.81E-05 4.64E+02 1.72E+03 Ba-140 1.14E-07 1.42E-06 3.85E+00 4.54E+01 Ce-141 4.09E-08 4.51E-07 1.45E+00 1.48E+01 Ce-144 6.95E-07 9.11 E-06 2.27E+01 2.90E+02 Other* 2.26E-06 9.24E-06 1.02E+02 3.91E+02

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

B.1-27 ODCM Rev. 25

TABLE B.1-15 DOSE AND DOSE RATE FACTORS SPECIEFIC FOR THE "ROCKS" FOR IODINE. TRITlUM. AND PARTICULATE RELEASES Critical Organ Dose Critical Organ Dose Critical Organ Dose Rate Critical Organ Dose Rate Factor Factor for Ground Factor for Elevated Factor for Ground Level for Elevated Release Level Release Point Release Point Point Release Point mrem mnrem mrem- sec Radio- DFGicoR(e) ( -. ) DFGiw) ( Ei ) DFGkoR~) ( Ci- yr DFG'iw (rem-sec ILo~g p Ci- yr/ ) nuclide PCi 4 uCi H-3 6.85E-10 6.45E-09 2.16E-02 2.03E-01 Cr-5I 2.68E-08 1.75E-07 1.07E+00 6.53E+00 Mn-54 5.84E-06 3.18E-05 2.55E+02 1.3 IE+03 Fe-59 1.74E-06 1.17E-05 6.78E+01 4.29E+02 Co-58 2.01E-06 1.25E-05 8.1 1E+O I 4.79E+02 Co-60 8.83E-05 4.09E-04 3.97E+03 1.85E+04 Zn-65 3.23E-06 1.80E-05 1.37E+02 7.29E+02 Sr-89 1.23E-06 1.15E-05 3.88E+01 3.63E+02 Sr-90 5.48E-05 5.14E-04 1.73E+03 1.62E+04 Zr-95 2.22E-06 1.68E-05 8.14E+01 5.83E+02 Nb-95 8.59E-07 5.79E-06 3.37E+01 2.13E+02 Mo-99 1.50E-07 1.34E-06 4.92E+00 4.32E+01 Ru-103 7.74E-07 5.47E-06 2.95E+01 1.96E+02 Ag- 1l0m 1.54E-05 8.77E-05 6.47E+02 3.53E+03 Sb-124 4.04E-06 2.80E-05 1.56E+02 1.01E+03 1-131 8.27E-06 7.73E-05 2.61E+02 2.44E+03 1-133 1.95E-06 1.83E-05 6.18E+01 5.77E+02 Cs-134 2.78E-05 1.29E-04 1.25E+03 5.80E+03 Cs-137 4.19E-05 1.94E-04 1.89E+03 8.77E+03 Ba-140 1.10E-06 9.99E-06 3.56E+01 3.19E+02 Ce-141 3.59E-07 3.14E-06 1.20E+01 1.02E+02 Ce-144 7.02E-06 6.46E-05 2.25E+02 2.05E+03 Other* 9.56E-06 5.09E-05 4.16E+02 2.12E+03

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

B.1-28 ODCM Rev. 25

2.0 METHOD TO CALCULATE OFF-SITE LIQUID CONCENTRATIONS Chapter 2 contains the basis for station procedures used to demonstrate compliance with ODCM Part A Control C.6.1 A, which limits the total fraction of ECL in liquid pathways, other than noble gases (denoted here as F ENG

                                   ) at the point of discharge from the station to the environment (see Figure B.6-1).

F is limited to less than or equal to one, i.e., F EVG 1. The total concentration of all dissolved and entrained noble gases at the point of discharge from the multiport diffuser from all station sources combined, denoted CN0 , is limited to 2E-04 pCi/mil, i.e., C < 2E-04 pCi/ml. Appendix C, Attachments 3 and 4, provide the option and bases for the use of the EMS determination of liquid concentration limits for plant discharges to the environment. 2.1 Method to Determine FENG AND CING First, determine the total fraction of ECL (excluding noble gases), at the point of discharge from the station from all significant liquid sources denoted F, NG; and then separately determine the total concentration at the point of discharge of all dissolved and entrained noble gases from all station sources, denoted ClG j as follows: FzrG = <10 (2-1) ECLl'i C pCi/mi pCilml) and: Cl C 2E-04 (2-2) (,uCi/ml) ([tCi/ml) (,uCi/ml) where:. IF. = Total fraction of ECL in liquids, excluding noble gases, at the point of discharge from the multiport difuser. B.2-1 ODCM Rev. 25

I I Cp; = Concentration at point of discharge from the multiport diffuser of radionuclide "i", except for dissolved and entrained noble gases, from all tanks and other significant sources, p, from which a discharge may be made (including the waste test tanks and any other significant source from which a discharge can be made). Cpi is determined by dividing the product of the measured radionuclide concentration in liquid waste test tanks, PCCW, steam generator blowdown, or other effluent streams times their discharge flow rate by the total available dilution water flow rate of circulating and service water at the time of release (Aci/m1). ECL, = Effluent concentration limit (ECL) for radionuclide "i" (except for dissolved and entrained noble gases) in igCi/ml as specified in 10 CFR 20, Appendix B, Table 2. See Appendix B for a list of ECL values. C1NG = Total concentration at point of discharge of all dissolved and entrained noble gases in liquids from all station sources (gCi/ml) CliNG = Concentration at point of discharge of dissolved and entrained noble gas "i" in liquids from all station sources (pCi/ml) 2.2 Method to Determine Radionuclide Concentration for Each Liquid Effluent Source 2.2.1 Waste Test Tanks Cpj is determined for each radionuclide detected from the activity in a representative grab sample of any of the waste test tanks and the predicted flow at the point of discharge. The batch releases are normally made from two 25,000-gallon capacity waste test tanks. These tanks normally hold liquid waste which may have been processed through the installed vendor equipment. The waste test tanks can also contain other waste such as liquid taken directly from the floor drain/chemical drain treatment tanks when that liquid does not require processing in the evaporator, from the installed vendor resin skid, distillate from the boron recovery evaporator when the BRS evaporator is substituting for the waste evaporator, or waste distillate from the Steam Generator Blowdown System when that system must discharge liquid off site. If testing indicates that purification of the waste test tank contents is required prior to release, the liquid can be circulated through the waste demineralizer and filter. The contents of the waste test tank may be reused in the Nuclear System if the sample test meets the purity requirements. Prior to discharge, each waste test tank is analyzed for principal gamma emitters in accordance with the liquid sample and analysis program outlined in Part A to the ODCM. B.2-2 ODCM Rev. 25

2.2.2 Turbine Building Sump The Turbine Building sump collects leakage from the Turbine Building floor drains and discharges the liquid unprocessed to the circulating water system. Sampling of this potential source is normally done once per week for determining the radioactivity released to the environment (see Table A.6.I-1). 2.2.3 Steam Generator Blowdown Flash Tank The primary method to process radioactive secondary liquid from the steam generators is to direct steam blowdown flash tank bottoms cooler discharge to the floor drain tanks. If no secondary pressure is available, the steam blowdown and wet lay-ups pumps can be used. From the floor drain tanks, processing through the installed vendor resin skid (WLSKD-135) to the waste test tanks is the preferred method. Othermethods maybe used as defined below. The steam generator blowdown evaporators may process the liquid from the steam generator blowdown flash tank when there is primary to secondary leakage. Distillate from the evaporators can be sent to the waste test tanks or recycled to the condensate system. When there is no primary to secondary leakage, flash tank liquid is processed through the steam generator blowdown demineralizers and returned to the secondary side. Steam generator blowdown is only subject to sampling and analysis when all or part of the blowdown liquid is being discharged to the environment instead of the nornal recycling process (see Table A.6.1-1). 2.2.4 PrimarM Component Cooling Water (PCCW) System The PCCW System is used to cool selected primary components. The system is normally sampled weekly to determine if there is any radwaste in-leakage. If leakage has been determined, the Service Water System is sampled to determine if any release to the environment has occurred. B.2-3 ODCM Rev. 25

3.0 OFF-SITE DOSE CALCULATION METHODS Chapter 3 provides the basis for station procedures required to meet the Radiological Effluent Control Program (RECP) dose and dose rate requirements contained in ODCM Part A Controls. A simple, conservative method (called Method I) is listed in Tables B.1-2 to B. 1-7 for each of the requirements of the RECP. Each of the Method I equations is presented in Part B, Sections 3.2 through 3.9. As an alternate to Method I,'the EMS computer program documented in Appendix C can be used to determine regulatory compliance for effluent doses and dose rates. The use of the EMS software is designated as Method IA in Chapter 3. In addition, those sections include more sophisticated methods (called Method B) for use when more refined results are needed; This chapter provides the methods, data, and reference material with which the operator can calculate the needed doses, dose rates and setpoints. For the requirements to demonstrate compliance with Part A off-site dose limits, the contribution from all measured ground level releases must be added to the calculated contribution from the vent stack to determine the Station's total radiological impact. The bases for the dose and dose rate equations are given in Chapter 7.0. Method IA bases and software verification documentation are contained in Appendix C. The Annual Radioactive Effluent Release Report, to be filed after January 1 each year per Technical Specification 6.8.1.4, and Part A, Section 10.2, requires that meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, be used for determining the gaseous pathway doses. For continuous release sources (i.e.,- plant vent, condenser air removal exhaust, and gland steam packing exhauster), concurrent quarterly average meteorology will be used in the dose calculations along with the quarterly total radioactivity released. For batch releases or identifiable operational activities (i.e., containment purge or venting to atmosphere of the Waste Gas System), concurrent meteorology during the period of release will be used to determine dose if the total noble gas or iodine and particulates released in the batch exceeds five percent of the total quarterly radioactivity released from the unit; otherwise quarterly average meteorology will be applied. Quarterly average meteorology will also be applied to batch releases if the hourly met data for the period of batch release is unavailable. Annual dose assessment reports prepared in accordance with the requirements of the ODCM will include a statement indicating that the appropriate portions of Regulatory Guide 1.109 (as identified in the individual subsections of the ODCM for each class of effluent exposure) have been used to determine dose impact from station releases. Any deviation from the methodology, assumptions, or parameters given in Regulatory Guide 1.109, and not already identified in the bases of the ODCM, will be explicitly described in the effluent report, along with the bases for the deviation. B.3-1. ODCM Rev. 25

II 3.1 Introductory Concepts In Part A Controls, the RECP limits for dose or dose rate are stated. The term "dose" for ingested or inhaled radioactivity means the dose comrnmitment, measured in mrem, which results from the exposure to radioactive materials that, because of uptake and deposition in the body, will continue to expose the body to radiation for some period of time after the source of radioactivity is stopped. The time frame over which the dose commitment is evaluated is 50 years. The phrases "annual dose" or "dose in one year" then refers to the 50-year dose commitment resulting from exposure to one year's worth of releases. "Dose in a quarter" similarly means the 50-year dose commitment resulting from exposure to one quarter's releases. The term "dose," with respect to external exposures, such as to noble gas clouds, refers only to the doses received during the actual time period of exposure to the radioactivity released from the plant. Once the source of the radioactivity is removed, there is no longer any additional accumulation to the dose commitment.

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

In addition to limits on dose commitment, gaseous effluents from the station are also controlled so that the maximum or peak dose rates at the site boundary at any time are limited to the equivalent annual dose limits of 10 CFR Part 20 to unrestricted areas (if it were assumed that the peak dose rates continued for one year). These dose rate limits provide reasonable assurance that members of the public, either inside or outside the site boundary, will not be exposed to annual averaged concentrations exceeding the limits specified in Appendix B, Table 2 of 0 10 CFR Part 20. See Appendix B for a listing of these concentration limits. The quantities AD and D are introduced to provide calculable quantities, related to off-site doses or dose rates that demonstrate compliance with the RETS. Delta D, denoted AD, is the quantity calculated by the Part B, Chapter 3, Method I dose equations. It represents the conservative increment in dose. The AD calculated by Method I equations is not necessarily the actual dose received by a real individual, but usually provides an upper bound for a given release because of the conservative margin built into the dose factors and the selection and definition of critical receptors. The radionuclide specific dose factors in each Method I dose equation represent the greatest dose to any organ of any age group. (Organ dose is a function of age because organ mass and intake are functions of age.) The critical receptor assumed by "Method I" equations is then generally a hypothetical individual whose behavior - in terms of location and intake - results in a dose which is higher than any real individual is likely to receive. Method IA dose calculations using the EMS software evaluate each age group and organ combination to determine the maximum organ dose for each mix of radionuclides specified in a release period. Method II also allows for a more exact dose calculation for each individual if necessary. B.3-2 ODCM Rev. 25

D dot, denoted D, is the quantity calculated in the Part B, Chapter 3 dose rate equations. It is calculated using the station's effluent monitoring system reading and an annual or long-term average atmospheric dispersion factor. D predicts the maximum off-site annual dose if the peak observed radioactivity release rate from the plant stack continued for one entire year. Since peak release rates, or resulting dose rates, are usually of short time duration on the order of an hour or less, this approach then provides assurance that 10 CER 20.106 limits will be met. Each of the methods to calculate dose or dose rate is presented in the following subsections. Each dose type has two levels of complexity. Method I is the simplest and contains many conservative factors. As an alternate to Method I the EMS computer program documented in Appendix C cani be used to determine regulatory compliance for effluent doses and dose rates. The use of the EMS system is designated as Method IA in Chapter 3 of Part B. Method II is a more realistic analysis which makes use of the models in Regulatory Guide 1. 109 (Revision 1), as noted in each subsection of Part B, Chapter 3 for the various exposure types. A detailed description of the methodology,- assumptions, and input parameters to the dose models that are applied in each Method II calculation, if not already explicitly described in the ODCM, shall be documented and provided when this option is used for NRC reporting and ODCM, Part A RECP dose compliance. B.3-3 ODCM Rev. 25

l L 3.2 Method to Calculate the Total Body Dose from Liquid Releases Part A Control C.6.2.1 limits the total body dose commitment to a member of the public from radioactive material in liquid effluents to 1.5 mrem per quarter and 3 mrem per year per unit. Part A Control C.6.3.1 requires liquid radwaste treatment when the total body dose estimate exceeds 0.06 mrem in any 31 -day period. Part A Control C.8. 1.1 limits the total body dose commitment to any real member of the public from all station sources (including liquids) to 25 mrem in a year. Use Method I or Method IA first to calculate the maximum total body dose from a liquid release from the station as it is simpler to execute and more conservative than Method I. Use Method II if a more refined calculation of total body dose is needed, i.e., Method I or Method IA indicates the dose might be greater than Part A Control limits. To evaluate the total body dose, use Equation 3-1 to estimate the dose from the planned release and add this to the total body dose accumulated from prior releases during the month. See Part B, Section 7.1.1 for basis. 3.2.1 Method I The total body dose from a liquid release is: Dth = kZ Qi DFLib (3-1) (mrem) = ( ) Ci) where DFL", = Site-specific total body dose factor (mrem/QCi) for a liquid release. It is the highest of the four age groups. See Table B.1-11. Qi = Total activity (pCi) released for radionuclide "i". (For strontiums, use the most recent measurement available.) k = 918/Fd; where Fd is the average (typically monthly average) dilution flow of the Circulating Water System at the point of discharge from the multiport diffuser (in &/sec). For normal operations with a cooling water flow of 918 ft3 /sec, k is equal to 1. During periods when no or low flow is recorded from the Discharge Transition Structure (DTS), a minimum dilution flow of 23 ft3 /sec (10,500 gpm for one service water pump) can be used since this would be the minimum flow available when discharges to the tunnel are reestablished. Alternately, the monthly average discharge flow for the period in which the release occurs can be used when this value is available. B.3-4 ODCM Rev. 25

3.2 Method to Calculate the Total Body Dose from Liquid Releases 3.2.1 Method I (Continued). Equation 3-1 can be applied under the following conditions (otherwise, justify Method I or consider Method II):

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

2. Any continuous or batch release over any time period up to 1 year. For annual dose estimates, the annual average discharge flow from the DTS should be used as the dilution flow estimate. - -

Method IA is implemented by the EMS software as described in Appendix C. Liquid release models are detailed in sections 2.1 - 2.6 of the EMS Technical Reference Manual (Attachment 4 of Appendix C). 3.2.2 Method II Method II consists of the models, input data and assumptions (bioaccumulation factors, shore-width factor, dose conversion factors, and transport and buildup times) in Regulatory Guide 1.109, Rev. I (Reference A), except where site-specific data or assumptions have been identified in the ODCM. The general equations (A-3 and A-7) taken from Regulatory Guide 1.109, and used in the derivation of the simplified Method I approach as described in the Bases section, are also applied to Method II assessments, except that doses calculated to the whole body from radioactive effluents are evaluated for each of the four age groups to determine the maximum whole body dose of an age-dependent individual via all existing exposure pathways. Table B.7-1 lists the usage factors of Method II calculations. As noted in Section B.7.1, the mixing ratio associated with the edge of the I IF surface isotherm above the multiport diffuser may be used in Method II calculations for the shoreline exposure pathway (Mp = 0.025). Aquatic food ingestion pathways shall limit credit taken for mixing zone dilution to the same value assumed in Method I (MP= 0.10). B.3-5 ODCM Rev. 25

l L 3.3 Method to Calculate Maximum Organ Dose from Liquid Releases Part A Control C.6.2.1 limits the maximum organ dose commitment to a Member of the Public from radioactive material in liquid effluents to 5 mrem per quarter and 10 mrem per year per unit. Part A Control C.6.3.1 requires liquid radwaste treatment when the maximum organ dose projected exceeds 0.2 mrem in any 31 days (see Part B, Subsection 3.11 for dose projections). Part A Control C.8.1.1 limits the maximum organ dose commitment to any real member of the public from all station sources (including liquids) to 25 mrem in a year except for the thyroid, which is limited to 75 mrem in a year. Use Method I or Method IA first to calculate the maximum organ dose from a liquid release to unrestricted areas (see Figure B.6-1) as it is simpler to execute and more conservative than Method II. Use Method II if a more refined calculation of organ dose is needed, i.e., Method I or Method IA indicates the dose may be greater than the limit. Use Equation 3-2 to estimate the maximum organ dose from individual or combined liquid releases. See Part B, Section 7.1.2 for basis. 3.3.1 Method I The maximum organ dose from a liquid release is: D,= k Qj DFL, (3-2) (mrem) =() Ci) F where DFLj, = Site-specific maximum organ dose factor (mrem/pCi) for a liquid release. It is the highest of the four age groups. See Table B.1-11 . Q = Total activity (pQCi) released for radionuclide "i". (For composited analyses of strontiums, use the most recent measurement available.) k = 918/Fd; where Fd is the average (typically monthly average) dilution flow of the Circulating Water System at the point of discharge from the multiport diffuser (in ft3 /sec). For normal operations with a cooling water flow of 918 ft 3 /sec, k is equal to l. During periods when no or low flow is recorded from the Discharge Transition Structure (DTS), a minimum dilution flow of 23 ft3 /sec (10,500 gpm for one service water pump) can be used since this would be the minimum flow available when discharges to the tunnel are reestablished. Alternately, the monthly average discharge flow for the period in which the release occurs can be used when this value is available. B.3-6 ODCM Rev. 25

3.3 Method to Calculate Maximum Organ Dose from Liquid Releases 3.3.1 Method I (Continued) Equation 3-2 can be applied under the following conditions (otherwise, justify Method I or consider Method A):

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

2. -Anycontinuous or batch release over any time period up to 1 year. For annual dose estimates, the annual average discharge flow from the DTS should be used as the dilution flow estimate.

Method IA is implemented by the EMS software as described in Appendix C. Liquid release models are detailed in sections 2.1 - 2.6 of the EMS Technical Reference Manual (Attachment 4 of Appendix C). 3.3.2 Method II Method II consists of the models, input data and assumptions (bioaccumulation factors, shore-width factor, dose conversion factors, 'and transport and buildup times) in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific data or assumptions have been identified in the ODCM. The general equations (A-3 and A-7) taken from Regulatory Guide 1.109, and used in the derivation of the simplified Method I approach as described in the Bases section, are also applied to Method HI assessments, except that doses calculated to critical organs from radioactive effluents are evaluated for each of the four age groups to determine the maximum critical organ of an age-dependent individual via all existing exposure pathways. Table B.7-1 lists the usage factors for Method II calculations. As noted in Section B.7.1, the mixing ratio associated with the edge of the 1PF surface isotherm above the multiport diffuser may be used in Method II calculations for the shoreline exposure pathway (Mp = 0.025). Aquatic food ingestion pathways shall limit credit taken for mixing zone dilution to the same value assumed in Method I (Mp = 0.10). 7. B.3-7 ODCM Rev. 25

I I 3.4 Method to Calculate the Total Body Dose Rate from Noble Gases Part A Control C.7.1.1 limits the dose rate at any time to the total body from noble gases at any location at or beyond the site boundary to 500 mrem/year. The Part A Control indirectly limits peak release rates by limiting the dose rate that is predicted from continued release at the peak rate. By limiting Db to a rate equivalent to no more than 500 mrem/year, we assure that the total body dose accrued in any one year by any member of the general public is less than 500 mrem. Use Method I or Method IA first to calculate the Total Body Dose Rate from the peak release rate via the station vents or ground level effluent release points. Method I applies at all release rates. Use Method II if a more refined calculation of Dd, is desired by the station (i.e., use of actual release point parameters with annual or actual meteorology to obtain release-specific XIQs) or if Method I or Method IA predicts a dose rate greater than the Part A Control limit to determine if it had actually been exceeded during a short time interval. See Part B, Section 7.2.1 for basis. Compliance with the dose rate limits for noble gases are continuously demonstrated when effluent release rates are below the plant vent noble gas activity monitor alarm setpoint by virtue of the fact that the alarm setpoint is based on a value which corresponds to the off-site dose rate limit, or a value below it. Determinations of dose rate for compliance with Part A Control are performed when the effluent monitor alarm setpoint is exceeded, or as required by the Action Statement (Part A Control C.5.2, Table A.5.2-1) when the monitor is inoperable. 3.4.1 Method I The Total Body Dose Rate to an off-site receptor due to noble gases in effluents released via the plant vent can be determined as follows: D d,(e) = 0.85 * (0 W

• DFBI) (3-3a) mrem (pCi - sec) (pCi) (mren -m3 yr .pci-m) 3sec pCi-yr )

where D = The off-site total body dose rate (mrem/yr) due to noble gases in elevated effluent tb(e) releases, Q = the release rate at the station vents (IiCi/sec), for each noble gas radionuclide, "i", shown in Table B.1-10, and DFB; = total body gamma dose factor (see Table B.l-10). The Total Body Dose Rate (to an off-site receptor) due to noble gas in ground level effluent releases can be determined as follows: B.3-8 ODCM Rev. 25

3.4 Method to Calculate the Total Body Dose Rate from Noble Gases 3.4.1 Method I (Continued) D =beg) 3.4 * (j *DFBi) (3-3b) mrrnre PO -sec ) _ rnrem-m3 yr pjCim3J t sec) pCi-yr ) where Dtb~g) = The total off-site body dose rate (mnrem/yr) due to noble gases in ground level equivalent effluent releases, and Q and DFBi are as defined for Equation 3-3a. For the special on-site receptor locations, the Science & Nature Center and the "Rocks," the total body dose rates due to noble gases in effluent discharges can be determined as follows: For the Science & Nature Center, elevated effluent release: DbE(.) = 0.0015 * (Qj *DFBi) (3-3c) i For the Science & Nature Center, ground level effluent release: DzbEw = 0.0074 * (* O

• DFBi) (3-3d)

For the "Rocks," elevated effluent release: DbRWe) = 0.038

• E
• DFBi) (3-3e)

For the "Rocks," ground level effluent release: Dp = 0.2 * (Q* DFBi) (3-3f) where DeE(e)v DcbE~g)9 DdtR(e)' and bR = The total body dose rate (mrem/yr) at the Science & Nature Center and the "Rocks," respectively, due to noble gases in gaseous discharges from elevated (e) and ground level (g) release points, and Q; and DFBI are as defined previously. B.3-9 ODCM Rev. 25

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

1. Normal operations (nonemergency event), and

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

Method IA is implemented by the EMS software as described in Appendix C. Gaseous release models are detailed in Section 6.7.3 of the EMS Software Requirements Specification (Attachment 3 of Appendix C). 3.4.2 Method II Method II consists of the model and input data (whole body dose factors) in Regulatory Guide 1.109, Rev. I (Reference A), except where site-specific data or assumptions have been identified in the ODCM. The general equation (B-8) taken from Regulatory Guide 1.109, and used in the derivation of the simplified Method I approach as described in the Bases section, is also applied to a Method II assessment. No credit for a shielding factor (SF) associated with residential structures is assumed. Concurrent meteorology with the release period may be utilized for the gamma atmospheric dispersion factor identified in ODCM Equation 7-3 (Part B, Section 7.2.1), and determined as indicated in Part B, Section 7.3.2 for the release point (either ground level or vent stack) from which recorded effluents have been discharged. B.3-10 ODCM Rev. 25

3.5 METHOD TO CALCULATE THE SKIN DOSE RATE FROM NOBLE GASES Part A Control C.7.1.1 limits the dose rate at any time to the skin from noble gases at any location at or beyond the site boundary to 3,000 mrem/year. The Pait A Control indirectly limits peak release rates by limiting the dose rate that is predicted from continued release at the peak rate. By limiting ])ski, to a rate equivalent to no more than 3,000 mrem/year, we assure that the skin dose accrued in any one year by any member of the general public is less than 3,000 mrem. Since it can be expected that the peak release rate on which f),k, j1s derived would not be exceeded without corrective action being taken to lower it, the resultant average release rate over the year is expected to be considerably less than the peak release rate. Use Method I or Method IA first to calculate the Skin Dose Rate from peak release rate via station vents. Method I applies at all release rates. Use Method II if a more refined calculation of Dkil is desired by the station (i.e., use of actual release point parameters with afnnual or actual meteorology to obtain release-specific X/Qs) or if Method I or Method IA predicts a dose rate greater than the Part A Control limit to determine if it had actually been exceeded during a short time interval. See Part B, Section 7.2.2 for basis. Compliance with the dose rate limits for noble gases are continuously demonstrated when effluent release rates are below the plant vent noble gas activity monitor alarm setpoint by virtue of the fact that the alarm setpoint is based on a value which corresponds to the off-site dose rate limit, or a value below it. Determinations of dose rate for compliance with Part A Controls are performed when the effluent monitor alarm setpoint is exceeded. 3.5.1 Method I For an off-site receptor and elevated effluent release, the Skin Dose Rate due to noble gases is: DsdDe) = E (

• DFK'.) ) (34a)

Inrem = (PCi (mrem-sec) yr sec ) uCi - yr where Dae) = the off-site skin dose rate (mre/yr) due to noble gases in an effluent discharge from an elevated release point, Qa = as defined previously, and DF;C) = the combined skin dose factor for elevated discharges (see Table B.1-I0). B.3 ODCM Rev. 25

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

                            =Di;;o                                                                      (3-4b) where D 3 sk)       =    The off-site skin dose rate (mrem/yr) due to noble gases in an effluent discharge from a ground level release point,
                      =    as defined previously, and DFjO          = The combined skin dose factor for ground level discharges (see Table B.1-10).

For an on-site receptor at the Science & Nature Center and elevated release conditions, the skin dose rate due to noble gases is: D.ciEF(,) =0.0014* (Qi

• DFj,)) (34c) where DsId ,~e) = The skin dose rate (mrem/yr) at the Science & Nature Center due to noble gases in an elevated release,

                      =   as defined previously, and DF6,C)         =   the combined skin dose factor for elevated discharges (see Table B.1-13).

For an on-site receptor at the Science & Nature Center and ground level release conditions, the skin dose rate due to noble gases is: DE,,9E=0.0014*E(Qi*DF;W) (34d) where

                 = the skin dose rate (mrem/yr) at the Science & Nature Center due to noble gases in a ground level release,
                =  as defined previously, and DF,)      = The combined skin dose factor for ground level discharges (see Table B.1-13).

B.3-12 ODCM Rev. 25

For an on-site receptor at the "Rocks" and elevated release conditions, the skin dose rate due to noble gases is: DskinR(e) =0.0076 E (Qi

• DFj(,)) (3-4e) where DskinR(e) = the skin dose rate at the "Rocks" due to noble gases in an elevated release, QI = as defined previously, and DFRC) = The combined skin dose factor for elevated discharges (see Table B.I-13).

For an on-site receptor at the "Rocks" and ground level release conditions, the skin dose rate due to noble gases is: DskinrJ) = 0.0076

• E (Q;
• DFW) (3-4f) where

                 =t4   the skin dose rate (rnrem/yr) at the "Rocks" due to noble gases in a ground level release,
                =       as defined previously, and DFj(z) =         the combined skin dose factor for ground level discharges (see Table B.1-13).

Equations 3-4a through 34f can be applied under the following conditions (otherwise, justify Method I or consider Method II).

1. Normal operations (nonemergency event), and

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

Method IA is implemented by the EMS software as described in Appendix C. Gaseous release models are detailed in Section 6.7.3 of the EMS Software Requirements Specification (Attachment 3 of Appendix C). B.34-1 ODCM Rev. 25

I I 3.5.2 Method II Method II consists of the model and input data (skin dose factors) in Regulatory Guide 1. 109, Rev. I (Reference A), except where site-specific data or assumptions have been identified in the ODCM. The general equation (B-9) taken from Regulatory Guide 1.109, and used in the derivation of the simplified Method I approach as described in the Bases section, is also applied to a Method II assessment, no credit for a shielding factor (SF) associated with residential structures is assumed. Concurrent meteorology with the release period may be utilized for the gamma atmospheric dispersion factor and undepleted atmospheric dispersion factor identified in ODCM Equation 7-8 (Part B, Section 7.2.2), and determined as indicted in Part B, Sections 7.3.2 and 7.3.3 for the release point (either ground level or vent stack) from which recorded effluents have been discharged. B.3-14 ODCM Rev. 25

3.6 Method to Calculate the Critical Organ Dose Rate from Iodines, Tritium and Particulates with Tin Greater Than 8 Days Part A Control C.7.1.1 limits the dose rate at any time to any organ from 131I, 133i, 3 H and radionuclides in particulate form with half lives greater than 8 days to 1500 mremlyear to any organ. The Part A Control indirectly limits peak release rates by limiting the dose rate that is predicted from continued release at the peak rate. By limiting Dlo to a rate equivalent to no more than 1500 mremlyear, we assure that the critical organ dose accrued in any one year by any member of the general public is less than 1500 rmrem. Use Method I or Method IA first to calculate the Critical Organ Dose Rate from the peak release rate via the station vents. Method I applies at all release rates. Use Method II if a more refined calculation of f). is desired by the station (i.e., use of actual release point parameters with annual or actual meteorology to obtain release-specific X/Qs) or if Method I or Method IA predicts a dose rate greater than the Part A Control limit to determine if it had actually been exceeded during a short time interval. See Part B, Section 7.2.3 for basis. 3.6.1 Method I The Critical Organ Dose Rate to an off-site receptor and elevated release conditions can be determined as follows: Dco(e) (Q *DFG') (3-5a) (mremj = (iCij * (mrem-secj yr ) Sec J ui-yr where D<.) = The off-site critical organ dose rate (mrem/yr) due to iodine, tritium, and particulates in an elevated release,

                =      the activity release rate at the station vents of radionuclide "i" in JiCi/sec (i.e.,

total activity measured of radionuclide "i" averaged over the time period for which the filter/charcoal sample collector was in the effluent stream. For i = Sr89 or Sr90, use the best estimates, such as most recent measurements), and DFG ico(e) = the site-specific critical organ dose rate factor (- s )for an elevated gaseous release (See Table B.1-12). B.3-15 ODCM Rev. 25

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

             =     (QDFG')                                                                             (3-5b) ic~g where Dd
                 =       the off-site critical organ dose rate (mrem/yr) due to iodine, tritium, and particulates in a ground level release, Qj         =       as defined previously, and DFG'-,s =           the site-specific critical organ dose rate factor for a ground level gaseous discharge (see Table B.1-12).

For an on-site receptor at the Science & Nature Center and elevated release conditions, the critical organ dose rate can be determined as follows: DC.E() = 0.0014 * (d DFG'.ce)) _ (3-5c) i. where DCOE() = The critical organ dose rate (mrem/yr) to a receptor at the Science & Nature Center due to iodine, tritium, and particulates in an elevated release,

                 =      as defined previously, and DFG icOE(e)          =    the Science & Nature Center-specific critical organ dose rate factor for an elevated discharge (see Table B.1-14).

B.3-16 ODCM Rev. 25

For an on-site receptor at the Science & Nature Center and ground level release conditions, the critical organ dose rate is: f.EU)= 0.0014* (Q,*DFG'C) (3-Sd) where -

                     =    the critical organ dose rate (mrem/yr) to a receptor at the Science & Nature Center due to iodine, tritium, and particulates in a ground level release, Q.          =    as defined previously, and DFG'i       z            the Science & Nature Center-specific critical organ dose rate factor for a ground level discharge (see Table B.1-14).

For an on-site receptor at the "Rocks" and elevated release conditions, the critical organ dose rate is: DCOp~e) = 0.0076 *; (Q,* DFG(C,)) (3-Se) where DCOR(e) = The critical organ dose rate (mrem/yr) to a receptor at the "Rocks" due to iodine, tritium, and particulates in an elevated release,

                     =    as defined previously, and DFG'icoR(.)        =     the "Rocks"-specific critical organ dose rate factor for an elevated discharge (see Table B.l-15).

For an on-site receptor at the "Rocks" and ground level release conditions, the critical organ dose rate is: DccRW= 0.0076

• Ej (
• DFG'R) (3-5f where DCOR and Q = are as defined previously, and DFG'iRw = the "Rocks"-specific critical organ dose rate factor for a ground level discharge (see Table B.l-lS).

B.3-17 ODCM Rev. 25

II Equations 3-5a through 3-Sf can be applied under the following conditions (otherwise, justify Method I or consider Method II):

1. Normal operations (not emergency event), and

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

Method IA is implemented by the EMS software as described in Appendix C. Gaseous release models are detailed in Section 6.7.3 of the EMS Software Requirements Specification (Attachment 3 of Appendix C). 3.6.2 Method I[ Method II consists of the models, input data and assumptions in Appendix C of Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific data or assumptions have been identified in the ODCM (see Tables B.7-2 and B.7-3). The critical organ dose rate will be determined based on the location (site boundary, nearest resident, or farm) of receptor pathways as identified in the most recent annual land use census, or by conservatively assuming the existence of all pathways (ground plane, inhalation, ingestion of stored and leafy vegetables, milk, and meat) at an off-site location of maximum potential dose. Concurrent meteorology with the release period may be utilized for determination of atmospheric dispersion factors in accordance with Part B, Sections 7.3.2 and 7.3.3 for the release point (either ground level or vent stack) from which recorded effluents have been discharged. The maximum critical organ dose rates will consider the four age groups independently, and take no credit for a shielding factor (SF) associated with residential structures. B.3-18 ODCM Rev. 25

3.7 Method to Calculate the Gamma Air Dose from Noble Gases Part A Control C.7.2.1 limits the gamma dose to air from noble gases at any location at or beyond the site boundary to 5 mrad in any quarter and 10 mriad in any year per unit. Dose evaluation is required at least once per 31 days. Use Method I or Method IA first to calculate the gamma air dose from the station gaseous effluent releases during the period. Use Method II if a more refined calculation is needed (i.e., use of actual release point parameter with annual or actual meteorology to obtain release-specific X/Qs), or if Method I or Method IA predicts a dose greater than the Part A Control limit to determine if it had actually been exceeded. See Part B, Section 7.2.4 for basis. 3.7.1 Method I The general form of the gamma air dose equation is: Dzir = 3.17E-02 * [>QI *t * (Qi DFr) (3-6) (mrad) = (I3ii sec) * (secj * ( ) (PCi) mC ra ) where Drair is the gamma air dose. 3.17E-02 is the number of pCi per piCi divided by the number of second per year, [X/Q]' is the l-hour gamma atmospheric dispersion factor, t-a is a unitless factor which adjusts the 1-hour [X/Q]T value for a release with a total duration of t hours, Q, is the total activity in 1iCi of each radionuclide "i" released to the atmosphere from the station gaseous effluent release point during the period of interest, and DFr is the gamma dose factor to air for radionuclide "i" (see Table B.1-10). Incorporating receptor location-specific atmospheric dispersion'factors ([X/Q]7), adjustment factors (t4 ) for elevated and ground-level effluent release conditions, and occupancy factors when applicable (see Section 7.2.7), yields a series of equations by which the gamma air dose can be determined. B.3-19. ODCM Rev. 25

                                                                                            -- - II

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

DYjK¢) = 3.2E-07 t-t 02 7 5

• Z (Qj* DFj) (3-6a)

(mrad) =(pci-

                                  ,U~i

( (Ci) M3pCi (YrCimrad

                                                                        -m
                                                                        -yr)

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

Da) = 1.6E-06

• t4293 * (Qj*DFr) (3-6b)

               =a      pCi-yr'(          *(                  Ci(mrad-m      3 aCi     - m'                                  pCi - yr

c. Science & Nature Center receptor; elevated release conditions:

Dair') = 4.9E- I0

• t-°-25 *
• DF0)

                                                   *(Q                               (3-6c)

(mrad) PCO yr)*( )E(#Ci* m m3

                 /iCi-rni                               pCi- yr

d. Science & Nature Center receptor, ground-level release conditions:

Drft)=4.4 E- 09

• t-0.32 * (Q;
• DF(Y) (3-6d)

( pCi- yr mrad- m 3 (mrad) = ( )*( )MUci* ~) p1/Ci_-Mn pCi- yr

e. Receptor at the "Rocks"; elevated release conditions:

Dr.) = 5.1 E- 09

• t' *(Qj
• DFi) (3-6e)

_pCi- yr morad- m3 (mrad)=( ,C ,3)*( )E(,pCi* )

                ,pCi-r                                  pCi-yr B.3-20                  ODCM Rev. 25

f. Receptor at the "Rocks"; ground-level release conditions:

DarRe) =4.1 E- 08

• t"-O'* (Qi
• DF`) (3-6f)

                                        'i 3

(ma) PCi-yr (mrad( Ci-rn 3 )i mrad-m p)Zi-yr Equations 3-6a through 3-6f can be applied under the following conditions (otherwise justify Method I or consider Method II):

1. Normal operations (nonernergency event), and

2. Noble gas releases via station vents to the atmosphere.

Method IA is implemented by the EMS software as described in Appendix C. Gaseous release models are detailed in Section 6.7.3 of the EMS Software Requirements Specification (Attachment 3 of Appendix C). 3.7.2 Method II Method II consists of the models, input data (dose factors) and assumptions in Regulatory Guide 1.109, Rev. 1 (ReferenceA), except where site-specific data or assumptions have been identified in the ODCM. The general equations (B4 and B-5) taken from Regulatory Guide 1.109, and used in the derivation of the simplified Method I approach as described in the Part B Bases Section 7.2.4 are also applied to Method II assessments. Concurrent meteorology with the release period may be utilized for the gamma atmospheric dispersion factor identified in ODCM Equation 7-14, and determined as indicated in Part B, Section 7.3.2 for the release point (either ground level or vent stack) from which recorded effluents have been discharged.

                                                .      .1.

B.3-21 ODCM Rev. 25

I L 3.8 Method to Calculate the Beta Air Dose from Noble Gases Part A Control C.7.2.1 limits the beta dose to air from noble gases at any location at or beyond the site boundary to 10 mrad in any quarter and 20 mrad in any year per unit. Dose evaluation is required at least once per 31 days. Use Method I or Method IA first to calculate the beta air dose from gaseous effluent releases during the period. Method I applies at all dose levels. Use Method II if a more refined calculation is needed (i.e., use of actual release point parameters with annual or actual meteorology to obtain release-specific X/Qs) or if Method I or Method IA predicts a dose greater than the Part A Control limit to determine if it had actually been exceeded. See Part B, Section 7.2.5 for basis. 3.8.1 Method I The general form of the beta air dose equation is: D°,, = 3.17 E- 02 *(X/Q)1 M

• t-a
• E(Qi
• DFP') (3-7) mrad) IjpCi- yr*(

Ci- 3 in3*

                                             *(    *      ~*PET yr where D° is the beta air dose, 3.1 7E-02 is the number of pCi per liCi divided by the number of seconds per year, (X/Q),h is the 1-hour undepleted atmospheric dispersion factor, t-a is a unitless factor which adjusts the 1-hour X/Q value for a release with a total duration of t hours, Q2 is the total activity (gCi) of each radionuclide "i" released to the atmosphere during the period of interest, and DFf is the beta dose factor to air for radionuclide "i" (see Table B.1-10).

Incorporating receptor location-specific atmospheric dispersion factor (X/Q), adjustment factors (t')for elevated and ground-level effluent release conditions, and occupancy factors when applicable (see Section 7.2.7) yields a series of equations by which the Beta Air Dose can be determined. B.3-22 ODCM Rev. 25

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

Dap(e) 4.1E-7

• t* Z*(Q* DFr°) (3-7a) mad (pCi-(mrad)= Ci m' )

(pCi-)yr3 * ( ) E (PCi

• mrad-P ci-pMi-yr

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

Da) = 6.0 E-06 *t-03 ' 9 *Z*(Qj*DFP) (3-7b) p i- .r -r-d., (mrad) PCi-Yr)*( )X(CiUp*mim) p m-n pCi- yr

c. Science & Nature Center receptor; elevated release conditions:

Dp 1.8 E- 09

• t05 * (Q DFR ) (3-7c)

(mrad) = (pCi- yr3 )*( )1(PCi

• mrad-m 3 )

p Ci-rM pCi- yr

d. Science & Nature Center receptor; ground-level release conditions:

a;5= 2.4 E- 08

• t0347
• Q;
• DFf) (3-7d)

(mrad) ( T )E(Pci* pCi- yr

e. Receptor at the "Rocks"; elevated release conditions:

D!ROe) = 3.9E-08

• t024 9 *(Q
• DFIP) (3-7e)

(mrad)=(PC 3i_)*( 3 p~irn

                                )7(/.Ci*         yra) pCi- yr,

f. Receptor at the "Rocks"; ground-level release conditions:

D = 4.6 E-07

• t" 2 6 **E(Q;*DFr) (3-7f)

(rnrad)= ( pCi-r3 )*( )y(rCi rad*m3 (m) P Ci-rn pCi- yr B.3-23 ODCM Rev. 25

II Equations 3-7a through 3-7f can be applied under the following conditions (otherwisejustify Method I or consider Method II):

1. Normal operations (nonemergency event), and

2. Noble gas releases via station vents to the atmosphere.

Method IA is implemented by the EMS software as described in Appendix C. Gaseous release models are detailed in Section 6.7.3 of the EMS Software Requirements Specification (Attachment 3 of Appendix C). 3.8.2 Method II Method II consists of the models, input data (dose factors) and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific data or assumptions have been identified in the ODCM. The general equations (B4 and B-5) taken from Regulatory Guide 1.109, and used in the derivation of the simplified Method I approach as described in the Part B Bases Section 7.2.5, are also applied to Method II assessments. Concurrent meteorology with the release period may be utilized for the atmospheric dispersion factor identified in ODCM Equation 7-15, and determined, as indicated in Part B, Sections 7.3.2 and 7.3.3 for the release point (either ground level or vent stack) from which recorded effluents have been discharged. B.3-24 ODCM Rev. 25

3.9 Method to Calculate the Critical Organ Dose from Iodines, Tritium and Particulates Part A Control C.7.3.1 limits the critical organ dose to a member of the public from radioactive iodines, tritium, and particulates with half-lives greater-than 8 days in gaseous effluents to 7.5 mrem per quarter and 15 mnrem per year per unit. Part A Control C.7.3.1 limits the total body and organ dose to any real member of the public from all station sources (including gaseous effluents) to 25 mrem in a year except for the thyroid, which is limited to 75 mnrem in a year. Use Method I or Method IA first to calculate the critical organ dose from gaseous effluent releases as it is simpler to execute and more conservative than Method II. Use Method II if a more refined calculation of critical organ dose is needed (i.e., Method I or Method IA indicates the dose is greater than the limit). See Part B, Section 7.2.6 for basis. 3.9.1 Method I D= IQ)%/(X/Q).'*t * (Q *DFGi.) (3-8) (nrem) =3 (_, )c*L Ci) -nirem) mn m3u iO where D,0 is the critical organ dose from iodines, tritium, and particulates, (X/Q)d' is the 1-hour depleted atmospheric dispersion factor. (X/Q)d' is the annual average depleted atmospheric dispersion. t a is a unitless adjustment factor to account for a release with a total duration of t hours, Qi is the total activity in ilCi of radionuclide 'i" released to the atmosphere during the period of interest (for strontiums, use the most recent measurement), and DFGi, 0 is the site-specific critical organ dose factor for radionuclide "i", see Tables B.1-12, B.1-14, and B.1-15. (For each radionuclide, it is the age group and organ with the largest dose factor.) Incorporating receptor location-specific atmospheric dispersion factors ((X/Q) ¢b' and (X/Q)."n and adjustment factors (fa) for elevated and ground-level release conditions, and incorporating occupancy factors when applicable (see Section 7.2.7), yields a series of equations by which the critical organ dose can be determined. B.3-25 ODCM Rev. 25

II

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

Dc *) =14.8

• t-029 *(Qi
• DFGi.(,)) (3-8a) mrem (mrem)=( )*( )E(fCi* EC) 4 uCi

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

DCg = 17.7

• t-031 6 * (Qi
• DFGico) (3-8b)

(mrem)=( ( )( Ci

• mremr PCi

c. Science & Nature Center receptor, elevated release conditions:

D E(,) = 3.3 E- 02

• t-0 349* (Qi
• DFGicoEl¢)) (3-8c) pi (mrem) =( ) *( )Ez(,uCi
• mrern)

d. Science & Nature Center receptor ground-level release conditions:

D.E(g = 3.3 E- 02

• t-0-3 47 *
• DFGi1 Exg) (3-8d)

(mrem)=( )m*e( )(Ci* ) PuCi

e. Receptor at the "Rocks"; elevated release conditions:

D.Roc) = 7.3 E- 02

• t-024 8* A(Q.
• DFGacoR(e)) (3-8e)

(mrem)(mrem)=(mnem

             =( ) *( )ZCUflCi

• me) pUCi f Receptor at the "Rocks"; ground-level release conditions:

D R(g) = 8.6 E- 02* t-02 67 * (Q;

• DFGiCOJso) (3-80 (mrem)() * )( Ci
• Cie) pUCi B.3-26 ODCM Rev. 25

Equations 3-8a through 3-8f can be applied under the following conditions (otherwise, justify Method I or consider Method IT): '

1. Normal operations (nonemergency event),

2. Iodine, tritium, and particulate releases via station vents to the atmosphere, and

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

Method IA is implemented by the EMS software as described in Appendix C. Gaseous release models are detailed in Section 6.7.3 of the EMS Software Requirements Specification (Attachment 3 of Appendix C). 3.9.2 Method II Method II consists of the models, input data'and assumptions in Appendix C of Regulatory Guide 1.109, Rev. I (Reference A), except where site-specific data or assumptions have been identified in the ODCM (see Tables B.7-2 and B.7-3). The critical organ dose will be determined based on the location (site boundary, nearest resident, or farm) of receptor pathways, as identified in the most recent annual land use census, or by conservativ'elyassuming the existence of all pathways (ground plane, inhalation, ingestion of stored and leafy vegetables, milk and meat) at an off-site location of maximum potential dose. Concurrent meteorology with the release period may be utilized for determination of atmospheric dispersion factors in' accordance with Part B, Sections 7.3.2 and 7.3.3 for the release point (either ground level or vent stack) from which recorded effluents have been discharged. The maximum critical organ dose will consider the four age groups independently, and use a shielding factor (SF) of 0.7 associated with residential structures. B.3 ODCM Rev. 25

II 3.10 Method to Calculate Direct Dose from Plant Operation Part A Control C.8.1.1 restricts the dose to the whole body or any organ to any member of the public from all uranium fuel cycle sources to 25 mrem in a calendar year (except the thyroid, which is limited to 75 mrem). Direct radiation from contained sources is required to be included in the assessment of compliance with this standard. 3.10.1 Method The direct dose from the station will be determined by obtaining the dose from TLD locations situated on-site near potential sources of direct radiation, as well as those TLDs near the site boundary which are part of the environmental monitoring program, and subtracting out the dose contribution from background. Additional methods to calculate the direct dose may also be used to supplement the TLD information, such as high pressure ion chamber measurements, or analytical design calculations of direct dose from identified sources (such as solid waste storage facilities). The dose determined from direct measurements or calculations will be related to the nearest real person off-site, as well as those individuals on-site involved in activities at either the Education Center or the Rocks boat landing, to assess the contribution of direct radiation to the total dose limits of Part A Control C.8.1.1 in conjunction with liquid and gaseous effluents. B.3-28 ODCM Rev. 25

3.11 Dose Projections Part A Controls C.6.3.1 and C.7.4.i require that appropriate portions of liquid and gaseous radwaste treatment systems, respectively; be used to reduce radioactive effluents when it is projected that the resulting dose(s) would exceed limits which represent small fractions of the "as low as reasonably achievable" criteria of Appendix I to 10 CFR Part 50. The surveillance requirements of these Part A Controls state that dose projections be performed at least once per 31 days when the liquid radwaste treatment systems or gaseous radwaste treatment systems are not being fully utilized. Since dose assessments are routinely performed at least once per 31 days to account for actual releases, the projected doses shall be determined by comparing the calculated dose from the last (typical of expected operations) completed 3 1-day period to the appropriate dose limit for use of radwaste equipment, adjusted if appropriate for known or expected differences between past operational parameters and those anticipated for the next 31 days. 3.1 1.1 Liquid Dose Proiections

The 31-day liquid dose projections are calculated by the following:

a. Determine the total body Dt and organ dose Drno (Equations 3-1 and 3-2, respectively) for the last typical completed 31-dayperiod. The last typical 31-dayperiod should be one without significant identified operational differences from the period being projected to, such as full power operation vs. periods when the plant is shut down. For periods with identified operational differences, skip to subsection 3.1 1.1.e. below.

b. Calculate the ratio (RI) of the total estimated volume of batch releases expected to be released for the projected period to that actually released in the reference period.

c. Calculate the ratio (R2 ) of the estimated gross primary coolant activity for the projected period to the average value in the reference period. Use the most recent value of primary coolant activity as the projected value if no trend in decreasing or increasing levels can be determined.

d. Determine the projected dose from:

Total Body: Du pr Dab . RI . R2 Max. Organ: Df' ' = D. . RI .'Ri '

e. During periods when significant operational differences are identified, such as shutdowns vs. normal power operations, or when specific treatment components are expected to be bypassed or out of service for repair or maintenance, the projected dose should be based on an assessmentof the expected amount of radioactivity that could be discharged, both through treated and any untreated pathways, over the next 31 days. Specific consideration should be given to effluent streams and treatment systems noted on Figure B.6-1. The volume of liquid to be released,-the current or projected maximum radioactivity concentration in the effluent streams either prior to treatment or at the point of release to the environment, and the duration of expected release evaluations should be estimated as part of the projection of offsite dose.

B.3-29 ODCM Rev. 25

I I For these periods outside the bounds of steps 3.1 1.1.a. when significant operational differences exist from the last reference period, the projected dose to the total body Dib and organ dose Dno shall use Equations 3-1 and 3-2, respectively to project dose for each definable time segment of release evolution and summed over the next 31 days. The radioactive release quantity, Qi, in equations 3-1 and 3-2 represents the estimated quantity of radionuclide "i" estimated to be released over the next 31 days, or during short time periods for defined plant operational evaluations, based on expected volumes, concentrations and treatment options to be applied. The EMS software can also be used to perform monthly projected dose calculations as described in Appendix C. The methodology applied by EMS in projecting liquid doses is outlined in Section 2.7 of Attachment 4 to Appendix C (EMS Technical Reference Manual). 3.11.2 Gaseous Dose Proiections

1. For the gaseous radwaste treatment system, the 31-day dose projections are calculated by the following:

a. Determine the gamma air dose Dr, (Equation 3-6a), and the beta air dose DE (Equation 3-7a) from the last typical 3 i-day operating period. The last typical 31-day period should be one without significant identified operational differences from the period being projected to, such as full power steady state operation vs.

periods when the plant is shutdown. For periods with identified operational differences, skip to subsection 3.11.2.2.e. below.

b. Calculate the ratio (R3) of anticipated number ofcuries of noble gas to be released from the hydrogen surge tank to the atmosphere over the next 31 days to the number of curies released in the reference period on which the gamma and beta air doses are based. If no differences between the reference period and the next 31 days can be identified, set R3 to 1.

c. Determine the projected dose from:

Gamma Air: DfrDp, =f r..R3 Beta Air: Drp - D&. *R3

2. For the ventilation exhaust treatment system, the critical organ dose from iodines, tritium, and particulates are projected for the next 31 days by the following:

a. Determine the critical organ dose D,0 (Equation 3-8a) from the last typical 31-day operating period. (If the limit of Part A Control C.7.4.1.c (i.e., 0.3 mrem in 31 days) is exceeded, the projected controlled area annual total effective dose equivalent from all station sources should be assessed to assure that the 10 CFR 20.1301 dose limits to members of the pubilic are not exceeded.)* . The last typical 31-day period should be one without significant identified operational differences from the period being projected to, such as full power steady state operation vs. periods when the plant is shutdown. For periods with identified operational differences, skip to subsection 3.11.2.2.e. below.

B.3-30 ODCM Rev. 25

b. Calculate the ratio (R4 ) of anticipated primary coolant dose equivalent I-131 for the next 31 days to the average dose equivalent I-131 level during the reference period.

Use the most current determination of DE I-131 as the projected value if no trend can be determined.

c. Calculate the ratio (Rs) of anticipated primary system leakage rate to the average leakage rate during the reference period. Use the current value of the system leakage as an estimate of the anticipated rate for the next 31 days if no trend can be determined.

d. Determine the projected dose from:

Critical Organ: Dco- = Dco. R4.1

e. During periods when significant operational differences are identified, such as shutdowns vs. normal power operations, or when specific treatment components are expected to be bypassed or out of service for repair or maintenance, the projected dose should be based on an assessment of the expected amount of radioactivity that could be discharged, both through treated and any untreated pathways, over the next 31 days. Specific consideration should be given to effluent streams and treatment systems noted on Figure B.6-2. The volume or flow rate of gas to be released, the current or projected maximum radioactivity concentration in the effluent streams either prior to treatment or at the point of release to the environment, and the duration of expected release evaluations should be estimated as part of the projection of offsite dose.

For these periods outside the bounds of steps 3.11.2.1.a or 3.11.2.2.a. when significant operational differences exist from the last reference period, the projected air dose from gamma and beta emissions from noble gases (Equations 3-6 and Equations 3-7, respectively), or from iodines, tritium, and particulates (Equations 3-8) shall use the referenced equations to project dose for each definable time segment of release evolution and summed over the next 31 days. The radioactive release quantity, Qi in the dose equations represents the estimated quantity of radionuclide "i" estimated to be released over the next 31 days, or during short time periods for defined plant operational evaluations, based on expected volumes, concentrations and treatment options to be applied.

3. Alternate Projection Method for Use with Containment Ventilation Exhaust Treatment System (Charcoal Filters)

During periods when the Containment Building air needs to be vented to the atmosphere, the decision to use the Containment charcoal filter train to exhaust Containment air can be based on dose conversion factors and critical organ dose equation that reflect only those real exposure pathways in the offsite environment as indicated by the annual Land Use Census. This reduces the excess conservatism associated with the standard Method I assumptions that all typical (potential) exposure pathways (including milk) may exist at the most limiting atmospheric dispersion point off site. B.3-31 ODCM Rev. 25

I I In place of the dose conversion factors found in Table B.I-12, and critical organ dose equation 3-8a for Dco, Chemistry Department technical evaluation CHSTI) 02-004 contains the dose conversion factors (DFG) and critical organ dose equation which were developed in the same manner as the current Method I factors and time dependent dose equation, but which utilize the most recent Land Use Census data to define which exposure pathways and identified receptor locations exist. CHSTID 02-004 documents the development of this altemate dose projection method. After the Land Use Census is performed each year, and before application to any Containment venting evolution, CHSTID 02-004 will be reviewed to see if any new receptor location impacts the selection of controlling dose location. The EMS software can also be used to perform monthly projected dose calculations as described in Appendix C. The methodology applied by EMS in projecting gaseous dose is outlined in Section 3.8 of Attachment 4 to Appendix C (EMS Technical Reference Manual). B.3-32 ODCM Rev. 25

• Note: This action is based on the assumption that tritium is the controlling nuclide for whole body exposures through the inhalation pathway. Maximum annual average on-site X/Q's for station effluent release points are approximately 100 times the values used for the site boundary dose calculations. However, the site boundary doses calculated by the ODCM for iodines, tritium, and particulates with half lives greater than 8 days, includes all potential off-site exposure pathways. For tritium, the inhalation pathway only accounts for 10% of the total dose contribution being calculated. As a result, if the monthly calculation indicates that the site boundary maximum organ dose reached 0.3 mrem, the on-site maximum dose due to inhalation would be approximately 3.0 mrem for this period. If this were projected to continue for a year with a 2000 hour occupancy factor applied, the projected inhalation whole body dose would be approximately 8 mrem, or 8% of the 10 CFR 20.1301 limit. This is a reasonable trigger value for the need to consider the dose contribution from all station sources to members of the public in controlled areas.

B333 ODCM Rev. 25

Il 3.12 Method to Calculate Total Dose From Plant Operations ODCM Control C.8.1.1 restricts the annual dose to the whole body or any organ of a member of the public from all uranium fuel cycle sources (including direct radiation) to 25 mrem (except the thyroid, which is limited to 75 mrem). These cumulative dose contribution limits from liquids and gaseous effluents, and direct radiation, implement the Environmental Protection Agency (EPA) 40 CFR 190, "Environmental Standards for the Uranium Fuel Cycle." 3.12.1 Method Compliance with the Seabrook Station Effluent Controls dose objectives for the maximum individual, as calculated by the methods described in sections B.3.2, B.3.3, B.3.7, B.3.8, B.3.9 of the ODCM also demonstrates compliance with the EPA limits to any member of the public. This indirect determination of compliance is based on the fact that the Effluent Control liquid and gaseous dose objectives are taken from 10 CFR 50, Appendix I, and represent lower values than the 40 CFR 190 dose limits. Direct radiation dose from contained sources is not expected to be a significant contributor to the total dose to areas beyond the site boundary. If the operational dose objectives in the Seabrook ODCM Effluent Controls C.6.2.1.a, C.6.2.l.b, C.7.2.l.a, C.7.2.L.b, C.7.3.1 .a, or C.7.3.1.b are determined to be exceeded by a factor of two, a Special Report must be prepared. The purpose of this Special Report is to determine by direct assessment ifthe cumulative dose (calendar year) to any member of the public (real individual) from all sources is within the limits of the Total Dose Control C.8. 1.1. In addition, section A.10.2, "Annual Radioactive Effluent Release Report," requires that an assessment of radiation doses to the likely most exposed member of the public from all effluent and direct radiation sources be included for the previous calendar year to show compliance with 40 CFR 190. When required, the total dose to a member of the public will be calculated for all significant effluent release points for all real pathways, including direct radiation. Only effluent releases from Seabrook Station need be considered since no other uranium fuel cycle facilities exist within five miles. EPA has determined that for fuel cycle facilities separated by more than five miles, their contribution to each other's total dose would not be significant and cause dose Standard for the Uranium Fuel Cycle to be exceeded. The calculations will be based on the liquid and gaseous Methods II dose models as described in Section B.3, including usage factors and other documented site-specific parameters reflecting realistic assumptions, where appropriate. The liquid and gaseous effluent Method II models are derived from the methods given in Regulatory Guide 1.109, Rev. 1, October 1977. The direct radiation component from the facility can be determined using environmental TLD results as noted in Section B.3.I0.1 (or alternately, high pressure ion chamber measurements or analytical design calculations for estimating the direct radiation dose from identified contained radioactive sources within the facility). B.3-34 ODCM Rev. 25

4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM The radiological environmental monitoring stations are listed 'in Table B.4-1. The locations of the stations with respect to the Seabrook Station are shown on the maps in Figures B.4-1 to B.4-6. Direct radiation measurements are analyzed at the station. All other radiological analyses for environmental samples are performed at a contractor laboratory. The contractor laboratory participates in an Interlaboratory Comparison Program for all relevant species in an aqueous (water) matrix. An independent vendor (Analytics) supplies the remaining cross check samples. These samples are presented on an air filter and in'milk and water matrices. Pursuant to Part A Surveillance S.9.2.1, the land use census will be conducted "during the growing season" at least once per 12 months. The growing season is defined, for the purposes of the land use census, as the period from June I to October 1. The method to be used for conducting the census will consist of one or more of the following, as appropriate: door-to-door survey, visual inspection from roadside, aerial survey, or consulting with local agricultural authorities. Technical Specification 6.8.1.3 and Part A, Section 10.1 of the ODCM require that the results of the Radiological Environmental Monitoring Program be summarized in the Annual Radiological Environmental Operating Report "in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, 1979." The general table format will be used with one exception and one clarification, as follows. The mean and range values will be based not upon detectable-measurements only, as specified in the NRC Branch Technical Position, but'uponall measurements. This will prevent the positive bias associated with the calculation of the mean and range based upon detectable measurements only. Secondly, the Lower Limit of Detection column will specify the LLD required by ODCM Table A.9.1-2 for that radionuclide and sample medium. B.4-1 ODCM Rev. 26

TABLE B.4-1 RADIOLOGICAL ENVIRONMENTAL MONITORING STATIONS(a) Exposure Pathway Sample Location Distance From Direction From and/or Sample and Designated Code Unit I the Plant Containment (km)

1. AIRBORNE (Particulate and Radioiodine)

AP/CF-01 PSNH Barge Landing Area 2.6 ESE AP/CF-02 Harbor Road 2.5 E AP/CF-03 SW Boundary 1.0 SW AP/CF-04 W. Boundary 1.2 W AP/CF-05 Winnacunnet H.S.(b) 4.0 NNE AP/CF-07 (PSNH Substation) 5.7 NNW AP/CF-08 E&H Substation(b) 3.4 SSE AP/CF-09 Georgetown 21.4 SSW Electric Light (Control)

2. WATERBORNE

a. Surface WS-01 Hamrpton-Discharge Area 5.3 E WS-51 Ipswich Bay (Control) 16.9 SSE

b. Sediment SE-02 Hampton-Discharge AreaPb) 5.3 E SE-07 Hampton Beach(b) 3.1 E SE-08 Seabrook Beach 3.2 ESE SE-52 Ipswich Bay (Control)(b) 16.9 SSE SE-57 Plum Island Beach 15.9 SSE (Control)(b)

3. INGESTION

a. Milk TM-09 Hampton, NH (d) 5.3 NNW TM-15 Hampton Falls, NH (d) 6.9 NW TM-20 Rowley, MA (d) 17.0 S TM-23 Newbury, MA(d) 12.0 S

b. Fish and Invertebrates(C)

FH-03 Hampton - Discharge Area 4.5 ESE FH-53 Ipswich Bay (Control) 16.4 SSE HA-04 Hampton - Discharge Area 5.5 E HA-54 Ipswich Bay (Control) 17.2 SSE MU-06 Hampton - Discharge Area 5.2 E MU-09 Hampton Harbor"b) 2.6 E MU-56 Ipswich Bay (Control) 17.4 SSE MU-59 Plum Island(b) 15.8 SSE B.4-2 ODCM Rev. 26

TABLE B.4-1 RADIOLOGICAL ENVIRONMENTAL MONITORING STATIONS(a) (Continued) Exposure Pathway Sample Location Distance From Direction From and/or Sample and Designated Code Unit I the Plant Containment (km)

c. Food Products TG-01 Site Boundary 1.05 W TG-02 Site Boundary .94 SW TG-03 Georgetown Light 21.4 SSW

4. DIRECT RADIATION TL-1 Brimmers Lane, .97 N Hampton Falls TL-2 Landing Rd., Hampton 3.0 NNE TL-3 Glade Path, Hampton 2.9 NE Beach TL-4 Island Path, Hampton 2.3 ENE Beach TL-5 Harbor Rd., Hampton 2.6 E Beachli TL-6 PSNH Barge Landing 2.7 ESE Area TL-7 Cross Rd., Seabrook Beach 2.6 SE TL-8 Farm Lane, Seabrook 1.3 SSE TL-9 Farm Lane, Seabrook 1.3 S TL-10 Site Boundary Fence 1.2 SSW TL-1 1 Site Boundary Fence 1.0 SW TL-12 Site Boundary Fence 1.2 WSW TL-13 Inside Site Boundary 1.2 W TL-14 TrailerPark, Seabrook 1.3 WNW TL-15 Brimmer's Lane, 1.4 NW Hampton Falls TL-16 Brimmer's Lane, 1.2 NNW Hampton Falls TL-17 South Rd., N. Hampton 7.8 N TL-18 Mill Rd., N. Hampton 7.6 NNE TL-19 Appledore Ave., 7.7 NE N. Hampton TL-20 Ashworth Ave., 3.2 ENE Hampton Beach TL-21 Route 1A, Seabrook Beach 3.7 SE TL-22 Cable Ave., 7.6 SSE Salisbury Beach B.4-3' ODCM Rev. 26

I L TABLE B.4-1 RADIOLOGICAL ENVIRONMENTAL MONITORING STATIONS(a) (Continued) Exposure Pathway Sample Location Distance From Direction From and/or SaMple and Designated Code Unit 1 the Plant Containment (km) TL-23 Ferry Rd., Salisbury 8.1 S TL-24 Ferry Lots Lane, 7.2 SSW Salisbury TL-25 Elm St., Amesbury 7.6 SW TL-26 Route 107A, Amesbury 8.1 WSW TL-27 Highland St., S. Hampton 7.5 W TL-28 Route 150, Kensington 7.5 WNW TL-29 Frying Pan Lane, 7.2 NW Hampton Falls TL-30 Route 27, Hampton 7.6 NNW TL-31 Alumni Drive, Hampton 3.8 NNE TL-32 Seabrook Elementary School 2.0 S TL-33 Dock Area, Newburyport 9.8 S TL-34 Bow St., Exeter 12.0 NW TL-35 Lincoln Ackerman School 2.3 NNW TL-36 Route 97, Georgetown 22.6 SSW (Control) TL-37 Plaistow, NH (Control) 21.5 WSW TL-38 Hampstead, NH (Control) 27.7 W TL-39 Fremont, NH (Control) 27.0 WNW TL-40 Newmarket, NH (Control) 21.6 NNW TL41 Portsmouth, NH, (Control)(b) 21.0 NNE TL-42 Ipswich, MA (Control)pb) 22.8 SSE B.4-4 ODCM Rev. 26

TABLE B.4-1 RADIOLOGICAL ENVIRONMENTAL MONITORING STATIONS(a)

                                            '(Continued)

(a) Sample locations are shown on Figures B.4-lt'6B.4-6. (b) This sample location is not required by monitoring program defined in Part A of ODCM; program requirements specified in Part Ado not apply to samples taken at this location. (c) Samples will be collected pursuant to ODCM Table A.9.1-1. Samples are not required from all stations listed during any sampling interval (FH = Fish; HA = Lobsters; MU = Mussels). Table A.9.1-1 specifies that "one sample of three commercially and recreationally important species" be collected in the vicinity of the plant discharge area, with similar species being collected at'a control location. (This wording is consistent with the NRC Final Environmental Statement for Seabrook Station.) Since the discharge area is off-shore, there is a great number of fish species that could be considered commercially or recreationally important.. Some are migratory (such as striped bass), making them less desirable as an indicator of plant-related radioactivity. Some pelagic species (such as herring and mackerel) tend to school and wander throughout a large area, sometimes making catches of significant size difficult to obtain. Since the collection of all species would be difficult or impossible, and would provide unnecessary redundancy in terms of monitoring important pathways to man, three fish and invertebrate species' have been specified as a minimum requirement. Samples may include marine fauna such as' lobsters, clams, mussels, and bottom-dwelling fish, such as flounder or hake. Several similar species may be grouped together into one, sample if sufficient sample mass for a single species is not available'afler a reasonable effort has been made (e.g., yellowtail flounder and winter flounder) (d) Monitoring program defined in Part A of ODCM does not require'this sample location; food product sampling is being implemented in lieu of an insufficient number ok milk locations. BA4-5 ODCM Rev. 26

FIGURE B.A-1i RADIOLOGICAL ENVIRONMENTAL MONITORING LOCATIONS WITHI 4 KILOMETERS OF SEABROOK STATION B.466 ODCM Rev. 26

FIGURE B.4-2 RADIOLOGICAL ENVIRONMENTAL MONITORING LOCATIONS BETWEEN 4 KILOMETERS AND 12 KILOMETERS FROM SEABROOK STATION B.4-7 ODCM Rev. 26

I I FIGURE B.4-3 RADIOLOGICAL ENVIRONMENTAL MONITORING LOCATIONS OUTSIDE 12 KILOMETERS OF SEABROOK STATION 0 B.4-8 ODCM Rev. 26

FIGURE B.4-4 DIRECT RADIATION MONITORING LOCATIONS WITHIN 4 KILOMETERS OF SEABROOK STATION HE

                                             \JL-4     ENE.

lTL-t35 A nL-16  : 0 ATL-23

             //TAAPTO                               HARER TL-I,   13A B.4-9                      ODCM Rev. 26

II FIGURE B.4-5 DIRECT RADIATION MONITORING LOCATIONS BETWEEN 4 KILOMETERS AND 12 KILOMETERS FROM SEABROOK STATION B.4-10 ODCM Rev. 26

FIGURE B.4-6 DIRECT RADIATION MONITORING LOCATIONS OUTSIDE 12 KILOMETERS OF SEABROOK STATION B.4-11 ODCM Rev. 26

5.0 SETPOINT DETERMINATIONS Chapter 5 contains the methodology for the calculation of effluent monitor setpoints to W implement the requirements of the radioactive effluent monitoring systems Part A Controls C.5. 1 and C.5.2 for liquids gases, respectively. Example setpoint calculations are provided for each of the required effluent monitors. 5.1 Liquid Effluent Instrumentation Setpoints Part A Control C.5.1 requires that the radioactive liquid effluent instrumentation in Table A.5.1-1 of Part A have alarm setpoints in order to ensure that Part A Control C.6.1.1 is not exceeded. Part A Control C.6.1 .1 limits the activity concentration in liquid effluents to ten times the ECL values in '10 CFR 20, Appendix B, Table 2, and a total noble gas MPC. 5.1.1 Liquid Waste Test Tank Monitor (RM-6509) The liquid waste test tank effluent monitor provides alarm and automatic termination of release prior to exceeding ten times the concentration limits specified in 10 CFR 20, Appendix B, Table 2, Column 2 to the environment. It is also used to monitor discharges from various waste sumps to the environment. 5.1.1.1 Method to Determine the Setpoint of the Liguid Waste Test Tank Monitor (RM-6509) The alarm setpoint is based on ensuring that radioactive effluents in liquid waste are in compliance with Control limits which are based on the concentration limits in Appendix B to 10 CFR 20. The alarm point depends on available dilution flow through the discharge tunnel, radwaste discharge flow rate from the test tanks, the isotopic composition of the liquid waste, and the monitor response efficiency and background count rate applicable at the time of the discharge. The alarm/trip setpoint is determined prior to each batch release taking into account current values for each variable parameter. The following steps are used in determining the monitor setpoint: First, the minimum required dilution factor is determined by evaluating the isotopic analysis of each test tank to be released along with ECL'requirements for each radionuclide. The most recent analysis data for tritium and other beta emitters that are analyzed only monthly or quarterly on composite samples can be used as an estimate of activity concentration in the tank to be released. For noble gases, the Control limit (C.6.1.1) is defined as 2E-04 pCi/ml total for all dissolved and entrained gases. Therefore, DF&i = Ci or E CNG , whichever is larger. (5-3) IOEMb 2E-04 Where: DFmn = Minimum required dilution factor necessary to ensure that the sum of the ratios for each nuclide concentration divided by its ECL value is not greater than 10 (dimensionless). B.5-l ODCM Rev. 25

I I CQ = Activity concentration of each radionuclide "i" (except noble gases) determined to be in the test tank (IiCi/mil). This includes tritium and other non gamma emitting isotopes either measured or estimated from the most recent composite analysis. C.s = The sum of all dissolved and entrained noble gases identified in each test tank (tCiUml). ECL, = Effluent concentration limit (ECL) for radionuclide "i" (except for dissolved an entrained noble gas) in pCi/ml as specified in 10 CFR 20, Appendix B, Table 2 See ODCM, Appendix B, for a listing. In the event that no activity is expected to be discharged, or can be measured in the system, the liquid monitor setpoint should be based on the most restrictive ECL for an "unidentified" mixture or a mixture known not to contain certain radionuclides as given in 10 CFR 20, Appendix B, notes. 2E-04 = The total dissolved and entrained noble gas Technical Specification concentration limit in liquid effluents from the plant (AiCi/mil). Next, the available dilution flow through the discharge tunnel (Fd), or a conservative estimate for it, is divided by the minimum dilution factor (DFmmn) to determine the maximum allowable discharge flow rate (Fog) that the test tanks could be released at without exceeding the ECL limits, assuming no additional radioactive flow paths are discharging at the time of release of the test tanks. Therefore, FnW Fd DF Where: Fnx = The maximum allowable discharge flow rate from the test tank past the monitor which would equate to the Control concentration limit for the radioactivity mixture determined to be in the test tank (gpm). Fd = The actual or conservative estimate of the flow rate out of the discharge tunnel (gpm).

• B.5-2 ODCM Rev. 25

For Waste Test Tank (WTT) releases, tritium is expected to be the radionuclide with the highest concentration, and therefore requires the highest dilution flow in order to satisfy the discharge concentration limits: Unlike concentrations of other dissolved or suspended radionuclides, tritium concentrations are not expected to vary because they are unaffected by plant cleanup systems used to reduce or control waste radioactivity levels. As such, events that cause sudden increases in the concentrations of other dissolved or suspended radionuclides, such as changes in waste cleanup efficiencies, crud bursts or failed fuel fractions would not change the tritium concentrations. As long as the minimum required dilution factor (DFm.n) for all radionuclides present in the liquid waste is satisfied, the alarm setpoint for the Waste Test Tank monitor need only consider the potential changes to the concentrations of detectable gamma-emitting radionuclides. Therefore, the required dilution for detectable activity by the WTT monitor can be determined by applying the definition of DFmin (given in equation (5-3)) to only the gamma-emitting radionuclides present in the waste. DFin1 = E (Ciy/1 OECL,- ) (5-3a) Where: DFmn = Minimum required dilution factor necessary to ensure that the sum of the ratios for the concentration of each gamma-emitting radionuclide to the respective ECL value is not greater than 10 (dimensionless). Ci = Activity concentration of each detectable gamma-emitting radionuclide "i" in the mixture (pCi/ml). ECI, = As defined previously. As in the determination of Fr, for the total radioactivity mixture, the maximum allowable discharge flow rate that the waste from the test tanks could be released at without exceeding the concentration limit for gamma-emitters, Fn,,,, is obtained by dividing the discharge tunnel flow, Fd, by DF,,fin. This determination is based on the assumption that there are no additional discharges of liquid waste at the time of release from the test tanks. Therefore, Fmrt,=Fd ,DFmn-d Where: r= The maximum allowable discharge flow rate from the test tank past the monitor which would equate to the control concentration limit for the gamma radioactivity mixture determined to be in the test tank (gpm). Fd = The actual or conservative estimate of the flow rate out of the discharge tunnel (gpm). B.5-3 ODCM Rev. 25

I I The selection of the actual discharge flow rate (Fm) from the test tanks compared to the maximum allowable discharge rate based on all radionuclides that are present (F,,.) and the maximum allowable discharge rate based on only gamma-emitting radionuclides that are present (F",,.y) must satisfy the following: Fm < Fmax

• f<Fmaxy
• ftt Where the ftt represents an administrative fraction of the maximum allowable discharge flow from the test tanks. This fraction provides additional margin in meeting ECL limits for non-gamma emitters (such as tritium) at the discharge point to the ocean when other flow paths may contribute to the total site release at the time of tank discharges and minimum dilution flow conditions exist.

With the above conditions on discharge and dilution flow rates satisfied, the alarm/trip setpoint for the monitor which corresponds to the maximum allowable concentration at the point of discharge is determined as follows: Rw = f x Fd xZCXi (5-1) Where: Rsetpoint = The maximum allowable alarm/trip setpoint for an instrument response (jiCi/ml) that ensures the limiting concentration at the point of discharge is not exceeded.

                       =     The fraction of the total contribution of ECL at the discharge point to be associated with the test tank effluent pathway, where f2 , f3 , and 4 are the fractions for the Turbine Building Sump, Steam Generator Blowdown, and Primary Component Cooling pathways contributions to the total, respectively (f1 +f2+f3 +f4 < 1). Each of the fractions may be conservatively set administratively such that the sum of the fractions is less than 1. This additional margin can be used to account for the uncertainty in setpoint parameters such as estimated concentration of non gamma emitters that are based on previous composite analyses of the waste stream.

B.54 ODCM Rev. 25

- 5.1.1.2 Liquid Waste Test Tank Monitor Setpoint Example The radioactivity concentration of each radionuclide, Ci, in the waste test tank is determined by analysis of a representative grab sample obtained at the radwaste sample sink, and analyzed prior to release for gamma emitters, or as part of a composite analysis for non gamma emitters. The maximum allowable instantaneous effluent concentrations (i.e., ten time the ECL values in 10 CFR 20, Appendix B, Table 2) are used to illustrate a monitor setpoint determination. This setpoint example is based on the following data: Ten Times i C, (KUCiml) ECLI (piCi/ml) ECLI (pCi/ml) Cs-134 2.15E-05 9E-07 9E-06 Cs-137 7.48E-05 1E-06 lE-05 Co-60 2.56E-05 3E-06 3E-05 H-3 - 1.50E-01 lE-03 lE-02 The minimum required dilution factor for this mix of radionuclides (including beta-emitters) is: C1 2 ___ 7.48E-065 2.56E-0 - 1.50E-01 DFR = - _ 2.15E-05+i + E + = 26 IOECLi 9E-06 1E-05 3E-05 IE-02 The release flow rate (Fo)from the waste test tanks can be set between 10 and 150 gpm. The cooling water tunnel discharge dilution flow rate (Fd) can typically vary from approximately 8,800 to 412,000 gpm depending on the operating status of the plant. In this example, if the dilution flow (Fd) is taken as 412,000 gpm, the maximum allowable discharge rate (F,) is: Fd

                        -;-   -~       D 412,000g
                                 ,26
                            = 15,846 gpm Next, the required dilution factor for only gamma emitters in the mix is:

2.1 5E -05 :7.48E -'05 +2.56E - 05 DFYiy (C,7 /IOECL~).

                                     -    9E-06,-      1E 05      - 3E-05 The maximum allowable discharge flow rate (F,,C) considering only gamma emitters is given as:

DFd - 12 00 Fnuxy DFd1200 gpm = 37,455 gpmn B.5-S ODCM Rev. 25

I I With the selected release rate from the test tank set at 150 gpm, and the administrative flow fraction (ft) assumed in this example to be 0.7, the condition for the control concentration limits is met since: Fm (equal to 150) < Fn.,, (equal to 15,846 gpm) x ft (set at 0.7)

                   < Frm (equal to 37,455 gpm x F., (set at 0.7) 150     <    11092     <   26219 and the monitor response due to the mix of the gamma emitters is:

i CatO(giCi/ml) Cs-134 2.15E-05 Cs-137 7.48E-05 Co-60 2.56E-05 ECE= 1.22E-04 pCi/ml Under these conditions, the alarm/trip setpoint for the liquid radwaste discharge monitor is: Ras, fi x Fd x E C-i (5-1) F.mxDFi,m,.

                        =  0.4x 1'2,00 x1.22E-04 150 x I
                        = 1.22E - 02 gCi/ml In this example, the alarm/trip setpoint of the liquid radwaste discharge monitor can be put at 1.22E-02 ACi/ml above background. For the example, it is assumed that the test tank release pathway will be limited to only 40% of the total site discharge allowable concentration.

5.1.2 Turbine Building Drains Liquid Effluent Monitor (RM-6521) The Turbine Building drains liquid effluent monitor continuously monitors the Turbine Building sump effluent line. The only sources to the Sump Effluent System are from the secondary steam system. Activity is expected in the Turbine Building Sump Effluent System only if a significant primary-to-secondary leak is present. If a primary-to-secondary leak is present, the activity in the sump effluent system would be comprised of only those radionuclides found in the secondary system, with reduced activity from decay and dilution. B.5-6 ODCM Rev. 25

The Turbine Building drains liquid effluent monitor provides alarm and automatic termination of release prior to exceeding ten times the concentration limits specified in 10 CFR 20, Appendix B, Table 2, Column 2 to the environment. The alarm setpoint for this monitor will be determined using the same method as that of the liquid waste test tank monitor 'if the total sump activity is greater than the ECL, as determined by the most recent grab sample isotopic analysis: If the total activity is less than the ECL, the setpoints of RM-6521 are calculated as follows: High Trip Monitor Setpoint (iCi/ml) = f2 (DF') ("unidentified mix ECL" (gCilml)) (5-21) where: Circulating water flow rate (gpm) DF' Flow rate pass- monitor (gpm) unidentified mix ECL = most restrictive ECL value (piCi/ml) for an unidentified mixture or a mixture known not to contain certain radionuclides as given in 10 CFR 20, Appendix B, Notes. f2 1- (f1 + f3 + 4); where the f values are described above. In addition, a warning alarm setpoint can be determined by multiplying the high trip alarm point by an administratively selected fraction (as an example, 0.25). [Warning AlarmHihTp Monitor Setpoint =igh Trip. (0.25) Monitor Sepoint (MCi/mi) - 5.1.3 Steam Generator Blowdown Liguid Sample Monitor (RM-65l9) The steam generator blowdown liquid sample monitor is used to detect abnormal activity concentrations in the steam generator blowdown flash tank liquid discharge. The alarm setpoint for the steam generator blowdown liquid sample monitor, when liquid is to be discharged from the site, will be determined using the same approach as the Turbine Building drains liquid effluent monitor. For any liquid monitor, in the event that no activity is expected to be discharged, or can be measured in the system, the liquid monitor setpoint should be based on the most restrictive ECL for an "unidentified" mixture or a mixture known not to contain certain radionuclides given in 10 CFR 20, Appendix B notes.

1. I I B.5-7 ODCM Rev. 25

I I 5.1.4 PCCW Head Tank Rate-of-Change Alarm Setpoint A rate-of-change alarm on the liquid level in the Primary Component Cooling Water (PCCW) head tank will work in conjunction with the PCCW radiation monitor to alert the operator in the Main Control Room of a leak to the Service Water System from the PCCW System. For the rate-of-change alarm, a setpoint is selected based on detection of an activity level equivalent to I 0-8 gCi/ml in the discharge of the Service Water System. The activity in the PCCW is determined in accordance with the liquid sampling and analysis program described in Part A, Table A.6.I-1 of the ODCM and is used to determine the setpoint. The rate-of-change alarm setpoint is calculated from: RC,,, =IxIO"e* SWF

• 1 (5-23)

(gal) P(oi ) (gal) ( ml ) hr) ml h-r (;Fi where: RC, = The setpoint for the PCCW head tank rate-of-change alarm (in gallons per hour). lxlO- = The minimum detectable activity level in the Service Water System due to a PCCW to SWS leak (pCi/ml). SWF = Service Water System flow rate (in gallons per hour). PCC = Primary Component Cooling Water measured (decay corrected) gross radioactivity level (giCi/ml). As an example, assume a PCCW activity concentration of lx10-5 pCi/mI with a service water flow rate of only 80 percent of the normal flow of 21,000 gpm. The rate-of-change setpoint is then: RC,, = lxlO-s .x0gph (1/lxlO-5 C'1) ml ml As a result, for other PCCW activities, the RCset which would also relate to a detection of a minimum service water concentration of Ix10 8 pCi/ml can be found from: RC^, - x10 x uCi/mlx 1000 gph (5-24) RCS& ~PCC (-4 B.5-8 ODCM Rev. 25

5.1.5 PCCW Radiation Monitor

     . The PCCW radiation monitor-will alert the operator in the Main Control Room of a leak to the PCCW System from a radioactively contaminated system.

The PCCW radiation monitor alarm is based on a trend of radiation levels in the PCCW System. The background radiation of the PCCW is determined by evaluating the radiation levels over a finite time period. The alert alarm setpoint is set at 1.5 x background, and the high alarm setpoint is set at 2 x background, per Technical Specification Table 3.3-6.

.B.5-9 ODCM Rev. 25

IL 5.2 Gaseous Effluent Instrumentation Setpoints Part A Control C.5.2 requires that the radioactive gaseous effluent instrumentation in Table A.5.2-1 of Part A have their alarm setpoints set to insure that Part A Control C.7.1.1 is no exceeded. 5.2.1 Plant Vent Wide-Ranze Gas Monitors (RM-6528-1. 2 and 3) The plant vent wide-range gas monitors are shown on Figure B.6-2. 5.2.1.1 Method to Determine the Setpoint of the Plant Vent Wide Range Gas Monitors (RM-6528-1. 2 and 3) The maximum allowable setpoint for the plant vent wide-range gas monitor (readout response in p.Ci/sec) is set by limiting the off-site noble gas dose rate to the total body or to the skin, and denoted R.stpoin. R oint is the lesser of: R=588 I f, (5-5) DFBC Cisc mrem-,u

            ,uCi/sec=(          p.i Ci-niM se      ( pCi- yr 3) yr-pCi-sec         mrem-m and:

Rs 3 X000

                      ,=           f,                                                             (5-6)

Cise mremr A Ci-Yyr yr mrem-sec where: Rh = Response of the monitor at the limiting total body dose rate (,iCi/sec) 500 (mrem-,Ci- m3 588 = (1E+06) (8.5E-07) yr-pCi-sec 500 = The offsite limiting total body dose rate (mrem/yr) from all release points 1E+06 = Number of pCi per gCi (pCi/tCi) 8.5E-07 = [X/Q]r, maximum off-site long-term average gamma atmospheric dispersion factor for primaxy vent stack releases (sec/m 3 ) B.5-10 ODCM Rev. 25

DFBC = t Composite total body dose factor (mrem-m 3 /pCi-yr)

           =  E:QDFBI                                                                 (5-7) fE          =    The fraction of the offsite limiting total body dose rate to be administratively assigned to the plant vent (f, < I -fg, where fg is the fraction of the limiting dose rate to be assigned to monitored ground level releases)
           =   The relative release rate of noble gas "i" in the mixture, for each noble gas identified or postulated to be in the off-gas (jiCi/sec)

DFBi = Total body dose factor (see Table B.1-10) (mrem-m 3 /pCi-yr) Rskin = Response of the monitor at the limiting skin dose rate (AtCi/sec) 3,000 = The offsite limiting skin dose rate (mrem./yr) DF', = Composite skin'dose factor (mrem-sec/pCi-yr)

               -E  QjDF'ive)                                                    (5-8)

DF'i(c) = Combined skin dose factor for elevated release point (see Table B.1-10) (mrem-sec/ACi-yr) I- -B.5-11 ODCM Rev. 25

I I 5.2.1.2 Plant Vent Wide Range Gas Monitor Setpoint Example for Limiting Case The following setpoint example for the plant vent wide range gas monitors demonstrates the use of equations 5-5 and 5-6 for determining setpoints. Evaluations of potential releases rates associated with the limiting offsite dose rates (Control C.7. 1.1 .a) have been made considering different noble gas mixes related to normal operations, observed periods with fuel defects, and potential UFSAR accident conditions. The bounding noble gas mix case for setpoint alarm indications was found to be related to projected fuel gap activity at the time of shutdown from power operations (UFSAR Table 15.7-20). By setting the maximum alarm setpoint in accordance with this assumed mix, other potential or realistic release conditions will not create an effluent discharge at or above the limiting offsite dose rates without the monitor going into alarm. This limiting setpoint example is based on the following data (see Table B.1-10 for DFBi(,) and DF'i(c)): O; DFBi DF'i(,) (ILCi) mrem-m 3 ,mrem-sec: i._ sec) pCi- yr u Ci- yr Xe-138 2.52E+02 8.83E-03 1.20E-02 Kr-87 7.90E+01 5.92E-03 1.38E-02 Kr-88 1.15E+02 1.47E-02 1.62E-02 Kr-85m 4.49E+01 1.17E-03 2.35E-03 Xe-135 6.82E+01 1.81E-03 3.33E-03 Xe-133 3.23E+02 2.94E-04 5.83E-04 Kr-85 4.13E+00 1.61E-05 1.1 lE-03 Xe-131m 1.15E+00 9.15E-05 5.37E-04 Xe-133m 4.67E+01 2.51E-04 1.1 2E-03 Xe-135m 6.64E+01 3.12E-03 3.74E-03 B.5-12 ODCM Rev. 25

DFB = Q1DFBi (5-7) 2 Q1DFB, = (2.52E+02)(8.83E-03) + (7.90E+01)(5.92E-03) + (1.15E+02) (1.47E-02)

                         + (4.49E+01)(1.17E-03) + (6.82E+01)(1.81E-03) + (3.23E+02) (2.94E-04)
                         + (4.13E+00)(1.61E-05) + (1.15E+00)(9.15E-05) + (4.67E+01) (2.51E-04)
                         + (6.64E+01) (3.12E-03)
                         =  4.86E+00 (ptCi-mrem-m 3 /sec-pCi-yr) 2   Qi          =      2.52E+02 + 7.90E+01 + 1.15E+02 + 4.49E+01 + 6.82E+01
                           +      3.23E+02 + 4.14E+00 + 1.16E+00 + 4.67E+01 + 6.64E+01
                           =      1.OOE+03 pCi/sec 4.86 E+ 00 DFB,                    1.00 E+ 03
                           -     4.86E-03 (mrem-m3/pCi-yr) and therefore:

Rtb = 588 - f (5-5) DFBC

                     -     =     (588)       .          0.7 (4.86 E- 03)
                           =     8.47E+04 pCi/sec
                                               ! ,.B.5-13                 ODCM Rev. 25

I L and next; (5-8) E Di j DF;' = (2.52E+02)(1.20E-02) + (7.90E+01)(1.38E-02) + (1.15E+02) (1.62E-02)

                          +    (4.49E+Ol)(2.35E-03) + (6.82E+01)(3.33E-03) + (3.23E+02) (5.83E-04)
                          + (4.13E+00)(1.1lE-03) + (l.15E+00)(5.37E-04) + (4.67E+01) (1.12E-03)
                          + (6.64E+01) (3.74E-03)
                          =    6.80E+00 (giCi-mrem-sec/sec-giCi-yr) 6.80E+00
                 -1.0E+     03
               = 6.80E - 03       (mrem -sec/pCi - yr) and therefore:

R = 3,000 1 fy (5-6) DFc'

                = (3,000)

(6.8 I- 0.7

                 =   3.09E+05 jiCi/sec The setpoint, R."im, is the lesser of Rth and Rskin. For the limiting noble gas mixture, Rtb is less than Rskin, indicating that the total body dose rate is more restrictive. Therefore, the plant vent wide-range gas monitor should be set at no more than 8.47E+04 pCi/sec above background, or at some administrative fraction of the above value.

B.5-14 ODCM Rev. 25

5.2.2 Waste Gas System Monitors (RM-6504 and RM-6503) Process radiation monitors in the waste gas system provide operational information on the performance of the system before its discharge is combined and diluted with other gas flows routed to the plant vent for release to the environment. The setpoints for the waste gas system monitors are administratively set as small multiples of the expected activity concentration to provide operational control over unexpected changes in gas discharges from the system. Typically, the alert alarm setpoint for both monitors is placed at 1.5 times the expected activity concentration passing the monitor, with the high alarm trip set at 2.0 times the expected concentration flow. Under all conditions, the maximum allowable alarm trip shall not exceed a concentration equivalent to 62.5 pCi/cm 3 . This concentration limit, based on system design flow of 1.2 cfm, assures that any release from the waste gas system to the plant vent will not exceed the site boundary dose rate limits of Part A Control C.7.l.1 .a. 5.2.3 Main Condenser Air Evacuation Monitor (RM-6505) The process radiation monitor on the main condenser air evacuation system provides operational information about the air being discharged. The discharge typically occurs either directly from the turbine building during start up (hogging mode) or through the plant vent during normal operations. -During maintenance activities or other temporary operational conditions, discharges to the turbine roof may also occur. This process monitor is also used as an indicator of potential releases from the Turbine Gland Seal Condenser exhaust. Early indications of a potential release (i.e., monitor count rate at twice the normal background) should be evaluated by collecting a grab sample of the exhausts from both the main condenser and the Turbine Gland Seal Condenser. The operational setpoints for the air evacuation monitor are administratively set as small multiples of the expected background response of the detector to provide operational control over unexpected changes in the activity discharged from the system. Typically, the alert setpoint is 1.5 times background, with the high alarm set at 2 times background. Maximum allowable setpoint determinations assure that the site boundary dose rate limits of Part A Control C.7.1.1.a will not be exceeded. For the air evacuation detector an efficiency of 1.87E + 08 cpm-cm 3 rnCi, (the AR-41 response value determined by HPSTID 20-021), flow rates of 10 to 50 cfm and 10,000 cfm for the normal and hogging modes of operation, respectively, and assuming that all the response is due to the most restrictive noble gas mixture associated with fuel gap activity inventory at the end of power operations (same mixture as used for the limiting mixture for the plant vent Wide Range Gas Monitor setpoint given in section 5.2.1.2), the following examples illustrate the calculation of the limiting setpoint for different operational conditions.

                                               -B.5-1 5                         ODCM Rev. 25

II Case 1: For start-up operations (i.e., 10,000 cfm hogging flow to the Turbine Building roof), the maximum allowable alarm setpoint is calculated as: RAE 147 I fg fglnd DFBc gfln where: RAE = Release rate equivalent to the assigned fraction of the limiting offsite total body dose rate (pCi/sec) 147 500 (mrem- i.Ci-m33 (I E + 06) (3.4E - 06) yr -pCi - sec 500 = The site boundary limiting total body dose rate (mrem/yr) from all release points lE+06 = Number of pCi per pCi (pCi/pCi) 3.4E-06 Maximum off-site long-term average gamma atmospheric dispersion factor for ground level releases (sec/m 2 ) DFBc = Composite total body dose factor (defined for the WRGM in Section 5.2.1.2 to be equal to 4.86E-03 [mrem-m 3 /pCi-yr] for the limiting fuel gap activity mix) fg = The fraction of the site boundary total body dose rate limit to be administratively assigned to monitored ground level releases (for this illustration = 0.3) such that the combination of the plant vent fraction (fQ) and ground fraction (fg) is less than or equal to 1 (fg < 1 - fQ). fgland = Release reduction factor to be administratively assigned to account for potential unmonitored contributions from the Turbine Gland Seal Condenser exhaust (for this illustration = 0.7). RAE 147 1 03(07 4.86E - 03 (0) (0)

            =     6.36E+03 IiCilsec release rate limit and for the 10,000 cefm (4.72E+06 cm 3 /sec) exhaust flow, the count rate response of the air evacuation monitor would be:

Monitor Response = RAE 1.87E + 08 cpm-cm3/jxCi 4.72E + 06cm 3 / sec

                        = (6.36E+03) (1.87E+08) / (4.72E+06)
                        = 2.520E + 05 cpm B.5-16                      ODCM Rev. 25

Case 2: As an extension of Case 1 which assumed the full startup hogging flow was released to the Turbine Building roof, maintenance requirements could direct normal operating main condenser offgas flow (assume 50 cfm or equivalent 2.36E+04 cm3/sec) to the Turbine Building Roof (ground level release point) instead of the elevated main plant vent. In this situation, the samne release rate limit as calculated above (i.e., 6.36E+03 piCi/sec) would apply. However, the reduced gas flow from 10,000 cfm down to 50 cfm would permit a higher alarm setpoint to be used. Monitor Response 23 AE e 1.87E + 08 cpm-cm 3/4Ci

              -                2.36E + 04cffi s'ec
                          =  (6.36E+03) (1.87E + 08) / (2.36E+04)
                          =  5.04E + 07 cpm Case 3:    For normal operations which direct main condenser offgas flow (assume 50 cfm or equivalent 2.36E+04 cm3/sec) to be released to the atmosphere via the main plant vent, the maximum allowable alarm setpoint would be:

RAE =588 f, fgud DFM where: RAE = Release rate equivalent to the assigned fraction of the limiting offsite total body dose rate (pCi/sec) 588 = 500 (mrem -juCi -M3 (1E + 06) (8.5E - 07) yr - pCi -sec 5.8E-07 = Maximum off-site long-term average gamma atmospheric dispersion factor for elevated (mixed mode) releases (sec/M2 ) DFBc = Composite total body dose factor (defined for the WRGM in Section 5.2.1.2 to be equal to 4.86E-03 [mrem-m3 /pCi-yr] for the limiting fuel gap activity mix) fgband = Same as listed above (i.e., 0.7) f = The fraction of the site boundary total body dose rate limit to be administratively assigned to plant vent releases such that the combination of the plant vent fraction (f ) and ground fraction (fg) is less than or equal to 1 (f, < I - fg). For the case that main condenser offgas is discharged to the main plant vent, there is no ground release fraction to be assigned (i.e., fg = 0), and f, maybe set at 1. RAE = 588 0.7 4.86E - 03

         =      8.47E+04 pCi/sec release rate limit
                                          -B.5-17                        ODCM Rev. 25

I I and for the 50 cfm (2.36E+06 cm3/sec) Main Condenser offgas exhaust flow, the count rate response of the air evacuation monitor would be: Monitor Response = RAE 1.87E + 08 cpm-cm3 4/Ci 2.36E + 06cm', sec

                        =  (8.47E+04) (1.87E + 08) ! (2.36E+04)
                        =  6.71E+08cpm The operation of the Main Condenser Evacuation System assumes 670 lbs./hour of steam flow through the Turbine Gland Seal Condenser exhaust (very small fraction of total steam flow), 1.5E+07 lbs./hour steam flow to the main condenser, and that the Turbine Gland Seal Condenser exhaust mostly air at a flow rate of 1,800 cfm which goes directly to the Turbine Building Vents (does not pass RM-6505). The main condenser offgas which goes past the Air Evacuation monitor during power operations is combined with other plant ventilation and process gas streams before being monitored by the WRGM and discharged to the atmosphere via the Plant Vent as a single release point.

The maximum allowable setpoints during startup and normal power operations may be recalculated based on identified changes in detector efficiency, discharge flow rate, radionuclide mix distribution, or administrative apportionment of potential contributions from the plant vent and ground level release points following the methods identified in Part B, Section 8.5. B.5-18 ODCM Rev. 25

6.0 LIQUID AND GASEOUS EFFLUENT STREAMS, RADIATION MONITORS AND RADWASTE TREATMENT SYSTEMS. Figure B.6-1 shows the liquid effluenft streams, radiation monitors and the appropriate Liquid Radwaste Treatment System. Figure B.6-2 shows the gaseous effluent streams, radiation monitors and the appropriate Gaseous Radwaste Treatment System. For more detailed information concerning the above, refer to the Seabrook Station Final Safety Analysis Report, Sections 11.2 (Liquid Waste System), 11.3 (Gaseous Waste System) and 11.5 (Process and Effluent Radiological Monitoring and Sampling System). The turbine gland seal condenser exhaust iodine and particulate gaseous releases will be determined by continuously sampling the turbine gland seal condenser exhaust. The noble gas releases will be determined by periodic noble gas grab samples. A ratio of main condenser air evacuation exhaust and turbine gland seal condenser exhaust noble gas will be determined periodically. B.-6-1 ODCM Rev. 25

I L

1. Figure B.6-1 Liquid Effluent Streams, Radiation Monitors, and Radwaste Treatment System at Seabrook Station C93-__

          /     CCLT-2172-1     \

CCLT-2172-2

          \RM-6515, RM-6516 O   CVCS Letdown Diversion Equipment Drainage                          WU Tritium Control Release Equipment Leakage (O) PAB Floor Drains h Secondary Side Steam
                                              ~   Generator Blowdown Non-Recyclable and Misc Containment Sumps Laboratory Drains                             Service Water System Decontamination Water CCLT Level Transmitter OD Turbine Building Sump                      RM Radiation Monitor OOCM61 B.6-2                             ODCM Rev. 25

Figure B.6-2 Gaseous Effluent Streams, Radiation Monitors, and Radwaste Treatment System at Seabrook Station Turbine Gland (During Hobbing Seal Condenser Exhaust Mode Only) LEGEND: H - HEPA Filter C - Charcoal Filter ODCM62 RM- Radiation Monitor B.6-3 ODCM Rev. 25

7.0 BASES FOR DOSE CALCULATION METHODS @ 7.1 Liquid Release Dose Calculations This section serves: (1) to document the development and conservative nature of Method I: equations to provide background information to Method I users, and (2) to identify the general equations, parameters and approaches to Method 11-type dose assessments. Appendix C provides the bases for the EMS software which is used to implement the dose and dose rate calculations indicated as Method IA.: Method I may be used to show that the Part A RECP which limit off-site total body dose from liquids (C.6.2.1 and C.6.3.1) have been met for releases over the appropriate periods. The quarterly and annual dose limits in Part A Control C.6.2.1 are based on the ALARA design objectives in IOCFR50, Appendix I Subsection HI A. The minimum dose values noted in Part A Control C.6.3.1 are "appropriate fractions," as determined by the NRC, of the design objective to ensure that radwaste equipment is'used as required to keep off-site doses ALARA. Method I was developed such that "the actual exposure of an individual ... is unlikely to be substantially underestimated" (IOCFR50, Appendix I). The definition, below, of a single "critical receptor" (a hypothetical or real individual whose behavior results in a maximum potential dose) provides part of the conservative margin to the calculation of total body dose in

     *Method I. Method II all6ws that act ual individuals, associated with identifiable exposure pathways, be taken into account for any given release. In fact, Method I was based on a Method II analysis for a critical receptor assuming all principal pathways present instead of any real individual. That analysis was called the "base case;" it was then reduced to form Method I. The general equations used in the base case analysis are also used as the starting point in Method II evaluations. Th7e base case, the method of reduction, and the assumptions and data used are presented below.

The steps performed in the Method I derivation follow. First, the dose impact to the critical receptor [in the form of dose factors DFIj, (mrem/pCi)] for a unit activity release of each radioisotope in liquid effluents was derived. The base case analysis uses the general equations, methods, data and assumptions in Regulatory Guide 1.109 (Equations A-3 and A-7, Reference A). The liquid pathways contributing to an individual dose are due to consumption of fish and invertebrates, shoreline activities, and 'winming and boating near the discharge point. -A nominal operating piant discharge flow rate of 918 fR3 /sec was used with a mixing ratio of 0.10. The mixing ratio of 0.10 corresponds to the miinimum expected prompt dilution or near-field mixing zone created at the oc~ean surface'directly above the inultiport diffusers. (Credit for additional dilution to the outer edge of thepronpt mixing zone which corresponds to the IOF surface isotherm (mixing ratio .025) can be applied in the'Method II calculation for shoreline exposures only since the edge of this isothermi typically does not reach the shoreline receptor points during the tidal cycle. The mixing ratio for aquatic food pathways in Method II assessments shall be limited to the same value (0.10) as applied in Method I for near-field mixing, or prompt dilution only. B.7 ODCM Rev. 24

I L The requirements for the determination of radiological impacts resulting from releases in liquid effluents is derived from IOCFR50, Appendix I. Section I.A.2 of Appendix I indicates that in making the assessment of doses to hypothetical receptors, "The Applicant may take account of any real phenomenon or factors actually affecting the estimate of radiation exposure, including the characteristics of the plant, modes of discharge of radioactive materials, physical processes tending to attenuate the quantity of radioactive material to which an individual would be exposed, and the effects of averaging exposures over time during which determining factors may fluctuate." In accessing the liquid exposure pathways that characterize Seabrook Station, the design and physical location of the Circulating Water Discharge System needs to be considered within the scope of Appendix L Seabrook utilizes an offshore submerged multiport diffuser discharger for rapid dissipation and mixing of thermal effluents in the ocean environment. The 22-port diffuser section of the Discharge System is located in approximately 50 to 60 feet of water with each nozzle 7 to 10 feet above the sea floor. Water is discharged in a generally eastward direction away from the shoreline through the multiport diffuser, beginning at a location over one mile due east of Hampton Harbor inlet. This arrangement effectively prevents the discharge plume (at least to the 1 degree or 40 to 1 dilution isopleth) from impacting the shoreline over the tidal cycle. Eleven riser shafts with two diffuser nozzles each form the diffuser and are spaced about 100 feet apart over a distance of about 1,000 feet. The diffusers are designed to maintain a high exit velocity of about 7.5 feet per second during power operations. Each nozzle is angled approximately 20 degrees up from the horizontal plane to prevent bottom scour. These high velocity jets passively entrain about ten volumes of fresh ocean water into the near field jet mixing region before the plume reaches the water surface. This factor of 10 mixing occurs in a very narrow zone of less than 300 feet from the diffuser by the time the thermally buoyant plume reaches the ocean surface. This high rate of dilution occurs within about 70 seconds of discharge from the diffuser nozzles. The design of the multiport diffuser to achieve a 10 to 1 dilution in the near field jet plume, and a 40 to 1 dilution in the near mixing zone associated with the 1 degree isotherm, has been verified by physical model tests (reference "Hydrothermal Studies of Bifurcated Diffuser Nozzles and Thermal Backwashing - Seabrook Station," Alden Research Laboratories, July 1977). During shutdown periods, when the plant only requires service water cooling flow, the high velocity jet mixing created by the normal circulating water flow at the diffuser nozzles is reduced. However, mixing within the discharge tunnel water volume is significantly increased (factor of about 5) due to the long transit time (approximately 50 hours) for batch waste discharged from the plant to travel the three miles through the 19-foot diameter tunnels to the diffuser nozzles. Additional mixing of the thermally buoyant effluent in the near field mixing zone assures that an equivalent overall 10 to 1 dilution occurs by the time the plume reaches the ocean surface. B.7-2 ODCM Rev. 24

The dose assessment models utilized in the'ODCM are taken from NRC Regulatory Guide 1.109. The liquid pathway equations include'a parameter (Mp) to account for the mixing ratio. (reciprocal of the dilution factor) of effluents in the environment at the point of exposure. Table 1, in Regulatory Guide 1.109, defines the point of exposure to be the location that is anticipated to be occupied during plant lifetime, or have potential land and water usage and food pathways as could actually exist during the term of plant operation. For Seabrook, the potable water and land irrigation pathways do not exist since saltwater is used as the receiving water body for the circulating water discharge. The three pathways that have been factored into the assessment models are shoreline exposures, ingestion of invertebrates, and fish ingestion. With respect to shoreline exposures, both the mixing ratios of 0.1 and 0.025 are extremely conservative since the effluent plume which is discharged over one mile offshore never reaches the beach where this type of exposure could occur. Similarly, bottom dwelling invertebrates, either taken from mud flats near the shoreline or from the area of diffuser, are not exposed to the undiluted effluent plume. The shore area is beyond the reach of the surface plume of the discharge, and the design of the upward directed discharge nozzles along with the thermal buoyancy of the effluent, force the plume to quickly rise to the surface without affecting bottom organisms. Consequentially, the only assumed exposure pathway which might be impacted by the near field plume of the circulating water discharge is finfish. However, the mixing ratio of 0.1 is very conservative because fish will avoid both the high exit velocity provided by the discharge nozzles and the high thermal temperature difference between the water discharged from the diffuser and the ambient water temperature in the near field. In addition, the dilution factor of 10 is achieved 'within 70 seconds of discharge and confined to'a very small area, thus prohibiting any significant quantity of fish from reaching equilibrium conditions with radioactivity concentrations created in the water environment. The mixing ratio of 0.025, which corresponds to the 1 degree thermal near field mixing zone, is a more realistic assessment of the dilution to which finfish might be exposed. However, even this dilution credit is conservative since it neglects the plant's operational design which discharges radioactivity by batch mode. Batch discharges are on the order of only a'few hours in duration several times per week and, thus, the maximum discharge concentrations are not maintained in the environment long enough io allow fish to reach equilibrium uptake concentrations as assumed in the dose assessment modeling. Not withstanding the above expected dilution credit afforded at the 1 degree isotherm, all Method II aquatic food pathway dose calculations shall conservatively assume credit for prompt diluti6n only with an' M 0.10. When dose' impacts from the fish and invertebrate pathways are then added to the conservative dose impacts'derived for shoreline exposures, the total calculated dose is very unlikely to have underestimated the exposure to any real individual. The recommended value for dilution of 1.0 given in NUREG-0133 is a simplistic'assurmption provided so that a single model could be used with any plant design and physical discharge arrangement. For plants that utilize a surface canal-type discharge structure where little entrainment mixing in the environment occurs, a dilution factor of 1.0 is a reasonable assumption. However, in keeping with the guidance provided in Appendix I to I OCFR5O, Seabrook has determine site-specific mixing ratios which factor in its plant design. B.7-3 ODCM Rev. 24

II The transit time used for the aquatic food pathway was 24 hours, and for shoreline activity 0.0 hours. Table B.7-1 outlines the human consumption and use factors used in the analysis. The resulting, site-specific, total body dose factors appear in Table B.1-11. Appendix A provides an example of the development of a Method I liquid dose conversion factor for site-specific conditions at Seabrook. 7.1.1 Dose to the Total Body For any liquid release, during any period, the increment in total body dose from radionuclide "i" is: ADd, = kQj DFLit (mrem) () (uCi)( (7-1) where: DFLI, = Site-specific total body dose factor (mrem/,iCi) for a liquid release. It is the highest ofthe four age groups. See Table B.l-ll. Q= Total activity (jtCi) released for radionuclide "i". k = 918/Fd (dimensionless); where Fd is the average dilution flow of the Circulating Water System at the point of discharge from the multiport diffuser (in ft3/sec). Method I is more conservative than Method II in the region of the Part A dose limits because the dose factors DFL,,, used in Method I were chosen for the base case to be the highest of the four age groups (adult, teen, child and infant) for that radionuclide. In effect each radionuclide is conservatively represented by its own critical age group. 7.1.2 Dose to the Critical Organ The methods to calculate maximum organ dose parallel to the total body dose methods (see Part B, Section 7.1.1). For each radionuclide, a dose factor (mnrem/pCi) was determined for each of seven organs and four age groups. The largest of these was chosen to be the maximum organ dose factor (DFLi,) for that radionuclide. DFI,. also includes the external dose contribution to the critical organ. For any liquid release, during any period, the increment in dose from radionuclide "i" to the maximum organ is: ADMO = k Q; DFLhW, (mrem) () (,uCi) (r (7-2) B.74 ODCM Rev. 24

where: DFLITro = Site-specific maximum organ dose factor (mrem/piCi) for a liquid release. See Table B.1-1 1. Q= Total activity (gCi) released for radionuclide "i". k = 91 8/Fd (dimensionless); where Fd is the average dilution flow of the Circulating Water System at the point of discharge from the multiport diffuser (in ft3 /sec). B.7-5 ODCM Rev. 24

Table B.7-1 Usage Factors for Various Liquid Pathways at Seabrook Station (From Reference A, Table E-5*, except as noted. Zero where no pathway exists) AGE VEG. LEAFY MILK MEAT FISH INVERT. POTABLE SHORELINE SWIMMING"* BOATING"* VEG. WATER (KGJYR) (KG/YR) (LITERIYR) (KG/YR) (KGIYR) (KG/YR) (LITERIYR) (HR/YR) (HR/YR) (HR/YR) Adult 0.00 0.00 0.00 0.00 21.00 5.00 0.00 334.00*** 8.00 52.00 Teen 0.00 0.00 0.00 0.00 16.00 3.80 0.00 67.00 45.00 52.00 Child 0.00 0.00 0.00 0.00 6.90 1.70 0.00 14.00 28.00 29.00 Infant 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00

• Regulatory Guide 1.109.
  • HERMES; "A Digital Computer Code for Estimating Regional Radiological Effects from Nuclear Power Industry," HEDL, December 1971. Note, for Method II analyses, these pathways need not be evaluated since they represent only a small fraction of the total dose contribution associated with the other pathways.
    • Regional shoreline use associated with mudflats - Maine Yankee Atomic Power Station Environmental Report.

B.7-6 ODCM Rev. 24

7.2 Gaseous Release Dose Calculations 7.2.1 Total Body Dose Rate From Noble Gases This section serves: (1) to document the development of the Method I equation, (2) to provide background information to Method I users, and (3) to identify the general equations, parameters and approaches to Method fl-type dose rate assessments. Method I may be used to show that the Part A Controls which limit total body dose rate from noble gases released to the atmosphere (Part A Control C.7.1.1) has been met for the peak noble gas release rate. Method I was derived from general equation B-8 in Regulatory Guide 1.109 as follows: yb=lE+06[X/Q]r A QDFB,  :(7-3) nrem) (pCi ) (sec;Hpi ' (mrem-m3' t yr ) tjuCi) m 3 Asec pCi-yr) where: [X/Q]7 = Maximum off-site receptor location long-term average gamma atmospheric dispersion factor. Q

                  =     Release rate to the environment of noble gas "i" (giCi/sec).
                                                   ;,        ,     3 DFBi        =     Gamma total body dose factor, I        Ci m ). See Table B.1-10. (Regulatory
                                                        .. pCi-yr)

Guide 1.109, Table B-i). Elevated and ground level gaseous effluent release points are addressed separately through the use of specific [X/Q)r For an elevated gaseous effluent release point and off-site receptor, Equation 7-3 takes the form: D be) =(IE+ 06)* (8.5E- 07) *E ( *DFB;) rnrem) (_pCi pC i , rrem-rn yrT 1UuCi) Ltm') sec:. pCi-yr ) B.7-7, ODCM Rev. 24

I I which reduces to: Dfb)o = 0.85 * (Qi

• DFB;) (3-3a)

(mrem =(pCi- sec7(/C ')*(mrem-m 3 yr pCi _M3 sec pCi- yr ) For a ground level gaseous effluent release point and off-site receptor, Equation 7-3 takes the form: D)bwg = (1 E+ 06) * (3.4 E- 06) * (Q.* DFB1 ) which reduces to:

              = 3.4*(*DFB)                                                                           (3-3b)

(mrem (pCi- secf (,uCi *(mrem- m 3 ) yr ) ,sec pCi- yr ) The selection of critical receptor, outlined in Part B, Section 7.3 is inherent in the derived Method I, since the maximum expected off-site long-term average atmospheric dispersion factor is used. The sum of doses from both plant vent stack and ground level releases must be considered for determination of Technical Specification compliance. All noble gases in Table B.1-10 should be considered. A Method II analysis could include the use of actual concurrent meteorology to assess the dose rates as the result of a specific release. 7.2.2 Skin Dose Rate from Noble Gases This section serves: (1) to document the development of the Method I equation, (2) to provide background information to Method I users, and (3) to identify the general equations parameters and approaches to Method 11-type dose rate assessments. The methods to calculate skin dose rate parallel the total body dose rate methods in Part B, Section 7.2.1. Only the differences are presented here. Method I may be used to show that the Part A Controls which limit skin dose rate from noble gases released to the atmosphere (Part A Control C.7. 1.1) has been met for the peak noble gas release rate. The annual skin dose limit is 3,000 mrem (from NBS Handbook 69, Reference D, pages 5 and 6, is 30 rem/10). The factor of 10 reduction is to account for nonoccupational dose limits. B.7-8 ODCM Rev. 24

It is the skin dose commitment to the critical, or most limiting, off-site receptor assuming long-term site average meteorology and that the release rate reading remains constant over the entire year. Method I was derived from the general equation B-9 in Regulatory Guide 1.109 as follows: Ds=1.1IDA' +3.17E+04 O),[X/Q]DFS1 (7-4) (mrem) (mrem (mrad)(pCi-yr) Ci (sec) (mrem-m 3 yr ) mrad myra (DCi- see) 3) pCi-yr ) where: 1.11 Average ratio of tissue to air absorption coefficients (will convert mrad in air to mrem in tissue). DFSi = Beta skin dose factor for a semi-infinite cloud of radionuclide "il which includes the attenuation by the outer "dead" layer of the skin. Dfr = 3.17E+04 z Q1 [X/OJDF[ (7-5) (mrad)(pci-yr Ci ( sec mrad- i-) yr Cl- sIn m3 pCi-yr DFj = Gamma air dose factor for a uniform semi-infinite cloud of radionuclide "i". Now it is assumed for the definition of (XQ7) from Reference 8 that: D&A.D.=Dr Q[X/Qr/[X/Q] (7-6) (mrad) (mrad)( ) rm3in) and Q = 31.54 Q (7-7) (Ci) ( Ci-sec) (,uCi) yr) (pi ~-yr) se~c B.7-9!! ODCM Rev. 24

IL so: Dskm =1 11 IE+06 [NQJ1 O *DFr (7-8) mrer = (mrem) 2(sec ( pCi FCi)(mrad -m3) yrJ -'mradJ ,uiJP in' sec pCi-yr /

                              +IE+06 XIQ Z jDFSj (pCi) (secJ (PuCi) mrem-m3n pCi) .mO )       sec ) pCi-yr           )

Substituting atmospheric dispersion factors for an elevated gaseous effluent release point, Equation 7-8 takes the following form:

                   =[1.11

• 1E+ 06
• 8.5 E- 07 *(Q*DF)] + [ E+ 06*8.2 E- 07* (Q
• DFS3 )]

i i which yields: Dsci,,) = [0.94

• DFi-)] + [0.82 j (Q;
• DFS;j) i i

( )(pCi- sec-. re Cimrem-m)+pCi-sec r___ rem-m3 yrC Ci- m' - mrad sec pCi- yr ,lCi-m `sec pCi- yr defining: DFixe) = 0.94 DFr + 0.82 DFS; (7-lOa) Then the off-site skin dose rate equation for an elevated gaseous effluent release point is: w= E Q;* DFice) (34a) ( mren4juCi

• mrem-secJ tr asec ,uCi- yr)

For an off-site receptor and a ground level gaseous effluent release point, Equation 7-8 becomes: D5 = [1 1 1I

• 1E+ 06
• 3.4E- 06
• Z(Oi
• DFir)] +[1 E+ 06
• 10 E- °05 (Qi
• DFS)]

B.7-10 ODCM Rev. 24

which yields: Dl,1n,',, = [3.8E (Oi

• DFI)]+ [ I10 (j
• DFSi)J (7-9b) i i

                      =Zj[3.8DFRY +ODFSi1 defining:

DF~iw = 3.8 DFir +10 DFSi (7-1 Ob) Then the off-site skin dose rate equation for ground level gaseous effluent release points is: Dk) =Q*DFyo) (3-4b) The selection of critical receptor, outlined in Part B, Section 7.3, is inherent in the derived Method I, as it is based on the determined maximum expected off-site atmospheric dispersion factors. All noble gases in Table B.1-10 must be considered.' 7.2.3 Critical Organ Dose Rate from I6dines, Tritium and Particulates With Half-Lives Greater Than Eight Days This section serves: (1) to document the development of the Method I equation, (2) to provide background information to Method I users, and (3) to identify the general equation's parameters and approached to Method II type dose rate assessments. The methods to calculate skin dose rate parallel the total body dose rate methods in Part B, Section 7.2.1. Method I may be used to show that the Part A Controls which limit organ dose rate from iodines, tritium and radionuclides in particulate form with half lives greater than 8 days released to the atmosphere (Part A Control C.7.1.1) has been met for the peak above-mentioned release rates. The annual organ dose limit is 1500 mrem (from NBS Handbook 69, Reference D, pages 5 and 6). It is evaluated by looking at the critical organ dose commitment to the most limiting off-site receptor assuming long-term site average meteorology. The equation for D . is derived from a form of Equation 3-8 in Part B, Section 3.9 by applying the conversion factor, 3.154E+07 (sec/yr) and converting Q to Q pCi/sec: (7-12)

             = 3.151E+07*         o(Q* DFGi,4 )

(mrem (secj (*UCiJ*( yr ) r ( Lsec )t#C B.7-11 ODCM Rev. 24

I I Equation 7-12 is rewritten in the form: DCo = ( ic *DFG') (mrem) (PCi)* pr i-y) (7-12a) yr ~ sec u Ci- yr where: (7-13) DFGkO =3.154 E+ 07 *DFGi( ) nurem- sec) Escl( em) u Ci- yr J r Ci The dose conversion factor, DFGi 4., has been developed for both elevated gaseous effluent release points and ground level gaseous effluent release points (DFGic ,e) and DFGc1 ), respectively. These dose factors are used to determine accumulated doses over extended periods and have been calculated with the Shielding Factor (SF) for ground plane exposure set equal to 0.7, as referenced in Regulatory Guide 1.109. In the case of the dose rate conversion factors (DFGIic,<c) and DFG'i,,(,)), the dose conversion factors from which they were derived were calculated with the Shielding Factor (SF) for ground plane exposure set equal to 1.0. For an off-site receptor and elevated effluent release point, the critical organ dose rate equation is: Dco(Z) = (Qi *DFG')) (3-5a) (mrem) =,uCi *nmrem-sec yr sec ,u Ci- yr For an off-site receptor and ground level effluent release point, the critical organ dose rate equation is: f~c ZQ*DFGV) (3-5b) Cirem yr ) _(,aCi 1 sec

• mrem- sec

                                   /JCi-yr   )

The selection of critical receptor, outlined in Part B, Section 7.3 is inherent in Method I, as are the expected atmospheric dispersion factors. B.7-12 ODCM Rev. 24

In accordance with the Basis Statement 3/4.11.2.1 in NUREG-0472, and the base's section for the organ dose rate limit given for Part A Control C.7.1.1 a Method II dose rate calculation, for compliance purposes, can be based on restricting the inhalation pathway to a child's thyroid to less than or equal to 1,500 mrem/yr. Concurrent meteorology with time of release may also be used to assess compliance for a Method II calculation. 7.2.4 Gamma Dose to Air from Noble Gases. This section serves: (1) to document the development and conservative nature of Method I equations to provide background information to Method I users, and (2) to identify the general equations, parameters and approaches to Method fl-type dose assessments. Method I may be used to show that the Part A Control C.7.2.1 which limits off-site gamma air dose from gaseous effluents has been met for releases over appropriate periods. This Part A Control is based on the objective in 10CFR50, Appendix I, Subsection B.1, which limits the estimated gamma air dose in off-site unrestricted areas. NUREG/CR-2919 presents a methodology for determining atmospheric dispersion factors (CHI/Q values) for intermittent releases at user specified receptor locations (intermittent releases being defined as releases with durations between 1 and 8,760 hours). The CHIIQ values for;' intermittent releases are determined by linearly interpolating (on a log-log basis) between an hourly 15-percentile CHI/Q value and an annual average CHI/Q value as a function of release duration. This methodology has been adopted to produce a set of time-dependent atmospheric dispersion factors for Method I calculations. For any noble gas release, in any period, the increment in dose is taken from Equations B4 and B-5 of Regulatory Guide 1.109 with the added assumption that Dfiinm = DT [X/Q]J /[X/Q]: AD") = 3.17E+4 [X/QY XQiDFr - (mrad) (pCi-yr) (sec)(Ci) mrad-m3 (7-14) Ci-sec m rn) pCi-yr where: 3.17E+04= Number of pCi per Ci divided by the number of seconds per year. [X/Q] 7 = Annual average gamma atmospheric dispersion factor for the receptor location of interest. Q = Number of curies of noble gas !'i" released. DFT3 = Gamma air dose factor for a uniform semi-infinite cloud of radionuclide "i". Incorporating a unitless release duration adjustment term t.' (where "a" is a constant and "t" is the total release duration in hours), and the conversion factor for Ci to pCi (to accommodate the use of a release rate Q in pCi), and substituting the 1-hour gamma atmospheric dispersion factor in place of the annual average gamma atmospheric dispersion factor in Equation 7-14 leads to: B.7-13 ODCM Rev. 24

I I Dr. 3.17E-02*[X/Qr

• t-a* (Q*DFr)

(mrad) -- ( (a c-

                         ~

Puci -sec f yr ) (m _ (sec) m 33 - (uci

• na-nJ(-6 mradM3) p(3 -6 (36 For an elevated release, the equation used for an off-site receptor is:

DXC) =3.17E-02*[l.OE-05]*t 4-01*7 ( D.

                                                                    *DFiT) which leads to:

Drip) =3.2E-07* t 7 * (Qi *Fjr)-D (3-6a)

                      -o

_ -Yr mrad-m3 (mrad) = (oCi m3J * (

• pCi-yr)

For a ground-level release, the equation used for an off-site receptor is: Dr =3.l7E-02*[4.9E-05]*t4293*F (Qj*DFY) which leads to: Dr =1.6E-06*t2 933 * (Q*DFY) (3-6b) (mrad pci-yr ) f . mrad--m' (mMpd) = ( YJ

• C
• p mra-r The major difference between Method I and Method 11 is that Method II would use actual or concurrent meteorology with a specific noble gas release spectrum to determine [X/Q]i rather than use the site's long-term average meteorological dispersion values.

7.2.5 Beta Dose to Air from Noble Gases This section serves: (1) to document the development and conservative nature of Method I equations to provide background information to Method I users, and (2) to identif~r the general equations, parameters and approaches to Method 11-type dose assessments. Method I may be used to show that Part A Control C.7.2.1, which limits off-site beta air dose from gaseous effluents, has been met for releases over appropriate periods. This Part A Control is based on the objective in IOCFR50, Appendix I, Subsection B.l, which limits the estimated beta air dose in off-site unrestricted area locations. For any noble gas release, in any period, the increment in dose is taken from Equations B-4 and B-5 of Regulatory Guide 1.109: ADZ = 3.17E-02*XQX (Q.

• DFr) (7-15)

(mrd) ( pCi-yr) * (secj (loi)

                                                       -DOM Cmrad  -m:

B.Vi pCi - Yr B.7-14 ODCM Rev. 24

where: DFfRP= Beta air dose factors for a uniform semi-infinite cloud of radionuclide "i". Incorporating the term ta into Equation 7-15 leads to: Dt = 3.17E-02*X/Qlh,

• t4 *(Q, *DP) (3-7) mrad-m 3 (mrad) =J (

                        ,uCi - sec (s*e) m

• Pci pCi - yrJ Where X/Qlh, = average 1-hour undepleted atmospheric dispersion factor.

For an elevated release, the equation used for an off-site receptor is:

              - D'S3-17E02 3.1.

• 1.3E-05
• t43
• i (Q

(,

• DFf) mrad -m 3 (mrad) = (,pi-r) * (sec) * ( ) -(* Ci * -

pCi -yr) which leads to: Dai(C) = 4.IE-07

• t3*
• DF').

1(Q. (3-7a) (mrad) = (uPcirn ) ( ) * '(,-Ci .

• mrad-m pCi - yr For a ground-level release, the equation used for an off-site receptor is:

D6E = 3.17E-02*1.9E-I4*t-4 9 *V (Q1* DF_) (mrad) = (pCi-yr

                      ,uCS - -sec )
                                     * (sec) * ( )

• m CPci mrad-m3 pCi-yr )

B.7-15 ODCM Rev. 24

IL which leads to: D ) = 6.OE -06

• t-3"9
• E (Q,
• DF) (3-7b)

(mrad) (pCi PCI - m,

                                  *  ( ) *E   (rCi

• mrad-r pCi - yr J

7.2.6 Dose to Critical Organ from lodines. Tritium and Particulates with Half-Lives Greater Than Eight Days This section serves: (1) to document the development and conservative nature of Method I equations to provide background information to Method I users, and (2) to identify the general equations, parameters and approaches to Method 11-type dose assessments. Method I may be used to show that the Part A Controls which limit off-site organ dose from gases (C.7.3.1 and C.8.1.1) have been met for releases over the appropriate periods. Part A Control C.7.3.1 is based on the ALARA objectives in IOCFR50, Appendix L Subsection II C. Part A Control C.8.1.I is based on Environmental Standards for Uranium Fuel Cycle in 40CFRI90, which applies to direct radiation as well as liquid and gaseous effluents. These methods apply only to iodine, tritium, and particulates in gaseous effluent contribution. Method I was developed such that "the actual exposure of an individual ... is unlikely to be substantially underestimated" (10CFR50, Appendix 1). The use below of a single "critical receptor" provides part of the conservative margin to the calculation of critical organ dose in Method I. Method II allows that actual individuals, associated with identifiable exposure pathways, be taken into account for any given release. In fact, Method I was based on a Method II analysis of a critical receptor assuming all pathways present. That analysis was called the "base case"; it was then reduced to form Method I. The base case, the method of reduction, and the assumptions and data used are presented below. B.7-16 ODCM Rev. 24

The steps performed in the Method I derivation follow. First, the dose impact to the critical receptor [in the form of dose factors DFGjCO (mrem ljCi)] for a unit activity release of each iodine, tritium, and particulate radionuclide with half lives greater than eight days to gaseous effluents was derived. Six exposure pathways (ground plane, inhalation, stored vegetables, leafy vegetables, milk, and meat ingestion) were assumed to exist at the site boundary (not over water or marsh areas) which exhibited the highest long-term X/Q. Doses were then calculated to six organs (bone, liver, kidney, lung, GI-LLI, and thyroid), as well as for the whole body and skin for four age groups (adult, teenager, child, and infant) due to the seven combined exposure pathways. For each radionuclide, the highest dose per unit activity release for any organ (or whole body) and age group was then selected to become the Method I site-specific dose factors. The base case, or Method I analysis, uses the general equations methods, data, and assumptions in Regulatory Guide'1.109 (Equation C-2 for doses resulting from direct exposure to contaminated'ground plane; Equationi C-4 for doses associated with inhalation of all radionuclides to different organs of individuals of different age groups; and Equation C-13 for doses to organs of individuals in different age groups resulting from ingestion of radionuclides in produce, milk, meat, and leafy vegetables in Reference A). Tables B.7-2 and B.7-3 outline human consumption and environmental parameters used in the analysis. 'It is conservatively assumed that the critical receptor lives at the "maximum off-site atmospheric dispersion factor location" as defined in Section 7.3. The resulting site-specific dose factors are for the maximum organ which combine the limiting age group with the highest dose factor for any organ with each nuclide. These critical organ, critical age dose factors are given in Table B.1-12. Appendix A provides an example of the development of Method I gaseous dose conversion factor for site-specific conditions at Seabrook. For any iodine, tritium, and particulate gas release, during any period, the increment in dose from radionuclide Wi" is:

      -A=Dco      QiDFGico,                                                                          (7-16) where DFGicO is the critical dose factor for radionudide "i" and Qi is the activity of radionuclide "i" released in rmicrocuries.

Applying this information, it follows that the general form for the critical organ dose equation is: co= (X/Q)dCP /(X/Q)dCPp * ** (Qi

• DEG) (3-8)

                -sec)se                              mrem mrein(          3       {(    ) *(PCi*-

( A  : _. * . _ Substituting specific values associated with the maximum off-site receptor location and elevated release' condition yields:' Dco*)= (1.12E- 05)/(7.55E- 07)

• t0 29' (Qi *DFGic.(,))

which reduces to: Dco(e) = 14.8

• t 0.297
• Z (Qi
• DFGico*e)) (3-8a)

B.7-17 ODCM Rev. 24

Il For the maximum off-site receptor location and ground-level release conditions, the equation is: DCOW = (1.71 E- 04)/(9.64 E- 06)

• t * (Q
• DFGi;)

which reduces to: Do = 17.7

• t-3'6
• E (Qj
• DFG d) (3-8b) 7.2.7 Special Receptor Gaseous Release Dose Calculations Part A Section 10.2 requires that the doses to individuals involved in recreational activities within the site boundary are to be determined and reported in the Annual Radioactive Effluent Release Report.

The gaseous dose calculations for the special receptors parallel the bases of the gaseous dose rates and doses in Part B, Sections 7.2.1 through 7.2.5. Only the differences are presented here. The special receptor XQs are given in Table B.7-5. 7.2.7.1 Total Body Dose Rate from Noble Gases Method I was derived from Regulatory Guide 1.109 as follows: Do =1 E+ 06 [X/Q ]r DFBj (7-3) General Equation (7-3) is then multiplied by an Occupancy Factor (OF) to account for the time an individual will be at the on-site receptor locations during the year. There are two special receptor locations on-site. The "Rocks" is a boat landing area which provides access to Browns River and Hampton Harbor. The Seabrook Station UFSAR, Chapter 2.1, indicates little boating activity in either Browns River or nearby Hunts Island Creek has been observed upon which to determine maximum or conservative usage factors for this on-site shoreline location. As a result, a default value for shoreline activity as provided in Regulatory Guide 1.109, Table E-5, for maximum individuals was utilized for determining the "Rocks" occupancy factor. The 67 hours/year corresponds to the usage factor for a teenager involved in shoreline recreation. This is the highest usage factor of all four age groups listed in Regulatory Guide 1.109, and has been used in the ODCM to reflect the maximum usage level irrespective of age. Regulatory Guide 1.109 does not provide a maximum individual usage factor for activities similar to those which would be associated with the Seabrook Station Science & Nature Center. Therefore, the usage factor used in the ODCM for the Science & Nature Center reflects the observed usage patterns of visitors to the facility. Individuals in the public who walk in to look at the exhibits on display and pick up available information stay approximately 1.5 hours each. Tour groups who schedule visits to the facility stay approximately 2.5 hours. For conservatism, it was assumed that an individual in a tour group would return five times in a year, and stay 2.5 hours on each visit. These assumptions, when multiplied together, provide the occupancy factor of 12.5 hours/year used in the ODCM for public activities associated with the Science & Nature Center. For the Science & Nature Center, and the "Rocks", the occupancy factors (OFs) are: B.7-18 ODCM Rev. 24

Science & Nature Center 12-5 1hs/yr -0.0014 8760 hrs/yr 67 hrs/ yr~'" The "Rocks" - 8760 __At = 0.0076 8760 hrs/yr substituting in the annual average gamma XIQs: [X/Q] 7 1.IE-06 sec/M3 (Science & Nature Center) for primary vent stack releases.

               = 5.3E-06 sec/M3 (Science & Nature Center) for ground level releases.
               = 5.OE-06 sec/M3 (The "Rocks") for primary vent stack releases.
               = 2.6E-05 sec/m 3 (The "Rocks") for ground level releases.

and multiplying by: OF= 0.0014 (Science &Nature Center)

           = 0.0076 (The "Rocks")

gives: DtbE(,) = 0-0015 *X (Q.. *DFB;) (nrem/yr) (3-3c) DbbE(Z)= 0-0074*E(QO *DFBi) (mreni/yr) (3-3d) DeR(,) = 0.038

• E (Q
• DFBj) (mremIyr) (3-3e)

Dfpw,) = 0.2

• I (Qj
• DFB;) (mrem/yr) (3-3f) where:

bE f,),bES),ID, bRp,), and DbR= total body dose rates to an individual at the

                                                             'Science & Nature Center and the "Rocks" (recreational site), respectively, due to noble
                                               ; . igases            in an elevated (e) and ground level (g) release,

') Taken from Seabrook Station Technical Specificationfs (Figure 5.1-1). B.7-19 ODCM Rev. 24

I I Q and DFBi are as defined previously. 7.2.7.2 Skin Dose Rate from Noble Gases Method I was derived from Equation (7-8): I.11 E+06 [xIQlY 1.= Q.DFY + (7-8) IE+06X/Q ZQIDFSI substituting in the annual average gamma XIQs: [X/Q]y = 1.IE-06 sec/m 3 (Science & Nature Center) for primary vent stack releases.

                = 5.3E-06 sec/r  3  (Science & Nature Center) for ground level release points.
                = 5.OE-06 sec/M3 (The "Rocks") for primary vent stack releases.
                = 2.6E-05 sec/M3 (The "Rocks") for ground level release points.

and the annual average undepleted X/Qs: X/Q = 1.6E-06 sec/m 3 (Science & Nature Center) for primary vent stack releases.

                 = 2.3E-05 sec/rn 3 (Science & Nature Center) for ground level release points.
                 = 1.7E-05 sec/m 3 (The "Rocks") for primary vent stack releases.
                 = 1.6E-04 sec/M3 (The "Rocks") for ground level release points.

and multiplying by: OF = 0.0014 (Science & Nature Center)

               = 0.0076 (The "Rocks")

gives: DsnE(,) = 0.0014 E DFr;1-22

                                              +1.60 DFSj] for an elevated release point.
                    = 0.0014  E Q [5.88 DFj;+ 23 DFSj] for a ground level release point.

DstixR0e) . 0076EQ; 1555 DFr + 17.0 DFSj1 for an elevated release point.

                    = 0.0076 E Q [28.9 DFi +160 DFS i] for a ground level release point.

B.7-20 ODCM Rev. 24

and the equations can be written: Dbki,..o) =0.0014 *z(Qi* DFIe)) (3-4c) Dfs;,Rn,) = 0.0076 * (* DFrn,)) (3-4e) f = 0.0076*E (* DFiR) (34f) where: DSkiE(),,E(S), D skinRe) and D = the skin dose rate (mrem/yr) to an individual at the Science & Nature Center and the "Rocks", respectively, due to noble gases in an elevated (e) and ground level (g) release, defined previously, and DFjE,,DF,,,DFj(,), and DFjC) = the combined skin dose factors for radionuclide "i" for the Science & Nature Center and the "Rocks", respectively, for elevated (e) and ground level (g) release points (see Table B.1-13). 7.2.7.3 Critical Organ Dose Rate from Wodines. Tritium and Particulates with Half-Lives Greater Than Eight Davs The equations for DCO are derived in the same manner as in Part B, Section 7.2.2, except that the occupancy factors are also included. Therefore: D oE(e)= 0.0014* D (Q

• DFGLOE) )for an elevated release. (3-5c) bCEu =0.0014* j (O
• DFGE4) for a ground level release. (3-Sd)

DbOR(w)= 0.0076* (Q;* Q DFG.,,,C)) for an elevated release. (3-5e) i B.7-21 ODCM Rev. 24

IL where: Db Cc)DcoE )' DcOR(c), and DoR) = the critical organ dose rates (mrem/yr) to an individual at the Science & Nature Center and the "Rocks", respectively, due to iodine, tritium, and particulates in elevated (e) and ground level (g) releases, Qj = as defined previously, and DFG'Ee) ,DFG' ,DFG ,e),and DFG RX = the critical organ dose rate factors for radionuclide "i" for the Science & Nature Center and the "Rocks", respectively, for elevated (e) and ground level (g) release points (see Tables B.1-14 and B.l-15). 7.2.7.4 Gamma Dose to Air from Noble Gases Method I was derived from Equation (3-6): Dr = 3.17E- 02 * [XIQJ,

• t-
• Z (Q.
• DFi) (3-6) where all terms of the equation are as defined previously.

Incorporating the specific OF and the atmospheric dispersion factor, the gamma air dose equation for the Science & Nature Center for elevated releases: D e =3.17E-02*1.E-05t-t-2*0.0014*E(Qi*DFiy) which reduces to: Darur) =4.9E-10*tt02S2

• Z (Q.
• DFI) (3-6c)

(mrad) =- I I:* (,Ci

• ma-zICiM3 PEiyr For ground-level releases, the gamma air dose equation for the Science & Nature Center becomes:

DyMw = 3.17E- 02 *l.OE- 04 t0321*.0 *0 4*(Qj

• DF?)

B.7-22 ODCM Rev. 24

which reduces to: 32 DrE>)= 4.4 E- 09

• t 1
• E (Qj
• DFr) (3-6d)

(mrad) (-J) *(Pi

• r )

Incorporating the specific OF and atmospheric dispersion factors for the "Rocks" yields the gamma air dose equation for elevated releases: DairR(e) = 3.17E- 02 *2.1E-05* tflSS

• 0.0076* (Qj *DFr) which reduces to:

                    = 5.1 E- 09

• t4'ss
• E (Q
• DFr) (3-6e) rnrad)(V.3J*( "CiM3

                                          ) *E(a Ci* mrd- pCi_. yr   )

For ground-level releases, the gamma air dose equation for the "Rocks" becomes: Dr = 3.17 E- 02*1.7 E- 04 t 0 *0.076*E (Qj *DFir) which reduces to: Dl=4.1 E- 08 *X(Q2*DFr) (3-6f) (mrad) r (i-J( (

                                          *XP-*

ci *-mrad-m ra-m 7.2.7.5 Beta Dose to Air from Noble Gases Method I was derived as described in Pait B, Section 7.2.5. The general form of the dose equation is: D! =3.17E-02*X/Q-'*t *a (Qi*DFiP) (3-7) where all terms in the equation are as defined in Part B, Section 7.2.5. B.7-23 ODCM Rev. 24

I I Incorporating the specific OF and atmospheric dispersion factor for elevated releases into Equation 3-7 yields the following beta dose equation for the Science & Nature Center: Dpe) 3.17E- 02 *4.0E- 05i*Ete=s

• 0.0014 * (Q *DFO) which reduces to:

DP 1.8E-09*t43 *Z(Q*DFiO) (3-7c) (mrad) = (Piy (na)(U PC;- 3) (( )*F.(luCi* C mrJ pCi- yr) For ground-level releases, the beta air dose equation for the Science & Nature Center becomes: DarE}= 3.17E-02*5.5E-O4*t 4 3'7 *0.0014* (Qj*DFB) which reduces to: 7 DairE = 2.4 E- 08

• t-3'* (Qj
• DF16) (3-7d)

(mrad)( )*( *rad Incorporating the specific OF and atmospheric dispersion factors for the "Rocks" yields the beta air dose equation for elevated releases: DPR)=3.17E-O2*1.6E-04*t024 9 *0.0076* (Q;*DFjl) which reduces to: D- 3)

                  =39E-08*t02    49 *(Qj*DFtl)

(3-7e) (rnad -( yr3 * )* I(ci* mrad-rn3 (mrad) =( -C )() ( p i For ground-level releases, the beta air dose equation for the "Rocks" becomes: Da 3=3.17 E-02 *1.9E- 03

• t 0 267
• 0.0076* I (Q
• DF()

B.7-24 ODCM Rev. 24

which reduces to: D!YRat = 4.6 E- 07

• t 0267
• 2 (Q *DFf) (3-7f)

(mrad) - r Ci-v

( )*(

prai-yr

                                             /i3)

• mra-m 7.2.7.6 Critical Organ Dose from lodines. Tritium and Particulates With Half-Lives Greater Than Eight Days Method I was derived as described in Part B, Section 7.2.3. The Critical Organ Dose equations for receptors at the Science & Nature Center and the "Rocks" were derived from Equation 3-8. The following general equation incorporates (i) a ratio of the average 1-hour depleted atmospheric dispersion factor to the average annual depleted atmospheric dispersion factor, (ii) the unitless t"a term, and (iii) the OF:

             =      /Q)d/(X/Q)f'

• t* OF*E (Qj
• DFGkO)

                                 =-)* ( )*( )*             ;

Applying the Science & Nature Center-specific factors for elevated release conditions produces the equation: DcoE4) = (3.72 E- 05)1(1.56 E- 06)

• t'0349
• 0.0014* (Q
• DFGicoE(e))

which reduces to: DcOE() = 3.3E- 02* t49 * (Qj

• DFGiCOE(,)) (3-8c)

(mrem)=( )*( )* 2 ( Ci * - C ) For a ground-level release, the equation for a receptor at the Science & Nature Center is: DCgt) 2-=(5.21E- 04) /(2.23E- 05)

• 347
• 0.0014* E (Q,
• DFGiwEg))

which reduces to: D Ew= 3.3 E- 02

• t-0347 * > (Qj
• DFGi.Ew) (3-8d)

(mrem)=( )*( )* (Ci *j B.7-25 ODCM Rev. 24

I I The specific Critical Organ Dose equation for a receptor at the "Rocks" under elevated release conditions is: DcoR~c) = (1.54 E- 04)/(1.61 E- 05) *t0248

• 0.0076* (Q* DFG 1 coR(.))

which reduces to: DCOR(.) = 7.3E- 02

• t024 * (Qi
• DFGcoR(,)) (3-8e)

(mrem)=( )*

• P~Ci* mrer)

                                 -(        b Ci)

For a ground-level release, the equation for a receptor at the "Rocks" is: DOR(g) =(1.80E- 03)/(1.59 E- 04)

• t0267
• 0.0076* E (Qi
• DFGi. R(g) which reduces to:

D R., = 8.6E- 02

• t4 267* (Qi
• DFGi.R(s)) (3-8f)

(mrem)= *( ) ( Ci* mre) The special receptor equations can be applied under the following conditions (otherwise, justify Method I or consider Method I1):

1. Normal operations (nonemergency event).

2. Applicable radionuclide releases via the station vents to the atmosphere.

If Method I cannot be applied, or if the Method I dose exceeds this limit, or if a more refined calculation is required, then Method II may be applied. B.7-26 ODCM Rev. 24

Table B.7-2 Environmental Parameters for Gaseous Effluents at Seabrook Station (Derived from Reference A) Variable Vegetables Cow Milk Goat Milk Meat Stored Leafy Pasture Stored Pasture Stored Pasture Stored YV Agricultural Productivity (Kg/M2 ) 2. 2. 0.70 2. 0.70 2. 0.70 2 P Soil Surface Density (Kg/M2 ) 240. 240. 240. 240. 240. 240. 240. 240 T Transport Time to User (HRS) 48. 48. 48. 48. 480. 480 TB Soil Exposure Time ( (HRS) 131400. 131400. 131400. 131400. 131400. 131400. 131400. 131400. TE Crop Exposure Time to Plume (HRS) 1440. 1440. 720. 1440. 720. 1440. 720. 1440 TH Holdup After Harvest (HRS) 1440. 24. 0. 2160. 0. 2160. 0. 2160 QF Animals Daily Feed (Kg/DAY) 50. 50. 6. 6. 50. 50 FP Fraction of Year' on Pasture(2) 0.50 0.50 0.50 FS Fraction Pasture when on Pasture( 3 ) 1. 1. 1. FG Fraction of Stored Veg. Grown in Garden 0.76 - - -. FL Fraction of Leafy Veg. Grown in Garden 1.0 FI Fraction Elemental Iodine = 0.5 H Absolute Humidity 5.60(4.- _ ( Regulatory Guide 1.109, Rev. 1 B.7-27 ODCM Rev. 24

Table B.7-2 Environmental Parameters for Gaseous Effluents at Seabrook Station 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 hours (1 year) for all pathways. (2) For Method II dose/dose rate analyses performed for releases occurring during the first or fourth calendar quarters, the fraction of time animals are assumed to be on pasture is zero (nongrowing season). For the second and third calendar quarters, the fraction of time on e -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, Pergamnmon Press). B.7-28 ODCM Rev. 24

Table B.7-3 Usage Factors for Various Gaseous Pathways at Seabrook Station (from Reference A, Table E-5) Maximum Receptor: - I -; Age Leafy Group Vegetables Vegetables Milk Meat Inhalation (kg/yr) (kg/yr) (l/yr) (kg/yr) (m3 /yr) Adult 520.00 64.00 310.00 110.00 8000.00 Teen 630.00 42.00 400.00 65.00 8000.00 Child 520.00 26.00 330.00 41.00 3700.00 Infant 0.00 0.00 330.00 0.00 1400.00

       . k   "                         .ad The "Rocks" and Science & Nature Center:-

Age Leafy Group Vegetables Vegetables Milk Meat Inhalation (kgfyr) (kg/yr)i (l/yr) (kg/yr) (m3 /yr) Adult 0.00 0.00 0.00 0.00 8000.00 Teen . .0.00 0.00 0.00 0.00 8000.00 Child 0.00 0.00 0.00 0.00 .; r _. 3700.00 Infant 0.00 0.00 0.00 0.00 . . .1400.00 Regulatory Guide 1.109 B.7-29 . ODCM Rev. 24

I I 7.3 Receptor Points and Average Atmospheric Dispersion Factors for Important Exposure Pathways The gaseous effluent dose equations (Method I) have been simplified by assuming an individual whose behavior and living habits inevitably lead to a higher dose than anyone else. The following exposure pathways to gaseous effluents listed in Regulatory Guide 1.109 (Reference A) have been considered:

1. Direct exposure to contaminated air,

2. Direct exposure to contaminated ground;

3. Inhalation of air,

4. Ingestion of vegetables;

5. Ingestion of goat's milk; and

6. Ingestion of meat.

Part B, Section 7.3.1 details the selection of important off-site and on-site locations and receptors. Part B, Section 7.3.2 describes the atmospheric model used to convert meteorological data into atmospheric dispersion factors. Part B, Section 7.3.3 presents the maximum atmospheric dispersion factors calculated at each of the off-site receptor locations. 7.3.1 Recegtor Locations The most limiting site boundary location in which individuals are, or likely to be located as a place of residence was assumed to be the receptor for all the gaseous pathways considered. This provides a conservative estimate of the dose to an individual from existing and potential gaseous pathways for the Method I analysis. This point is the west sector, 974 meters from the center of the reactor units for undepleted, depleted, and gamma X/Q calculations, and the northwest section, 914 meters for calculations with D/Q the dispersion parameter. The site boundary in the NNE through SE sectors is located over tidal marsh (e.g., over water), and consequently are not used as locations for determining maximum off-site receptors (Reference NUREG 0133). Two other locations (on-site) were analyzed for direct ground plane exposure and inhalation only. They are the "Rocks" (recreational site) and the Education Center shown on Figure 5.1-1 of the Technical Specifications. 7.3.2 Seabrook Station Atmospheric Dispersion Model The time average atmospheric dispersion factors for use in both Method I and Method II are computed for routine releases using the AEOLUS-2 Computer Code (Reference B). B.7-30 ODCM Rev. 24

AEOLUS-2 produces the following average atmospheric dispersion factors for each location:

1. Undepleted X/Q dispersion factors for evaluating ground level concentrations of noble gases;

2. Depleted X/Q dispersion factors for evaluating ground level concentrations of iodines and particulates;

3. Gamma X/Q dispersion factors for evaluating gamma dose rates from a sector averaged finite noble gas cloud (multiple energy undepleted source); and

4. D/Q deposition factors for evaluating dry deposition of elemental radioiodines and other particulates.

Gamma dose rate is calculated throughout this ODCM using the finite cloud model presented in "Meteorology and Atomic Energy'- 1968" (Reference E, Section 7-5.2.5). That model is implemented through the definition of an effective gamma atmospheric dispersion factor, [X/Qi] (Reference B, Section 6), and the replacement of X/Q in infinite cloud dose equations by the [XIQr]. 7.3.3 Average Atmosnheric Dispersion Factors for Receptors The calculation of Method I and Method II atmospheric diffusion factors (undepleted CHII/Q, depleted CHI/Q, D/Q, and gamma CHI/Q values) utilize a methodology' generally consistent with US NRC Regulatory Guide 1.111 (Revision 1) criteria and the methodology for calculating routine release diffusion factors as represented by the XOQDOQ computer code (NUREG/CR-2919). The primary vent stack is treated as a "mixed-mode" release, as defined in Regulatory Guide 1.1 11. Effluents are considered to be part-time ground level/part-time elevated releases depending on the ratio of the primary vent stack effluent exit velocity relative to the speed of the prevailing wind. All other release points (e.g., Turbine Building and Chemistry lab hoods) are considered ground-level releases. In addition, Regulatory Guide 1.111 discusses the concept that constant mean wind direction models like AEOLUS-2 do not describe spatial and temporal variations in airflow such as the recirculation of airflow which can occur during prolonged periods of atmospheric stagnation. For sites near large bodies of water like Seabrook, the onset and decay of sea breezes can also result in airflow reversals and curved trajectories. Consequently, Regulatory Guide 1.111 states that adjustments to constant mean wind direction model outputs may be necessary to account for such spatial and temporal variations in air flow trajectories. Recirculation correction factors have been applied to the diffusion factors. The recirculation correction factors used are compatible to the "default open terrain" recirculation correction factors used by the XOQDOQ computer code. The relative deposition rates, D/Q values, were derived using the relative deposition rate curves presented in Regulatory Guide 1.111 (Revision 1). These curves provide estimates of deposition rates as a function of plume height, stability class, and plume travel distance. B.7-31 ODCM Rev. 24

I I Receptor Locations For ground-level releases, the downwind location of "The Rocks" (244m NE/ENE) and the Science & Nature Center (406m SW) were taken as the distance from the nearest point on the Unit 1 Administrative Building/Turbine Building complex. For the site boundary, the minimum distances from the nearest point on the Administration Building/Tuibine Building complex to the site boundary within a 45-degree sector centered on the compass direction of interest as measured from UFSAR Figure 2.1-4A were used (with the exception that the NE-NE-ENE-E-ESE-SE site boundary sectors were not evaluated because of their over-water locations). For primary vent stack releases, the distances from the Unit I primary vent stack to "The Rocks" (244m NE) and the Science & Nature Center (488m SW) as measured from a recent site aerial photograph were used. For the site boundary, the minimum distances from the Unit 1 primary vent stack to the site boundary within a 45-degree sector centered on the compass direction of interest as measured from UFSAR Figure 2.14A were used (with the exception that the NNE-NE-ENE-E-ESE-SE site boundary sectors were not evaluated because of their over-water locations). Meteorological Data Bases For "The Rocks' and Science & Nature Center receptors, the diffusion factors represent six-year averages during the time period January 1980 through December 1983 and January 1987 through December 1988 (with the exception that, because of low data recovery, April 1979 and May 1979 were substituted for April 1980 and May 1980). For the site boundary receptors, both six-year average growing season (April through September) and year-round (January through December) diffusion factors were generated, with the higher of the two chosen to represent the site boundary. The meteorological diffusion factor used in the development of the ODCM Method I dose models are summarized on Tables B.7-4 through B.7-6. B.7-32 ODCM Rev. 24

Table B.7-4 Seabrook Station Long-Term Average Dispersion Factors* Primary Vent Stack Dose Rate to Individual Dose to Air Dose to Critical Organ Total Skin Critical Gamma Beta Thyroid Body Organ XIQ depleted (s) 7.5E-07 7.5E-07 X/Q ndepleted ) 8.2-0 . .E0 (-1 (sec (8.2E-07 - . 8.2E-07 X/Q undepleted D/Q---- C. - 1.538)- l M3)E-08" 1.51-08 :S>1 West site boundary, 974 meters from Containment Building Northwest site boundary, 914 meters from Containment Building B.7-33 ODCM Rev. 24

1L Table B.7-5 Seabrook Station Long-Term Average Dispersion Factors for Special (On-Site) Receptors Primary Vent Stack Dose to Critical Dose Rate to Individual Dose to Air Organ Total Skin Critical Gamma Beta Thyroid Body Organ Education Center: (SW - 488 meters) ( X/Q depleted se) -- I. SE-06 -- 1.51E-06 sec-0 l.5E-066 15-0 X/Q udepleted,3 X/Q undepleted C(-3c) (in' _ 1.6E-06 -- 1 .6E D- - 2.7E-O8 - - XIQr(sec; l.lE-06 1.lE-06 1.lE-06 The "Rocks": (ENE - 244 meters) Qsec) _ 1.6E-05 1.6E-05 XIQ depleted m 3L-XsQ undepleted sec; 1.7E-05 1.7E-05

                                --                        1 I IE-07      -              -

X/(seQcs 5.OE-06 5.0E-06 5.0E-06 Qm 3 _ . _ .. O B.7-34 ODCM Rev. 24

Table B.7-6 Seabrook Station Long-Term Atmospheric Diffusion and Deposition Factors Ground-Level Release Pathway RECEPTOR(a) Diffusion Factor The Rocks Science & Nature. Off-Site Center Undepleted CHI/Q, sec/mr3 1.6 x 10' 2.3 x I05 1.0 x 10-5 (244m ENE) (406m SW) (823m W) Depleted CHI/Q, sec/M3 1.5 x 104 2.1 x 1 5 9.6 x 106 (244m ENE) (406m SW) (823m W) D/Q, m 2 5.1 x 10' 1.0 X 10-7 5.1 x lo-, (244m ENE) (406m SW) (823m W) Gamma CHI/Q, sec/M3 2.6 x 10-5 5.3 x 106 3.4 x 104 (244m ENE) (406m SW) (823m W) (a) The highest site boundary diffusion and deposition factors occurred during the April through September growing season. Note that for the primary vent stack release pathway, none of the off-site receptor diffusion and deposition factors (located at 0.25-mile increments beyond the site boundary) exceeded the site boundary diffusion and deposition factors. B.7-35 ODCM Rev. 24

t 8.0 BASES FOR LIQUID AND GASEOUS MONITOR SETPOINTS 8.1 Basis for the Liquid Waste Test Tank Monitor Setpoint The liquid waste test tank monitor setpoint must ensure that the limits of Part A Control C.5. 1 are not exceeded in combination with any other site discharge pathways. The liquid waste test tank monitor is placed upstream of the major source of dilution flow. The derivation of Equation 5-1 begins with the general equation for the response of a radiation monitor: R =E CT, SE (8-1) (cps) (imli)(cs.) where: R Response of the monitor to radioactivity (cps). Sl = Detector counting efficiency for radionuclide "i" (cps/(gCi/ml)). C7r = Activity concentration of each gamma emitting radionuclide "i" in the mixture that the monitor has a response efficiency sufficient to detect (pCi/ml). The detector calibration procedure for the liquid waste test tank monitor at Seabrook Station establishes counting efficiency by use of a known calibration source standard and a linearity response check. Therefore, in Equation 8-1 one may substitute SI for Sli, where SI is the detector counting efficiency determined from the calibration procedure. Therefore, Equation 8-1 becomes: R = Si F, Cr (8-2) (cps) = (cUC1z ) (ml ) B.8 ODCM Rev. 25

I I The ECL for a given radionuclide must not be exceeded at the point of discharge to the environment. When a mixture of radionuclides is present, 10 CFR 20 specifies that the concentration (excluding dissolved and entrained noble gases) at the point of discharge shall be limited as follows: ECL< (8-3) where: Cdi = Activity concentration of radionuclide "i" determined to be present in the mixture at the point of discharge to the environment (giCi/ml). ECL, = Effluent concentration limit (ECL) for radionuclide."i" (except for dissolved and entrained noble gas) in FiCi/ml as specified in 10 CFR 20, Appendix B, Table 2. The limit for the sum of all noble gases in the waste discharge is 2E-04 gCi/ml. (See ODCM Appendix B for listing.) The activity concentration of radionuclide "i" at the point of discharge is related to the activity concentration of each radionuclide at the monitor as follows: Cdi =F_ (Cyi + C/i1) Fd (,ui) gpm. (uCi)- t-Mml gpm) l and with equivalence of C; = (Cy3 + C13i), Equation 8-4 can be written as

                       = . Ci Fd where:

Fm = Flow rate past monitor (gpm) Fd = Flow rate out of discharge tunnel (gpm) Ci= Activity concentration of non gamma emitting radionuclide "i in the mixture at the monitor for which the monitor response is inefficient to detect (pCi/ml). C; = The activity concentration of each radionuclide "i" in the waste stream. This includes both gamma and non gamma emitters, such as tritium. B.8-2 ODCM Rev. 25

Substituting the right half of Equation 8-4 for Cd ini Equation 8-3, and solving for Fd/Fm yields the dilution factor needed to complete Equation 8-3: DF.j. < F. 2 C' (8-5)

                       -F.        ~10ECL1 CgpJ (gpM)         PCi -ml) i-m i JC; where:

ECLI = Effluent concentration limit (ECL) for radionuclide "i" (except for dissolved and entrained noble gas) in jICi/ml as specified in 10 CFR 20, Appendix B, Table 2. For noble gases, a value of 2E-04 pCi/ml is used for the limit of the sum of noble gases in the waste stream. If Fd/Fm is less than DFmin, then the tank may not be discharged until either Fd or Fm or both are adjusted such that: Fds DFi <Ed 7. (8-5) The maximum allowable discharge flow rate past the monitor can be found by setting Fm to Fnax and its equivalents, i.e:

                    =     Fd DF=i Usually Fd/Fm is greater than DFmjn (i.e., there is more dilution than necessary to comply with Equation 8-3), but must be satisfied since the monitor can only detect the gamma emitting portion of the waste stream. The response of the liquid waste test tank monitor at the setpoint is therefore:

Rs = fl x Fd x SIF Cy1 Fm x DFmi (cps) ()cps - ml(Ci (8-6) B.8-3. ODCM Rev. 25

I I or with Fnax substituted into Equation 8-6 for the maximum allowable discharge flow rate ( DFnin. ), the setpoint equation can be stated also as: DF~ Rsvow -`=fi x F- x Sj CY Fm where fl is equal to the fraction of the total concentration of ECL at the discharge point to the environment to be associated with the test tank effluent pathway, such that the sum of the fractions of the four liquid discharge pathways is equal to or less than one (f1 + f2 + f3 + f4 < 1). The monitoring system is designed to incorporate the detector efficiency, Si, into its software. This results in an automatic readout in )lCi/ml or pCi/cc for the monitor response. Since the conversion for changing cps to pCi/ml is inherently done by the system software, the monitor response setpoint can be calculated in terms of the total waste test tank activity concentration in pCi/ml determined by the laboratory analysis. Therefore, the setpoint calculation for the liquid waste test tank is: R , =fx Fd x Z Cyj (5-1) Fm x DF() m () () (ci B.8-4 ODCM Rev. 25

8.2 Basis for the Plant Vent Wide Range Gas Monitor Setpoints The setpoints of the plant vent wide range gas monitors must ensure that Part A Control C.7. 1. .a is not exceeded. Part B, Sections 3.4 and 3.5 show that Equations 3-3 and 3-4 are acceptable methods for determining compliance with that Part A Control. Which equation (i.e., dose to total body or skin) is more limiting depends on the noble gas mixture. For the limiting setpoint case, the gas mixture associated with the fuel gap activity at time of shutdown (UFSAR Table 15.7-20) indicates that the total body dose rate to the maximum offsite receptor is the limiting dose rate type. The derivations of Equations 5-5 and 5-6 begin with the general equation for the response R of a radiation monitor: R = ESg, Cn (8-7) (cpm) = (CPMCM) (C) where: R = Response of the instrument (cpm) Sgi = Detector counting efficiency for noble gas "i" (cpml(jiCi/cm 3 )) Cm;= Activity concentration of noble gas "i" in the mixture at the noble gas activity monitor (pCi/cm 3 ) Cm;, the activity concentration of noble gas "i" at the noble gas activity monitor, may be expressed in terms of Qj by dividing by F, the appropriate flow rate. In the case of the plant vent noble gas activity monitors the appropriate flow rate is the plant vent flow rate.

                            .1 C,,      Q     F                 .                                                      (8-8)

(,uCi) (PCi) (sec) tcm3 = tec) c3 where:

                      =      The relative release rate of noble gas "i" identified or postulated to be in the mixture.

F = Appropriate flow rate (cm 3 /sec) Substituting the right half of Equation 8-8 into Equation 8-7 for Cm; yields: R =ESg 1 1i (8-9) (cpm)= cPm Ccm) (Ui2) cms) B.8-5 ODCM Rev. 25

I I As in the case before, for the liquid waste test tank monitor, the plant vent wide range gas monitor establishes the detector counting efficiency by use of a calibration source. Therefore, S. can be substituted for Sgi in Equation 8-9, where Sg is the detector counting efficiency determined from the calibration procedure. Therefore, Equation 8-9 becomes: R = Ss (8-10) (p =Cip ) scm ) (seci The total body dose rate due to noble gases is determined with Equation'3-3a: Dbhw = 0.85 * *(Q1 )

                                           *DFB                                                     (3-3a)

C s) -pCi - secj (pCi)J mrem - m' J (Pt iM3) _ tsec) pCi -yr) where:

         )=               Total body dose rate (mrem/yr) 0.85        =       (1.OE+06) x (8.5E-07) (pCi-sec/pCi-m 3 )

IE+06 = Number of pCi per pCi (pCi/pCi) 8.5E-07 = [X/OJf, maximum off-site average gamma atmospheric dispersion factor (sec/m 3 ) for primary vent stack releases

                  ;=     The relative release rate of noble gas "i" identified or postulated to be in the gas mix (glCi/sec).

DFB; = Total body dose factor (see Table B.l-I0) (mrem-m 3 /pCi-yr) B.8-6 ODCM Rev. 25

A composite total body gamma dose factor, DFBc, may be defined such that: DFBC XQ; = E DFB, (8-11) i mrem - m3

                                    =     i'~     (recm pCi - yr      ksecJ Solving Equation 8-11 for DFBC yields:

ZQDFB3 DFB¢= i (5-7) XQi Part A Control C.7.1.1.a limits the dose rate tothe total body from noble gases at any location at or beyond the site boundary to 500 ffirem/yr. By setting fj, equal to 500 mrem/yr and . substituting DFBc for DFBj in Equation 3-3, one may solve forXQ 1 at the limiting whole body I . .I I i noble gas dose rate: 1 (Pi = 588 (8-12) DFBC (sc mrem - uCi - m) ( pCi -yr yr -pCi - sec ) nirem - m) Substituting this result for XQ 1 in Equation 8-10 yields Rtb, the response of the monitor at the limiting noble gas total body dose rate: 1 1 Rt = 588 SR (8-13) F DFBC (cpm)= mrem -puCi-3

            .yr -pCi -se (cPm _ cm3 Pi

( ( (cm3 sec pCi-yr ) mrem - m)3 B.8 ODCM Rev. 25

The skin dose rate due to noble gases is detennined with Equation 34a: Dskin) = I (Q

• DF')) (3-4a) nirem = (PCa _mrem-sec yr sec ,Ci - yr )

p where: D3s.ilce) - Skin dose rate (mrem/yr)

               ;=      As defined above.

DF 1 ()= Combined skin dose factor (see Table B. 1-1) (mrem-sec/pCi-yr) A composite combined skin dose factor, DFC, may be defined such that: DF4 C

• ZQ; = z (i
• DFi,) (8-14)

Cmrem-sec, C _-sec) pCi - yr (,uci) sec )

J Solving Equation 8-14 for DF'c yields:

0' ZQiDF, (e) DF, = i E (5-8) B.8-8 ODCM Rev. 25

Part A Control C.7.1. La limits the dose rate to the skin from noble gases at any location at or beyond the site boundary to 3,000 mrem/yr. By settingb5 u, equal to 3,000 mrem/yr and substituting DF'c for DF'i in Equation 3-4 one may solve forE Q, at the limiting skin noble gas dose rate:

                   =      3,000                                                                       (8-15)

DF' (lo~~~~ci mrm( uiy)- sec tyr )mrem - sec) Substituting this result forl:Q1 Oin Equation 8-10 yields Rskim, the response of the monitor at the limiting noble gas skin dose rate: Raki, = 3,000 Sz - F DF'C (me)(cpm-cm3 {see) ( ui-c yr (cpm) ((8-16)

                              <           yra      Ccm 3    mrem -sej As with the liquid monitoring system, the gaseous monitoring system is also designed to incorporate the detector efficiency, Sg, into its software. The monitor also converts the response output to a release rate (LCi/sec) by using a real time stack flow rate measurement input.

Therefore, multiplying by the main plant vent flow rate measurement (F), the Equations 8-13 and 8-16 become: Rt() 588 (5-5) DFBc (Ci) (mrem- LCiM3)( pCi-yr) sec yr - pCi - sec mrem-m R skin(e) 3,000 1 (5-6) DF'C pCi) nmrern pCi - yr sec yr mrem - sec These equations assume that the main plant vent is the only release point contributing to the determination of limiting offsite dose rate. The Control dose rate limits (500 mrem/yr and 3000 mrem/yr for total body and skin, respectively) apply to combination of all release points to the limiting offsite receptor. Administrative fractions (f) should be applied to main plant vent setpoint calculation as a multiplier, and any other release points, such that the summation of all fractions is less than or equal to 1. This provides for the combined impact of all release points to ensure that selected setpoints alarm at or before the site dose rate limits is exceeded. B.8-9 ODCM Rev. 25

II 8.3 Basis for PCCW Head Tank Rate-of-Change Alarm Setpoint The PCCW head tank rate-of-change alarm will work in conjunction with the PCCW radiation monitor to alert the operator in the Main Control Room of a leak to the Service Water System from the PCCW System. For the rate-of-change alarm, a setpoint based on detection of an activity level of 1O' iCi/cc in the discharge of the Service Water System has been selected. This activity level was chosen because it is the minimum detectable level of a service water monitor if such a monitor were installed. The use of rate-of-change alarm with information obtained from the liquid sampling and analysis commitments described in Table A.6.1-1 of Part A ensure that potential releases from the Service Water System are known. Sampling and analysis requirements for the Service Water System extend over various operating ranges with increased sampling and analysis at times when leakage from the PCCW to the service water is occurring and/or the activity level in the PCCW is high. B.8-10 ODCM Rev. 25

8.4 Basis for Waste Gas Processing System Monitors (RM-6504 and RM-6503) The maximum allowable setpoint for the waste gas system monitors (response in pCi/cm3 ) can be determined by equating the limiting off-site noble gas dose rate from the plant vent to the total body or skin dose rate limits of Part A Control C.7..1. La, assuming that all the activity detected by the vent wide-range gas monitors is due to waste gas system discharges. By evaluating the noble gas radionuclide with the most limiting dose factor as given on Table B.I-10, a conservative activity release rate from the plant vent for both whole body and skin dose rate conditions can be calculated. From Table B.1-10, Kr-89 is seen to be the most restrictive individual noble gas if it were present in the effluent discharge. Applying plant vent setpoint equation 5-5 for the whole-body, and equation 5-6 for the skin, the maximum allowable plant vent stack release rate can be calculated as follows: Rt = 588 l/DFBC (5-5) where: Rtb = plant vent maximum release rate (pCi/sec) based on'the whole body does rate limit of 500 mrem/yr DFBc = 1.66E-02 (mrem-m3 /pCi-yr), whole body dose factor for Kr-89 588 = conversion factor (mrem-pCi-m 3 /yr-pCi-sec) Therefore: Rh= 588 1/1.66E-02

              =      35,421 pCi/sec maximum release rate at plant vent Next, the skin dose rate limit is evaluated from equation 5-6 in a similar fashion as follows:

Rskin = 3000 I/DFc (5-6) where: Rskin = plant vent maximum release rate (pCi/sec) based on skin dose rate limit of 3000 mrem./yr. DF'c = 2.45E-02 mrem-sec/pCi-yr skin dose factor for Kr-89 3000 = Site boundary skin dose raie limit (mreiTn/yr) B.8-1 1 - ODCM Rev. 25

L1 therefore: Rski. = 3000 (mrem/yr) 1/2.45E-02(mrem-sec/dCi-yr)

                   =   122,449 piCisec from the plant vent Comparing the release rate limit for the whole body to that for the skin (i.e., 35,421 piCi/sec vs 122,449 gCi/sec, respectively) it is determined that the release rate for the whole body is limiting.

Next, to get the maximum plant vent release rate from the waste gas system discharge, equate the plant vent maximum release rate limit for the whole body equal to the waste gas system activity concentration times its flow rate to the plant vent, i.e.: Rtb = 35,421(tLCi/sec) = Rg(p Ci/cm3 ) F.(cm3/sec) or solving for Rwg: Rn(gUCi/cm 3 ) = 35,421(pCi/sec) / F n(cm 3/sec) where: Rwg = maximum concentration (setpoint limit) at the waste gas system monitors Fwg = waste gas design flow of 566.4 cm 3 /sec (1.2 cfmn) therefore: Rwg(gCi/cm 3 ) = 35,421(jxCi/sec) / 566.4(cm 3 lsec)

                           =      62.5 pLCi/cm 3 This represents the maximum waste gas discharge concentration which would equal the site boundary whole body dose rate limit for plant vent releases. Administrative controls may set alert alarm and high alarm (waste gas isolation) setpoints on the waste gas monitors as some multiple of expected activity concentration, such as 1.5 and 2 times, respectively, as long as the maximum setpoint does not exceed 62.5 pCi/cm 3 . This provides operational controls to be exercised before any waste gas discharges could equate to the Part A Control C.7.1.1.a.

The primary process monitor noted in Part A Control C.5.2 is RM-6504, which is downstream of the waste gas discharge compressor at the end of the process system. Monitor RM-6503 is on the inlet side of the compressor downstream of the charcoal delay beds, and is considered as an alternate monitor if RM-6504 is inoperable. For the purpose of setting the maximum discharge setpoint, RM-6503 is treated the same as RM-6504, which assumes no additional source reduction before discharge to the plant vent. B.8-12 ODCM Rev. 25

8.5 Basis for the Main Condenser Air Evacuation Monitor Setpoint (RM-6505) The maximum allowable setpoint for the main condenser air evacuation monitor must be evaluated for two modes of operation. For normal operations the monitor is responding to a low flow rate that is typically released through the plant vent stack. During start-up (hogging mode), the monitor response must be related to a high flow rate that is being released from the turbine building which is considered a ground level release. In both instances, the setpoint can be determined by equating the limiting off-site noble gas dose rate from the release point to the total body or skin dose rates of Part A Control C.7.1.1.a. The most restrictive noble gas mixture has been found to be represented by the noble gases associated with the fuel gap activity at the time of plant shutdown. This mixture is listed on UFSAR Table 15.7-20, and provides a limiting setpoint calculation that bounds other potential or observed offgas mix conditions. In addition to monitoring the main condenser air, the air evacuation monitor response is also used as an indicator for Turbine Gland Seal Condenser exhaust. Since this is a potential release pathway during both the normal and the hogging modes of operation; the impact is considered in the setpoint calculations. 8.5.1 Limiting Example for the Air Evacuation Monitor Setpoint During Normal Operations During normal power operation, the maximum allowable setpoint for the air evacuation monitor is determined by applying plant vent setpoint equation 8-13 for the total body, and equation 8-16 for the skin. Therefore, the maximum allowable stack release rate can be calculated as follows: Rtb(e) = (588) (Sg) (I/F) (l/DFBc) (8-13) (cpm): = (mrem-iiCi-m 3 /yr-pCi-sec) (cpm-cm3/4Ci) (sec/cm3 )(pCi-yr/nrem-m 3 ) where: Rtube) = count rate (cpm) for the plant vent maximum release rate based on the total body dose rate limit of 500-mrem/yr 588 = conversion factor (mrem-pCi-m 3 /yr-pCi-sec) S = the detector response efficiency (cpm-cm 3 /pCi) as determined from monitor calibration. For the air evacuation monitor, a typical value is

                    -1.87E+08 cpm-cm 3 /LCi. .

F = release flow rate. During normal operations, a typical flow value ranges from 10 to 50 cfm (2.36E+04 cc/sec maximum) for the air'evacuation pathway. DFBc = the composite total body dose factor, (mrem-mr3 /pCi-yr). For different gas mixes, the composite can be found from: DFBC = EQ DFBi0/ 7-Q (5-7) i i DFBc for the limiting gas mixture is 4.86E-03 mrem-m 3 /pCi-yr (See Section 5.2.1.2) B.8-13 ODCM Rev. 25

I I Therefore, Rtb(c) = 588 1.87E+08 (1/2.36E+04) (1/4.86E-03)

                =    9.59E+08 cpm detector count rate for a'maximum release rate at the plant vent based on the total body dose rate.

Next, the off-site skin dose rate limit is evaluated from equation 8-16 in a similar fashion as follows: Rsid,(,) = 3000 Sg (1/F) (I/DFc) (8-16) (cpm) = (mrem/yr) (cpm-cm 3 /Ci) (sec/cm 3 ) (pCi-yr/mrem-sec) where: Rsedn(e) = count rate (cpm) for a plant vent maximum release rate based on the skin dose rate limit of 3000 mrern/yr DF, = the elevated release skin dose factor for the limiting noble gas mix associated with fuel gap activity at shutdown is calculated in the example provided in Section 5.2.1.2, and is equal to 6.80E-03 (mrem-sec/lCi-yr). Therefore,' Rsldn(e) = 3000 1.87E+08 (1/2.36E+04) (1/6.80E-03)

                =   3.50E+09 cpm detector count rate for a maximum release rate at the plant vent based on the skin dose rate.

Comparing the release rate limit for the total body to that of the skin (i.e., 9.59E+08 cpm versus 3.50E+09 cpm, respectively) it is determined that the release rate for the total body is limiting in this case. Since during normal operations the Turbine Gland Seal Condenser exhaust has the potential to be a minor additional contribution to the total site release, the effective contribution from the main condenser exhaust must be' limited to some fraction of the calculated value. The contribution from the Turbine Gland Seal Condenser exhaust is expected to be minor because this system handles only 670 lbs/hour of steam which is a very small fraction of the 1.5E+07 lbs/hour of secondary side steam that the main condenser handles. Therefore, the maximum alarm is set at 6.71E+08 cpm, which is 70% of the calculated value, to ensure that the contribution of the two does not exceed the dose rate limit of Part A Control C.7.1.1.a. During normal operations, this would represent the maximum allowable count rate on the air evacuation monitor that would equate to the site boundary total body dose rate limit or less. B.8-14 ODCM Rev. 25

8.5.2 Example for the Air Evacuation Monitor Setpoint During Startup (Hogging Mode) During startup (hogging mode), the determination of the air evacuation setpoint must take into account a larger air flow rate that is also released as a ground level effluent. The flow rate must also include the contribution from the Turbine Glind Seal Condenser exhaust, which is a potential release pathway which the air evacuation monitor response must also take into account. For ground releases, the general equation 8-10 is used to represent the monitor count rate. R = (Sg) (1I/F) YQj (8-10) (cpm) = (cpm-cm 3 /pCi) (sec/cm 3 ) (piCi/sec) where: R = detector count rate (cpm) Sg = the detector efficiency (cpm-cm3I/Ci) F = release flow rate (cm3 /sec)

            =   the release rate of noble gas "i" in the mixture, for each noble gas listed in Table B.1-I0.

For a ground release, the off-site total body dose rate is based on: Dj).W=3.4Z(Q,DFB;) (3-3b) A composite total body dose factor, DFB, can be defined such that: DFBC,Q 3  ; (QDFB) (8-11) i -i B.8-15 ODCM Rev. 25

II By substituting 8-11 into 3-3b and rearranging to solve for yQ, the following equation is obtained: 2:Qi=(15(Ns,3.4) (O/DFBc) By inserting a limiting value of 500 mnrem/yr as f~bfg) this simplifies to: 147 (l/DFB,) Insertion of this equation into equation 8-10 yields: Rtwg) = 147 SWz (I/F)(I/DFBe) (cpm) = (mrem-ljCi-m3/yr-pCi-sec) (cpm-cm3 4LCi) (sec/cm i 3) Ci-yr/mrem-m 3 ) where: R, 8) = count rate (cpm) for the maximum ground release rate based on the total body dose rate limit of 500 mrem/yr. 147 = conversion factor (mrem-jiCi-m 3 /yr-pCi-sec) Sg = the detector response efficiency for the air evacuation monitor (a typical value of 1.87E+08 cpm-cm3/pCi is applied in this example). F = release flow rate. During the hogging mode of operation, a value of 4.72E+06 cm3/sec (10,000 cfm) is assumed. This represents the hogging flow that is discharged to the Turbine Building roof via the air evacuation monitor. An additional 1800 cfin is discharged from the Gland Seal Condenser exhaust directly to the Turbine Building roof without passing via the air evacuation monitor. To account for this unnionitored flow, an administrative fraction (fg1and) is applied to the setpoint calculation to ensure that the monitor would alarm before the dose rate limit for the combined release would be exceeded. One approach for determining a conservative fraction is to assume that the radioactivity concentration in the gland seal exhaust is equal to the main condenser offgas, even though the steam flow to the gland seal system is a very small fraction of the steam flow to the main condenser. Then the ratio of the Gland Seal Condenser exhaust flow to the total flow of hogging discharge and gland seal condenser provides for the relative flow of both sources. For the stated conditions, the unmonitored flow is about 15 % of the total (as additional conservatism, this could be doubled to 30% for the relative proportion assumed to be contributed by the unmonitored pathway). Therefore, fgld = 1-0.3, or 0.7 as the fraction applied to the air evacuation monitor setpoint. An additional fraction (fg) is also applied to account for the potential offsite dose rate contribution from this total ground source vs the plant main vent (fg < I - fQ). The split for this illustration is set at 0.3 for ground sources and 0.7 for the plant vent. B.8-16 ODCM Rev. 25

DFBc = Composite total body dose factor which weights the combination of total body dose factors (from ODCM Table B.1-10) of each radionuclide assumed to be in the gas mix in accordance with the fraction that it makes up of the total release. For the limiting noble gas mix associated with fuel gap activity at shutdown (see example calculation provided in Section 5.2.1.2), the value is equal to 4.86E-03 (mrem-m3 /pCi-yr). In addition, two administrative fractions are applied to the general calculation to account for other release contributions to the site dose that do not go by the air evacuation monitor. The first (fg ) is the fraction of the site boundary total body dose rate limit to be administratively assigned to monitored ground level releases (for this illustration 0.3) such that the combination of the plant vent fraction (f, ) and ground fraction (fg ) is less than or equal to I (fg < 1 - fQ). The second release reduction factor (fgld ) is administratively assigned to account for potential unmonitored contributions from the Turbine Gland Seal Condenser exhaust (for this illustration = 0.7) which discharges to the Turbine Building roof without going past the air evacuation monitor Therefore: Rt~bg) = (147) (1.87E+08) (1/4.72E+06) (1/4.86E-03) (0.3) (0.7)

               -   2.52E+05 cpm detector count rate for a maximum ground release rate based on the total body dose rate.

Next, the off-site skin dose rate limit for a ground release is evaluated from equation 3-4b in a similar fashion as follows: DSki = (3-4b) A composite skin dose factor, DF'c(g) can be defined such that: DFc(g) EQj = (QDFiZ)) (8-17) By substituting 8-17 into 3-4b and rearranging to solve for EQi the following equation is obtained: FM = Dfswz) (l/DF;) By inserting a limiting value of 3000 mrem/yr as Ds,.,) this simplifies to: XQj = 3000 (l/DF'c(g)) Insertion of this equation into equation 8-10 yields: Rskin(g) = 3000 Sg (1/F) (1/DFc(g)) (cpm) = (mrem/yr) (cpm-cm 3 /jiCi) (sec/cm 3 ) (tCi-yr/mrem-sec) B.8-1 7 ODCM Rev. 25

I I where: Rskin(g) = Count rate (cpm) for the maximum ground release rate based on the skin dose rate limit of 3000 mrem/yr. DFc(g) = The composite ground release skin dose factors which weights the combination of the combined skin dose factors (from ODCM Table B.1-10) of each radionuclide assumed to be in the gas mix in accordance with the fraction that it makes up of the total release. For the limiting noble gas mix associated with fuel gap activity at shutdown (see example calculation provided in Section 5.2.1.2), the value is equal to 6.80E-03 (mrem-sec/gCi-yr). As with the whole body dose rate above, the same two administrative fractions, fg and fglamd are also applied to the skin dose rate response. Therefore: Rskin(g) (3000) (1.87E+08) (1/4.72E+06) (1/6.80E-03) (0.3) (0.7)

              -      3.67E+06 cpm detector count rate for a maximum ground release rate based on the skin dose rate.

Comparing the release rate limit for the total body to that of the skin (i.e., 2.52E+05 cpm versus 3.67E+06 cpm, respectively) it is determined that the release rate for the total body is limiting in this case. B.8-18 ODCM Rev. 25

REFERENCES A. Regulatory Guide 1.109, "Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with I OCFR50, Appendix 1", U.S. Nuclear Regulatory Commission, Revision 1, October 1977. B. Hamawi, J. N., "AEOLUS A Computer Code for the Determination of Continuous and Intermittent-Release Atmospheric Dispersion and Deposition of Nuclear Power Plant Effluents in Open-Terrain Sites, Coastal Sites, and Deep-River Valleys for Assessment of Ensuing Doses and Finite-Cloud Gamma Radiation Exposures," Entech Engineering, Inc., March 1988. C. Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases From Light-Water Cooled Reactors", U.S. Nuclear Regulatory Commission, March 1976. D. National Bureau of Standards, "Maximum Permissible Body Burdens and Maximum Permissible Concentrations of Radionuclides in Air and in Water for Occupational Exposure", Handbook 69, June 5, 1959. E. Slade, D. H., "Meteorology and Atomic Energy - 1968", USAEC, July 1968. F. Seabrook Station Technical Specifications. R-I ODCM Rev. 21

APPENDIX A

                       . k r                            .

_ . j . DOSE CONVERSION FACTORS '; A-1 ODCM Rev. 21

I I APPENDIX A METHOD I DOSE CONVERSION FACTORS I. LIQUID PATHWAYS - SEABROOK SITE SPECIFIC DCF'S The models used to assess doses resulting from effluents into liquids is derived from Appendix A of Reg. Guide 1.109. Since Seabrook is a salt water site, the assumed pathways of exposure taken from Reg Guide 1.109 are Aquatic foods - fish; Aquatic foods -invertebrates; and dose from shoreline deposits (direct dose). No drinking water or irrigation pathways exist because of the salt water environment. In addition, exposures resulting from boating and swimming activities have been included for key radionuclides even though Reg. Guide 1.109 identifies these pathways as not contributing any significant contribution to the total dose, and therefore does not provide dose equations for them. For completeness, the swimming and boating pathways have been included using the dose models from the HERMES code (HEDL-TME-71-168, Dec. 1971) section G, Water Immersion. The Method I dose conversion factors are derived by calculating the dose impact to individuals via the site specific pathways for a unit activity release (I curie per nuclide). For each pathway, doses by radionuclide are calculated for each of the 7 organs (including whole body) for each of the four age groups (adult, teen, child, and infant). The Method I dose factor for each nuclide is then selected by taking the highest factor for any organ in any of the age groups for all the exposure pathways combined. The list of dose factors in the ODCM then represents a combination of different limiting organs and age groups which, when used to calculate a dose impact from a mix of radionuclides released in liquid effluents, gives a conservative dose since it combines the exposure to different organs and age groups as if there was a single critical organ-age group. As an example of how the liquid dose conversion factors are developed, the following calculation for Co-60 is shown. The critical organ/age group is selected based on the full assessment of all organs and age groups. Factor for fish Ingestion: The general equation for ingestion doses in RG 1.109 is eq. A-3. 1119.7* Up*MP *Qj *B*Dipj*e* Xp F The full assessment for the ODCM dose factors indicated that for i = Co-60, the maximum dose (mremn/yr) is to the GI-LLI of an adult as the target organ and age group, therefore: Uap  : 21 kg/yr adult usage factor for fish A-2 ODCM Rev. 21

MP 0.1 mixing ratio for near field dilution provided by submerged multiport diffuser. F 918 cu. ftlsec effluent flow rate for circulating water system 1.0 curies/year released of Co-60 assumed Bp  : 100 equilibrium bioaccumulation factor for Co-60 in salt water fish, in liters/kg Dmipj 4.02

• 10-5 mrem/pCi. adult GI-LLI ingestion dose factor from RG-1.109, table E-11.

1.501

• 10' -decay constant for Co-60 in 1/hrs.

tp 24 time between release and ingestion, in hrs. 1119.7 is the factor to convert from Ci/yr per ft 3 /sec to pCifliter. Note that RG 1 109 uses 1100 as a rounded approximation. Therefore the dose from fish to adult GI-LLI is (mrem/yr): 1119.7* U*MP

• Qj*B*D j*e 'P=0.0103 F

Factor for invertebrate ingestion: Next, the dose from invertebrates to the adult GI-LLI is given by the same general equation but with the following variables changed: UP 5 kg/yr usage factor Bp  : 1000 l/kg bioaccumulation factor all other variables the same as above therefore the dose from invertebrates is (mrein/yr): 1119.7

• Uw *Mp
• Q * *pDDa*e p= 0.0245 F

Factor for shoreline direct dose: The general equation for direct dose from shoreline deposits is taken from equation A-7 in RG-1.109 as (mrem/yr): A-3 ODCM Rev. 21

I I 111970

• PMPW* Z Q *T *Dp F
• e-l"' * [I eA b It is assumed that all internal organ doses also receive exposure from direct external sources, therefore each organ dose due to ingestion must have an external component added. For the above equation, the site specific variables for an adult exposure to a I curie per year release of Co-60 are:

UV  : 334 hrs/year usage factor used for assumed shoreline activities at Seabrook. MP 0.1 mixing ratio for near field dilution provided by the submerged multiport diffuser and assume to be extended to the beach continuously. W  : 0.5 shorewidth factor for ocean sites, dimensionless T 1.923*103 radioactive half life in days for Co-60 Daipj  := 1.70*104' dose factor for Co-60 due to deposits in sediments, units of (mrem/hr)/(pCi/m 2 ) tp 0.0 transit time to point of exposure, hirs tb 131400 period that sediment is assumed to be exposed to water contamination for long term buildup, set at 15 years for Method I DCFs Qi 1.0 curies per year, Co-60 assumed 111970 conversion factor to convert (Ci/yr)/(ft3 /sec) to pCi/liter and account for the proportionality constant used in sediment model Therefore the dose to the whole body and each organ due to direct exposure to the shoreline (mrem/yr) is: 111970* UaP*MP W

• Q. *T*Dip,*eA* [I -e-lt]= 00573 F

A-4 ODCM Rev. 21

Direct dose due to Swimming: The dose due to immersion in water (swimming) is taken from the HERMES computer code. The original ODCM calculation was based on some preliminary dilution assumptions which gave a near field prompt dilution factor for the multipart diffuser of 8. For single unit operation with both service water and circulating water flow (412,000 gpm), a value of 10 is more realistic. This surface area of the plume is restricted to a small area over the diffuser and does not touch the shoreline approx. 1 mile away. Since the over all impact from swimming is small when compared to the other exposure pathways, the original conservatism on dilution are kept here. The dose from swimming is given by the following equation: 1.0*1012 $

• E
• DFR. (mrem/yr)

F. Where: Up 45 hrs/yr, usage factor for swimming for maximum age group (teen) from HERMES. Fa  : 6.56*1011 liters/yr, estimated annual dilution effluent flow in multiport diffuser Qi 1.0 Curies/yr, assumed release rate of nuclide i. DFim  : 4.6*1 0 mrem-liters per hrs-pCi, dose factor for Co-60 for water immersion taken from HERMES. 1.0*101~2 constant for pCi/Ci Therefore the swimming dose for a 1 curie release of Co-60 is (mrem/yr): 1.0*112*U *MP*Qj*DFim=3.155 *IOs F. As can be seen, the contribution of the swimming dose is only about one 30000ths of the total of the RG 1.109 pathways, and can be ignored in the case of Co-60. Similarly, the boating dose as given in HERMES is taken as half of the swimming dose, (and corrected for change in usage assumptions). The resulting dose is found to be less than the swimming dose and can also therefore be discounted in this case. A-5 ODCM Rev. 21

II Total liquid Pathway dose: The sum of the above liquid pathway doses can now be added to give the total maximum individual dose to the critical organ (adult-GI-LLI) for Co-60. This gives: S 0.0 103 + 0.0245 + 0.0573 = 0.0921 mrem/yr Since the internal doses given by the RG-1.109 methods actually are 50 yr dose commitments resulting from one year exposure to the quantity of activity assumed to be released into the water, and the direct dose represents the dose received for the period assumed to be exposed to the pathway, and the activity release was taken as a unit quantity (i.e. Q = 1 Ci), the above total liquid pathway dose can be stated as site specific committed dose factor in mrem/Ci released. For Method I in the ODCM, the critical organ dose factor is seen to be 0.0921 mrem/Ci, as shown above. The value reported on Table B.1-1 1 (9.22 E-08 mremipCi) was generated by a computational routine which gives rise to the round-off difference between it and the above example. The whole body site specific dose factor for the ODCM was calculated in the same way treating the whole body as a separate organ. A-6 ODCM Rev. 21

I. GASEOUS PATHWAYS - SEABROOK SITE SPECIFIC DCF'S The models used to assess doses resulting from gaseous effluents in the form of iodines, tritium, and particulates are derived from Appendix C of Reg. Guide 1.109. For Seabrook, it is assumed that at the off site location which exhibits minimum atmospheric dilution for plant releases the following exposure pathways exist: inhalation, ground plane, ingestion of goats milk, meat, stored vegetables, and leafy vegetables. The Method I dose and dose rate factors are derived by calculating the dose impact to all age group individuals via the site specific pathways for a unit activity release (1 curie per nuclide). For each pathway, doses by nuclide are calculated for each of 7 organs (including the whole body) for each of the 4 age groups. The Method I dose factor for each nuclide is then selected by taking the highest factor for any organ in any of the age groups for all exposure pathways combined. The list of dose factors in the ODCM then represents a combination of different limiting organs and age groups which, when used to calculate the dose impact from a mix of radionuclides released into the atmosphere, gives a conservative dose since it combines the exposure to different organs and age groups as if they were for all the same critical organ-age group. As an example of how the gaseous particulate dose factors are developed, the following calculation for Mn-54 is shown. The critical organ/age group for Mn-54 was selected based on a full assessment ofall organ and age group combinations. For elevated releases from the plant vent stack to the maximum site boundary (max. dose point due to meteorology), the critical organ and age group for Mn-54 was determined to be the GI-LLI for the adult. PART A: INHALATION DOSE CONTRIBUTION The general equations for inhalation doses in RG 1.109 are eq. C-3, and C-4 which together give: 3.17*104*Ra*[-]* Q*DFAij Dj, Where for the case of Mn-54 releases, the variables above are defined as: 3.17* 104 is the number of pCi/Ci divided by the number of second per year R.  : 8000 the breathing rate for age group a (adults) in m3 /yr. X -7 7.5* 10- the long term average depleted atmospheric dispersion factor, in sec/m 3 , at the maximum exposure point off site (S.B.) Q the release rate of nuclide i to the atmosphere in Ci/yr A-7 ODCM Rev. 21

I I DFAija := 9.67*106 the inhalation dose factor for nuclide i (Mn-54), organ j (GI-LLI), and age group a (adult) taken from RG 1.109, table E-7, in mnrem/pCi inhaled.- Therefore, the inhalation dose to the maximum potential off site individual is given as: 3.17*10*R.4* *[ ] *Qj*DFAU 3 = 0.00184mrem/yrperCi PART B: GROUND PLANE DIRECT DOSE CONTRIBUTION The general equations for ground plane external direct dose in RG 1.109 are equations C-1 and C-2 which together give the dose DG as: 8760 *1.0*10o t*SF*[esQI vi' -hi DFGi Where for the case of Mnb-54 releases, the variables in the above equation are defined as: 1.0* 1012 is the number of pCi per Ci SF  : 0.7 the shielding factor provided by residential structures (dimensionless) for use in calculation accumulated doses over time. Note that for determination of dose rate factors (i.e. instantaneous dose rates) the shielding factor is set equal to 1.0, or in effect no credit for dose reduction is taken for determination of dose rates at points in time. D 1.5*10-" the long term average relative deposition factor at the maximum Q site boundary location, in 1/m2 A;  : 0.8105 is the radiological decay constant for Mn-54 (nuclide i in this case) in 1/yr. tb  : 15 is the time in years over which accumulation is evaluated (approx. midpoint of plant operating life) DFGj := 5.80*10-9 external dose factor to the whole body, or any internal organ j, for standing on contaminated ground from Mn-54 (RG 1.109 Table E-6) in mrem/hr per pCi/m2 Q  := 1.0 is the unit release quantity assumed for each nuclide i, in Ci/yr. A-8 ODCM Rev. 21

8760 is the number of hours in a year Therefore, the contribution to the total dose made by exposure to the ground plane at the maximum off site exposure location for Mn-54 is given as:, 8760*1-0*i 1° 2-*sF* I) *Q $I A $DFGi, = 0.658 - mrem per yr per Ci 11 A-9 ODCM Rev. 21

II PART C: INGESTION DOSE CONTRIBUTION: As an initial step to determining the dose contribution from ingestion of milk, 'Meat, stored vegetables, and leafy vegetables, we must first calculate the radionuclide concentration in forage, produce, and leafy vegetables resulting from atmospheric tranfers of the activity to the surface of the vegetation and onto the soil for root uptake. For all radioiodines and particulate nuclides (except tritium and C-14), the concentration of nuclide i in and on the vegetation at a point of interest can be calculated using R.G. 1.109 equations C-5 and C-6, which combined gives: 1.14*0 ]*Qj*[r* I-e +Bi,,*I-eft *e-A*It, Q] J Q.Y.,1O*[ A PART C.1: Concentration in Produce (stored vegetables) For the case of Mn-54 released in air emissions to the maximum site boundary, the concentration of Mn in produce grown in the hypothetical garden at that location can be calculated from the above equation where the variables are defined as:

1. 14* 1 0 is the number of pCi per Ci divided by the number of hours in a year (8760).

D =1.5* 1s is the relative deposition factor, in I/ff2 , at the maximum exposure point off site (S. B.) Q;  :=1 the release rate of nuclide i to the atmosphere in Ci/yr r  : 0.2 fraction of deposited activity retained on crops, leafy vegetables, or pasture grass (1.0 for iodines) XEi  : 0.00219 effective removal rate constant for Mn-54 from crops due to decay and weathering, in hr-I tb  : 131400 soil exposure time to deposition, in (equal to 15 yrs, or mid plant life) Y,  : 2.0 agricultural productivity (yield) for produce, in kg/m-2 Bi = 2.9*102 concentration factor for uptake of Mn-54 from soil by edible parts of crops in pCi/kg (wet weight) per pCi/kg dry soil A-10 ODCM Rev. 21

9.252*1 0 5 radioactive decay constant for Mn-54, in hrs-I P 240 effective surface density of soil, in kg/M2 th 1440 crop holdup time after harvest'and before ingestion, in hrs t, 1440 crop exposure time to plume, in hrs Therefore, the concentration of Mn-54 in stored vegetables produced at the location of maximum deposition for a unit activity release is given as: D 1-e&A.EIt. I- 'ejtb . 1 14*io*E.Q-1*Qi*r* +B1v** e^ =67.379 pCi/kg PART C.2: Leafy Vegetable Concentration For leafy vegetables, the above equation is repeated with the value for th, crop holdup time after harvest is changed from 1440 hrs to 24 hrs, i.e.: - th  : 24 crop holdup time after harvest, in hrs. Therefore the concentration of Mn-54 in leafyyvegetables at the maximum deposition point due to a unit activity release is given as: 1.14*10*[ ] *Qi* [rI I , =76.811 VueA pCi/kg PART C.3.a: Animal Feed concentration (pasture): Cp Next, we can repeat the above calculation to determine the concentration of Mn-54 in pasture grass used as animal feed. This will allow for the determination of dose contribution from milk and meat. For pasture grass, all the above variables remain the same except for: Y, 0.70 for agricultural productivity of pasture grasses, kg/M2 720 for grass exposure time to plume, hrs th 0.0 for holdup time after harvest A-11 ODCM Rev. 21

II Using these variables in the above equation gives the concentration in pasture grass as: 1.14*i0s*[ .]*Qi [r M PA +Bijv l' *e'*z=179.227 pCi/kg PART C.3.b: Animal Feed Concentration (stored feed): C, For stored feed that would be given to goats, or meat animals, the average concentration would be calculated by changing the following variables in the above calculation to: Y,  := 2.0 agricultural productivity for stored feed t  := 1440 feed crop exposure time to plume in hrs th  : 2160 feed crop holdup time after harvest, hrs Putting these values back into the above equation gives the concentration in stored animal feed (goat and meat animal) of Mn-54 for a unit activity release to the maximum exposure point. 1.14*10'*ID* *Q.* [ r*e +]R,* I; '=63.037 pCi/kg L QJI A Y" *,s PART C.3.c.: Concentration in Goat's Milk: Cm The Mn-54 concentration in milk is dependent on the amount and contamination level of the feed consumed by the animal. The radionuclide concentration in milk is estimated from RG 1.109 general equation C-10 as: F

• Cv
• QF
• e-A, = conc. in milk, pCi/liter where the variables are defined as:

Fm  : 2.5*104 average fraction of animal's daily intake of Mn-54 which appears in each liter of rmilk, in days/liter QF 6.0 amount of feed consumed by a goat per day, in kglday (50 kg/d for meat) tr  : 2.0 average transport time of activity from feed into milk and to receptor, in days. A-12 ODCM Rev. 21

i 2.22* 103 decay constant of Mn-54, in days-I In addition, the C, term for the concenitration of a nuclide in the animal's feed is given from RG 1.1 09 general equation C-Il as: ' C, = f

• f
• CP +1 - f]* C + fp* [-fJ]*C, where the following equals:

fp 0.5 fraction of the year that animals graze on pasture f 1.0 fraction of daily feed that is pasture grass when the animal grazes on pasture Cp 179.227 concenfration' of Mn-54 in pasture grass as calculated from above, pCi/kg Cs  : 63.037 concentration of Mn-54 in stored feed as calculated from above, in pCi/kg Therefore, the concentration in the total animal's feed is estimated to be: fp*f *CP+[l..-fJ* C.+f* [l-fj* C, 121.132pCiIkg When this value of 121.132 is put back into the above general equation for nuclide concentration in milk, we get: [C,  : 121.132 pCilkg I and Fm

• Cv
• QF* ei = 0.181 pCi/liter of Mn-54 in goats milk PART C.3.d.: Concentration in Meat: Cf Similar to milk, the concentration of the nuclide in animal meat is calculated. RG 1.109 general equation C-12 is given as: '-

Cf = Fr *Cv*QF*e "-'" Here the variables are set as:

wT Ff8.0* 10- fraction of animals daily intake of Mn-54 which appears in each of flesh, in days/kg ckg QF  := 50.0 animal's daily feed intake, in kg/day A-13 ODCM Rev.: 21

I I t, 20.0 average time from slaughter to consumption, in days C, 121.132 concentration on Mn-54 in animal's feed, same as calculated above for goat, in pCi/kg Therefore, the concentration of Mn-54 in animal meat is calculated to be: Fr

• Cv *QF*$e1 4 =4.635 pCi/kg in meat for Mn-54 PART D: DOSE FROM INGESTION OF FOODS PRODUCED AT MAXIMUM LOCATION Now that we have calculated the concentration of Mn-54 in milk, meat, leafy vegetables, and stored vegetables produced at a location of maximum air deposition, the resulting dose to any organ j and age group a can be calculated from the following general equation C-13 taken from RG 1.109:

ZDFIi.*[Uva *f, *Cv +Ur *C.+UF*Cf +UL.a*f *CL] For Mn-54 set equal to i, we find that from the evaluation of all organs for all age groups for combination of all exposure pathways, the adults GI-LLI is the critical age group/organ. Therefore, the variables in the above dose equation can be defined as: DFIija  : 1.40*10-5 ingestion dose factor for adults/GI-LLI for Mn-54, in mrem/pCi ingested (RG 1.109, Table E-11) Uva1  := 520.0 vegetable ingestion rates for adults, kg/yr fg  := 0.76 fraction of stored vegetables grown in the garden ft  := 1.0 of leafy vegetables grown in the garden

= 310.0 milk ingestion rate for adults, liter/yr uFa  := 110.0 meat ingestion rate for adults, kg/yr ULa  := 64.0 leafy vegetable ingestion rate for adults, kg/yr

= 67.379 concentration of Mn-54 in stored vegetables, in pCi/kg (from above)

Cm  := 0.181 concentration of Mn-54 in milk, in pCi/liter (from above) Cf  := 4.635 concentration of Mn-54 in meat, in pCi/kg (from above) A-14 ODCM Rev. 21

CL 76.811 concentration of Mn-54 in leafy vegetables, in pCi/kg (from above) The dose from the combination of ingestion pathways for this example is calculated by substituting the above listed variables back into the ingestion dose equation: DFIija * [1Ua

• f,
• Cv + U.a
• Cm + UF.
• Cf + UL.
• fl
• CL] = 0.4495 mrem-/yr per Ci By breaking the above dose equation down into the different pathways which combine to give the total ingestion dose, we can see the individual dose contribution made by each exposure pathway.

Therefore, we have: Dose for ingestion DFl4a *U,, *fg *C, = 0.373 of stored vegetables Dose for ingestion DFIjja *Uma *Cm = 7.855*104 of goat's milk Dose for ingestion DFlIja *UF, *Cf = 0.00714 of meat Dose for ingestion DFIjja *ULa *f1 *CL = 0.0688 of leafy vegetables PART E: TOTAL DOSE FROM ALL EXPOSURE PATHWAYS The total dose from all exposure pathways assumed to be present at the maximum receptor location can be found by simply adding the individual pathway doses calculated above. Since all the calculations above assumed a unit activity release from the plant vent stack, the combined dose can be stated as dose factor per unit activity released. This then demonstrates the development of the Seabrook ODCM Method I dose factors for gaseous release of particulates from the vent stack. Inhalation dose (Part A) 0.00184 mnrem/yr per Ci Ground plane dose (Part B) 0.658 mrem/yr per Ci Ingestion dose total (Part D) 0.449 mrem/yr per Ci Total dose all pathways 1.11 mrem/yrperCi (critical organ is GI-LLI of an adult for Mn-54) A-15 ODCM Rev. 21

APPENDIX B

                            . I.

ANNUAL AVERAGE EFFLUENT CONCENTRATION LIMITS TAKEN FROM 10 CFR 20, APPENDIX B B-1 ODCM Rev. 25

II App. B APO.B : PART 20 . STANDARDS FOR PROTECTION AGAINST RADIATION Table 1 table 2, Table 3 Occuationul Values Effluent Releasel to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. 1 Col. 2 Oral othly Ingestion Inhalatin . Average Atomic Radlrouclide Class ALt ALl . AC Ar 'dtr Concentration (MC) WO (pCI) (pCI/1l) (pC1/el) (pVI/1l) (pCI/al) I Hydrogen-3 Vter. DACincludes skin absorption 81E4 81E4 2E-S 11-7 1E-3 IE-2 1 Gas(NT or TI)Submersion : Ust abbowe values as MT and TI oxidfzo fn air nd In the body to KID. 4 Beryllfur-7 W. all cpounds except 4E14 2E-4 9E-6 3E-8 6E-4 6E-3

                                   . those given for Y Y oxfoea       halIdes, and nitrates                                         2E14             8E-6        3E-8            -

4 Beryllium-la V. sea' 1')3 2E+2 6E-8 2E-10 - LLI w all 7 (11.3) _ ._ 2E-S 2E-4

                            -         Y. see      B                      -             lE+1             6E-9'       2E-l            -

2 6 Carbon-ln Monoxide IE6 5E-4 2E-6 - Dioxide SE-5 3-4. 9E-7 - Cenpou-c 4E-S 4E15 2E-4 6E-7 6E-3 6E-2 6 Careon-14 monoxide 2E+1 7E-4 2E-6 - Dioxide 2E+5 9E-5 3E-7 - Comounds 2E*3 2E*3 1E-6 3E-9 3E-5 3E-4 9 Fluor1ne-u2 D n f iordes of HMLi. Pa. K. b. Cs. nd Fr SE-4 71E4 3E-S 1E-7 - St wal II (SE4) _ - - . 7E-4 71-3 V, fluorides of M. Mg. Cs, Sr. 158,-.AAl Co. In T1 AS Sb. II: Fe. Nu: os: Co, Fi, Fd, Pt u.0 Ag. h Zn. Cd. Mg. It Sc: V TI. Zr. V. lb. I.. Tx.I. ITc, andI - 9E14 4E-5 IE-7 . IL Y,.lanthanun f!uoride  ; 81E4 3E-5 11-7 - 11 Sodfhv-22 D. &al c Wounds 4E-2 61.2 3E-7 9E-10 6f-6 6E-S 11 Sodhow-24 D. all cpoumds 4E*3 5E-3 2E-6 7E-9 SE-S 5E-4 12 "Sim-28 0. all coounds except those givf for V 7E12 2E-3 7E-7 2E-9 9E-6 9f-S W. oxides. hydroxides. carbides, halides, and nitrates tE-3 SE-7 2E-9 - . 13 Aluinuiv-26 0. all cD ounds except thoae given for V 4E.2 6E-1 3E-8 9E-21 61-6 *E-5 V. oxides;.hydroxides. carbides halides, and nitrates 9E1 41-8i 1E-10 14 S111icon-31 . all* counds except those given for Vn Y 9E13 3E14 lE-5 4E-8 11-4 1E-3 V. eoxdes. hydroxides. carbides. and nitrates 3E14 lE-S SE-8 - Y. aluxroril(cate glass - 3E14 1E-S 41-U 14 SIlico)?Z D. sco n.S 2E13 21E2 1E-7 3E-10 - LLI wal 11 3 _ - - 4E-5 4E-4

                                      ,                             (313)             1E2             5E-8        2E-10          -

T: See 3Si SE+O 2E-9 7E-12 . - 15 Phosphorus-32 0, all comp ds' excext phosphatas given for V 61.2 9E-2 4E-7 11-9 9E-6 9E-S V, phosphates of Zn2, I 15 Phosphorus-33 S3+ D to 32 2+. and lanthniades V,se. 3 p1.3 p:: F34 3 6E?3 4E-2 8E'3 3E13 2E-7

                                                                                                    .41-6 1E-6 SE-10 1E-8 4E-9 SE-5       8E-4 B-2                                                                 ODCM Rev. 25

App. B App B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION. I Table I Table 2 Table 3 Occupational Values Effluent Releases t Cancetntrations Swe n Col. I Col.;Z Col. 3 Col. I Cal. 2

                                                                                 -Oral                                                                   Monthly.
                                                                              -            *S Cltien          Inhalatln                                      e ragte Atomi        ladlonuclioe      Clas                           -       AsLI -             LI          DAC       Air          Water   tencentration 0                                                (PM1           ON)         ("C11/6l)  (PCi/m1)-;(CSl            v/'

26 Sul fur-35 Vapor 1.E4. 6E-S 2E-8 -

0. sulfide and sulfates except those given for V 1E+4 2E14 71-6 2E-6 LU wallI (8E13) - - 1E-4 1E-3 VWelmtaUl sulfur. 6E13 sulfides of Sr. as. CG.

Sn. ft. AS. Sb. of, Cu. AuAu. Zn. Cd. No. W. and Mo. Sulfates of Ca. Sr.

a. R -As. Mb arid t1 2E+3 9E-7 31-9 17 C01arfto-36 0. cilerides et N. Lt1

s. K. fbCs, ad Fr 2E.3 2E13 1E-E 31-9 2E-5 2E-4 V chlorides of lafitha nWdes, q Ca. Sr.

Ba.Re Ai",.a In 711 Ga. Sn. ft. As. Sbe 11. F.. Ru. Os. CeaIth Ir. Ni. Pd. Pt. Cu: Ag, Au. Zn Cd. Mg: Sc. Y.T. Zr. Pi. V, b. . Cr. 1710 jMno W. i .:.4 n;Re - 2E12 YE-7 3E-10 2 17 Chiorfne-38 0.see 4 2*44 4E-4 2E-S 6E-8

                                                                                    -     St. wall
                                                                                   *(31.4)                                                      31E-4    3E-3 W.-ao36I                                                SE14           2E-S 17     -chlorine-ll9            0 B.. 3ee.                             2E-4            SE14           2E-S      7E-O. 6E-4 St. wall (4E14)                                     6E-S          SE-4     5E-3 6E44            2E-5 Argon-P    W.deS" tooI                                                         . 2-C.-                                                                                                 -

6E-3 1 is Argoa -39 *Su*,.orsfos 21-4 E8E-7 in is 'Argoe- I S.masiool, _3E_6 lE-8 1-S. 19 Potassium-40 D. all coqposands ,1,3E#2 4E12 2E-7 6E-10 4E-S JE-S

                                                                                                                                                       . 4.'

19' lotassism-42 U.'all compounds SE+3 SE+3 UE16 7E-9 SE-S 19 Potasasfmo-43 8, all - g wud 6E43 9ff3 4E-6 6E-4 19 Potiasia Iw-,4 P. .1I caoeonds 2E*4 7E14 3E-S SE-t 9E-S St. well 9-B

                                                                                      .(41.4)                                                  #E-4      51-3 19           Potassiaw-0S       D. all compounmds                                       1E-5            SE-5     2E-7 (SE-4)                                                            7E-3 20          .Calcifir-42        W. all c          emouas                31#;            4E13'           2E-6 sIo" surf       Sae surf.-

(41E3) (4E13) SE-i SE-S 6E-4 51-9 20 Calcliamr45 W.all le s 21.3 8E12 4E-7 2E-5 2£-4 20 - Calciam-47 W. all c houmds

• 342 91.2 4E-7 1E-9 IE-S 1E-4 21 - Scandlium43 V. all cFqInds 7E'3 2E-4 #-C 3E-t1 1E-4 1E-3 21 Scandtem-44w T. al tcoeposends 4143 51t2 7E12 3E-7 l E-9 7E-6 7E-S 21 .scandiwr-44 T. all coows 1E-4 SE 6 2E-t SE-S 5E-4 21 Scandfm-41 Y. all ctomnds 2E42 1E-7 3E-10 11-S 3E-4

               -21          Scandisar-47.       Y. all te                               21t3            3E13            1E-6     4E-9
                                                                                                                                            . JI W1 wal                                                          44-4 21           Scandisas-48        Y. utI compounds                                     1E.3           6E-7      21-9         3E-S      *E-4 2                                                                                     11E-9        3E-4       3E-3 21          Scandiwm-49         Y, all compomnds                          E40            SE-4           2E-5 22          Titanium-U          0. all comoumds as COPE those given for W a Rd Y                                lE-I            5E-9      2E-1.1        4E-6      4E-S WV.exides. kydroxidiit.

carbids, halides. KMd nitrates -3E*I1 lE-S 4E-ll - Y. SrTmOs - 61.0 2E-9 SE-12 - .I IB-3 ODCM Rev. 25

App.B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Table I Table 2 Ttb 1le 3 Or Occutational Values Effluent Ret.lates' to Concentrations Si wers Col. 1 Col. 2. Col. 3 Col. l Col. 2 Oral cIthly. Ingestion Inhala ion Ave rage Atosic Radionuclde Class AL ALI CAC Air Vater Conc entratfon to --- (yCI) 'OUC) (PCf/ml) (PCI/mI /31) N"1R)(CI 22 Titaniu-45 a' sea T1 9EU3 31E4 lE-S 3EU- 1E-4 IE-3

                                         -W.V     see 44T                     -              4E4              c lE-S             sE-r,     4 tt-                  -               1E4             if-S             4E-I 2

23 Vanadiiu-47 0. all comounds except thosa given for V 31E4 .E+4 3E-5 1£-7

St. wall t3E.4) - - - 4E-4 4E-3 V. "Ides, hydroxides.

carbides; and halides 4E-S 2E-7 23 Vnadiu48 see47V 6E*2 1EU3 SE-' 9E-6 91-S v to V 6E12 3E-7 91E-10 23 Vanr ol 49 D. Ste 4'V 7E.4 1.64 LUl wall Boe surf 47 1-~ (3E14) ..SE-8 7-35

• W. se ' V V, see  %::-j 2E^4 lE-6 2..4 2E-B 24 Chroefu-48 v. all

                                                 *1      pounds except those given for V and r         6E-i            11E4            SE-I                         84-S..         EE-4 i

i V. halides and nitrVteS 71E+3 1E-J lE-1,

                                           *r . ox des and ydroxides                                        3E-S 1£-5~

II 24 ChrO w4492 O se 4Cr 31E4 71.3 4E-S~ 1E-7 4E-4 4E-3 V see Cr ZE14 4E-S 1£-J 81.4 4E-5 48mcr 41.4 2E-S. . 6E-4 SE-4. SE-3 2E14 l1-S 3E-4. Y S Cr OE-6 3E-I 25 Meaase-52 0. all co*p1undsexcept these viven tfr V 2.44 SE-4 21.4 2E-S. * . .71-a 3E-4 3E-3 W. osides. y ee1 es V JE-4 halides, and nitrates 6EE4 3E-S 3..E IE-t 2

25. msre 5Se-52M D S" SI - 31E^4 4E-S av. _st4 (4E14) s£-4 SE-3

v. se IM 1E+S 4E-S9 1E-7. . . .

71 91.4

           ?S          Mangamese-52        0    S"    I                                     1E*4           SE-7              2E-9       lE-S

• 1E-4 4E-7 25 1engae-S3 0. Ste SIM SE-6 7E-4 7E-3 2E+3 Sone surf 51.4 (2E14) 31-S W,'see 1E*4 S-6 21-S 2S Manganest-54 0 see 21.3 912.2 9EUZ 41-7 lE-9 2E-8 31-S -3E-4 3E-7 IE-9 25 Nanganese-56 0 see 5i 2E-4 21.4 6E-6 1£-S 7E-S *7E-4

                             .,

• W, ste 1.3. 2E^4 9E-6 31-B.

B4 ODCM Rev. 25

App. B PART 20 STANDARDSIFOR PR AGAINST RADIATION I App.-.El

                                                                                                                                                                                                     .1II atilei4                                 Table 2               Tabl?3
                                                             --                                     occupationa1 Value;                                Effluent               Releases to Concentrations                Sr Co1.1             -. Col. 2.               ol. 3         Col 1            Col. 2 Oral.                                                                                  Vonthly IngCst.on                  tnhaltia"      P       .        i
             *Atomic '14adionucl~de,                Class                              'ALI . -'               A~.                OAC           ll                w  tr      trin       10n
                             *ISM                                           *             "r)         -      e~):,       . (t§I
                                                                                                                              )         (   00/01 SPatal}                    (g)    vC1/ 1) 26    ..       rlonS2               P . all csmpounds except
                  *    .                           those given for V                            9E2  ,-3E.3                       It 1-6                 04            1tf5         11
                                                                                                                                                                             'I-4     .
                                                  -V. olfd.S.-hydroxide.,

and hli des 2. z IE-S .31-9

                                                                                                !P',2101                                   - 3It--9              1E-4          1E-3 26               C"o-SI0.                 se.

52 F. - A 4 E. Z t5 $St9 5 14E2 3£#Z *-11-7 5£-la 1E-4 26I on- 9 V.- a.. Fe Sf,?

• 17 ' 7E-20 52 1i-1.0 26 ro - O0 a..

set .. 3E-g 9£-i? . 4-7 . 4E-6 W:sa Ft 2E11 atB -, 31-n 27 Cobalt-ISs ,~ all c.o,,onds excet, thoe. givir( for V 41-9 2E-5 2E-4 (_ . Y* esfdes. hyslroxildoe halfdet, and nitrates '3E+3 IE16 II 27 Cobalt-5S W. a.. 3t--O 4E-10 - 6E-6 SE'? : 3E#2 -1E-7 6E-S I 2T. set -e 4E#Z 2E4Z at-a II 2 C-abalt-57 W. se St.)' 3E+3 *E16 61-4 II 4E .) 7 1*2 *. 3E-7 IE-7 SE-4 6E-3 27 Cobalt-5ft -W a. 5 .E04 U-S5 7E-3 2fP3 U*#3 . SE'? 2E-4 2 IE*3 71.2 32-7 27 Co belt-50 W.. CE-6_

                 .                                                                                                                           *9E-S     -      2E-2

_ 2£-i

                                                                                               -3E46                            11-3 S-55 27            cobalt-So                                                                                                          2E-10             3E-6         31-S SE2 .2142                                7E£'
                                                                                                                            .1-S
                                             . V'     i:e SS                                                                                 SE-11           :-

2 I .(! 2I.4. 27 4 Cobalt-4l. 6E44 3E-S 3E-t 3E-4 3E-3 T: et CoFQ U14. 6E44 . 2E-5S! BE-S - 27 C ob alt-U v. Ie .C 5194 . 2E'S 7E-S 3E-3 st.uaell.

                                               'rY.a.                                              .      2E+4                 SE-S           SE              7E-4 2E-7              -
                                                                                                                                               -     .        ZE.4 2d              N1ckeS                . all compowds excpt to..given             for V.,,         LE4                  2143                5GE.7        3E-9              2E-5          2E-4
                                              'V. exaIds. hydroxidaf.

and carbidos I

• 11*3 SE-7 2E-9 Vapor.. * - 113 SE1-7 2E-9 28 Pickel-Si D 56 7E-9 2E-5 2E-4

                                                                                           -E#           SE43                 2E-S          4E.9 Vapor                                                                                         *E-9 B-5                                                                                    ODCM Rev. 25

App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B T ole I Table 2 Taber 3 OccupationalValuga Effluenft 161*8asI to Concentrat1n ls - ts Cot; I Col. 2 Col. 3 Col. I Col. 2 Imestion Inhalation Averac Ataft Radifoniuclfde Class ALi - RI - t Air Water -Corncnitratfen he. (tict) (PCQ) 00/1fZ) 0p0/10l} (Pi~tiffill NVI) t) 2a Nickat-6S 0 see StNI 2E+4 4*E3 2E-6 SE-9 3E-4 3E-3 I Y. sot 56i -7E*3 3E-C lE-O8 V2p.r - 2E* W-I 3E-9 - - 28 NIcWi1-6S of see 9E*3 2E*3 7£-' 2Ef- 9 1-4

                                                                                                                                           *-3 Vapor                              -                 86.2           3E-7        l-9            -

2s Nickel-65 0. see 65614 2.E4 - 1E-S 3E-S 1E-4 1E-3 W. see SMI 31.4 1E-S 4E-I - - vaport - ZE4 7E-6 2E-C 28 ' Nickol-66 0. ar cE2 56156*2) E3 7E-7 Z£-9 - (31.2)  : - - - 6E-6 SE-S W. $W MMI - F2 3E-7 XE-iD - V por -14 IEI E-C 4E-3 LL. wall 2 is Copper-GO O. all compuds oexret try S W" for W st t 5E4 ff 4 4E65 1-J St. wall 3E 44 - * *-4 *E-3 W. utIrfdn. halides. I'

.,Ia5 amd aItrata

r. oxides and " idus IE+S 15 56-5 4E-5 2E-7 1-7 - -

I- 29 Copper-Il D, see 6CM lE*4 3144 11-S. 4E-I 2- 2E-3

• a Y, *" tCn 4E44 2E-S 6E-8 - -

Y sto -cm 4E14 16-S SE-S _ 29 Copperi4 D see e IE* 3E 4 21E5 4E-8 2E-4 2E-3 Vi 6 1CM ' 2644 f-5 39E-9 W: te C - .ZE4 9t-6 51-5 - 29 Copper-67 0O ,e 6iCsa sE 843 E -4 16-8 63-5 1E-5 s eC so SE43 .2E-C 7E-9 - Y. se C4 - SE43 2E-S6 -9 - 30 -Zln-62 rY. a11 counds 1E13 1E63 1-6 4E-9 ZE-5 ZE-4 2 30-. Zlnc'-4 Y. all 2t14 76E4 IE-5 St-& - - St. wall (36E4)

• _ 31-4 3E-3 30 Zfnc-6S Y, all cwposad$ 4E12 3E.2 1E-7 41-10 5E-S S-5

             .30      z Inc-69            V. 11 c         qo.d,         44 3               2EI3            3E-C       SE-S          CE-S         CE-4 30-      Zinc-S92            Y. all comuis                    66E4                1E1             WS-S. 2E-7            864        OE-3 30       Zinc-71a           Y;all cr                          {ourd 61.3               ZE.4           71E-        2E-1            8E-5       BE-4 30       Zinc-72            Y. all c           _s              113               It.3           SE-7        2f-1            1E-5       1E-4 2

31 Gallifin65 0, all copounds except thet give. for V 21-S 2ES 7E-S 2E-7 - -

                                        ,                                  St. - a9ll (6-- . )                        -             9E-4         9E-3 V. ealsde.Jsydxdes'.

cr-bide*$, halides, ard nitrates 21+5 SE-S 3X-7 B-6 ODCM Rev. 25

App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App.B Table I. Table 2 labae 3 ccupational Values Efflu ent Releases to Concentrations Savers tCol. 2 Col 3 .Col. Col. 2 Oral Hanthly Ingestle n Inhalatitn Average Atomic Radienuclide "Class ALl ALI DWC Air Water ConcentratIon N.o. (pCi) (pCI) (pCI/ml) (pCi/al) (pCI/el) IpCI/ml) 31 ' "Mum-66 0. se ,65C 1E*3 4E*3 1E-6 SE-9 IE-S 1E-4

                                   *

• eee G 3E43 1E-6 4E-9 -

31 Callfim-67. 7E+3 1E-4 6E-6 2E-8 1E-4 1E-3 V, seea G 1E-4 4E-6 lE-8 V. .. 31 Gallium 682 Oe65c I. et 656.G 2E+4 4E-4 2E-S 6E-e 2E-4 2E-1 2.4 SE+4 2E-S 7E-E - 2 31 Callium-70 *O. seta CS 0i .a SE 2145 71-5 iE-7 - St. wall VE*4) - - , - 1E-~3 1E-2

• se 65X 2E.S SE-S 3E-7 -

V. ago 65c 31 allim-2 1E13 4E;3 lE-S SE-1 21-S 2E-4 3E13 1E-6 41-9 - 4 , 3 31 Galliu -73 . sea 65Ca .,I .

                                                                  '7SE473          2E.4           6E-6         2E-t          7E-5       7E-4 Y.

0 sCa 2E14 6E-li 2E-.2-32 Garani 4 0. all copounds except those given for V

• 44 3E+4 lE-S 4E-S 3E-4 3E-3 W, oxides, sulfides, and halides 2E*4 ISE - 3E-S -

72 0S ag Ga *M3.4 5E+4. 4E-5 1E - St. wall (41.4) -- - ~ 6E-4 6E-3 W. see "Ge 1E5 4-E-S 1E-7 - 32 t n lj-6 51+3 4143 2E-6 SE-9 6E-S 6E-4 - Vase C . 1E*2 41-5 1-20. - 32 G r a i mB *1104 2E*4 SE-li 21-9 2E-4 21-3

                                     . se        C"a                              3E+3           3E-l        1E-3          -

66

    - 32        C avuanlffn-71 1 Coran1-           D soxe G                         51*S          4E.5           2E-4         6E-7          7E-3       7E-2 4E-4           2E-S         6E-o          -

32 Gern1_77 52 C i66e 51E4 41.4 8E.4 3E-5 1-7 - 0.1.. iCe- St. well (7144) _ _ - 9E-4 -E-3 51.4 4E-S 1E-7. -

                                    ,see        Ge    *    .

32 Cvn1a7S 21.4 1E+4 4E-6 1E-S 1E-4 1E-3. V. . .a - 6E.3 2E-6 - -E - 32 CvU V. _e 2E44 9E-6 3-4 - St., _A11 (21E4) _ * - 3E-4 3E-3 W, s a-: 2E.4 9E-6 3E- - ODCM Rev. 25

APP B- PART 20. STANDARDS FOR PROTECTION AGAINST RADIATION. PP. B Ta7le I Table Z Table 3 PI

    ./i Occupatlsnal Values                   Effluent          Releases to Conecratiorwns           Swes tCl. I         Cl. 2         CDl. 3     Col.           Cl. 2 oral                                                               monthly I

Ingestion Inhslation Average Atomic Radlonuclide Class ALI L Air Water Concentratlon No. . (pC1) (yCi) (pCi/8l) (pCi/al) (pCi/al) (pCiV 1) 33 Arsanoci"92 V, all ceommonds 3E-4 lE S SE-S 2E-7  ; (4E-4) - ,_ _ 6E-4 6E-3 33 Arsenfc-702 v. all c~omon~s, IE4 5E-4 2E-5 7E-8 2E-4 2E-3

               .33        Arsenic-7l          W. all comounds              4E#3             1#.3       2E-S         t1-9          SE-5       SE-4 33        Arseoic-72          Wl,all Ceapoands             9E12          11E3          6E-7         2E-9          lE-5       11-4 33         Arsenic-73          W. all c~ompuds              9E13          21.3          7E-7          1E-9        11-4        1E-3 33         Arsenic-7i4        W. all comounds              11E3            3E12         3E-7        11-9          2E-5        2E-4 33        Arsenfc-76          W. all c~ompods              IE43           IE 3            E6-7      2E-9          IE-5         1E-4 33        Arsenic-77          W. all coampuds              4E13       .5E3              2E-6        7E-9          -          ;

(SE,3) - - - 6E-5 6E-4 r VI0 33 Ars~sic-itz W.-all c~ompouds 81.3 21+4 914 3E-3 3E-4 1E-3 c 0DI 34 Selefltw-"2 0. all coempuds. extept those given for W 21*4 4E14 2E-5 5E-e IE-4 IE-3 W. oxides. hydroxides. carbides, and alseental So E114- 4144 2E-5 6E-6 6E14 2E5S . 6E-S 2E-7 4<-4 4E-3

34. Selsniau-73 0, ss 7SO 3E14 1E5 6E-S 2E-7. - -

W.'see Se 3 1E-4 SE-S 2- 4E-5 4E-4 34 Solonuiwil 990sq 7S 2E14 7E-6 2E-A - - 70 1#-2 71.2 3E-7 11-9 7E-6. 7E-S W.see St - 5142 3E-7 SE-10 - - 34 Seleftium-7e 0 seso 6E12 81E2 3E-7 lf-9 8E-6 8E-5 W, eq S. - 6E-2 2E-7 lE-10 - - 4E14 7E14 3E-5 9E-3 3E-4 iE-3 2E-4 7E14 3E-5 1E-7 - - 0 34 Sajeftju.-U2 0, $00 " So 5144 21.S 9f-5 31-7 - - St. waill (8E-4) - l-3 IE lE-2

                          . . .             W. m"                     -                21ES           1E-4       31E-7        -          -

34 .Seltsair-~332 ', _71s

                                . ..         V. see70so               4E14             1E-5          SE-5        2E-7         4E-4       <E-3 3E-4             1E15          SE-S       2E-7           -         -

B-8 ODCM Rev. 25

APp. B App. 1B PART 20 -STANDARDS FOR PROTECTION AGAINST RADIATION able 2 T~b able 3 Occup~ational Values Effluent. Releases t Concentrations Sewers

                                                                          *~Col. 2        Cal.2             Col. 3    Col. I     Col. 2
                                                                   - .. Oral                                                             lmothly Ingestion           Inhalation                                   Average Atomic    Radionuclide      CIA$$                               ALI          -Ifl-                 Z ' Air           6ater      Concentration Na                             .              0cf)          (Mi)i          (0j0/0) (PCI/al) -.(jjci/nl)       (1,Ci/ml) 35        Sramlne-74om      0. bromides at M. Li.

Ka. K, Rb. Cs. andFt --- 1+44 4.44 i1-S SE-S

                       - .                                                 St. wall
                                                                         .(2E+4)-                                                3E-4   . 3E-3 Oldas. S7.0.y. C'4."S~r.;

Ra.R.A1.6Ga, in. 71. C.Sn. ft. As.3b.,si. F:..Ru: Os. Co. Rh, Ir, iii., Pd.' Pt. Cu. Ag. Au. In.' Cd. Nog.Sc. Y. Ti.. Zr. Hf. V. Nb, Ta, n.' IC, and Be 4E#4 ZE-S' 6E-S 2 35 blanine-74 0. seem,; 2F*4 71.4 3E-5 11-7 St. well (4E.4) 5E-4 f1-3 Vi, see 7% 61.4 .;E-5 IE-7 SE44 21-S 7E-B.- St. wall (4E.4) 5E-4 5E-3 35 *Srooine-75 D. -See74i 514 2-5 71-8 74r SE*3 21-6 71-9 SE-S SE-4 i sea 041. 2E-6 6E-9 35 . roeIne-ii. D. Siat 2E1-4 2E-3 21.4 11E-5 3E-S II 7 2144 71-5 2E-8 31-4 31-3 3SIrwinw-"lee %r 2*114 S16: ZE-S 2*215 K1-5 3E-7 35 Broeine-S&u a. a.. -% St. wall

                                    *    ,.sma 4r                                                                               11-3        11-2 21.5       *. 9E-5       3E-7 35        Sromine-se        V    seae'N

• 4E.3 21-5 61-9 4E-S ".4E-4

                                                                                       *441.3              2E-6      5E-9 Bromin 842      W; see 74m~r 52.4       *.3E-S            91-I St. wall (71.4)                                                  91-4        91-3 35                               see fi1114          r                                 61.4              31-S      9E-S~

W. see 2E#4 St. well 614 2-5 -81 2 (3t.4) 4E-4 4E-3 35 Irypmin-74 S4te 74io~s 6E.4 31- K1-1 3-a 2E-35 Irytan-74 suer a 6 4E-8 36 Kryptms-76 k Iasol 4E-6 .2-a

                                                                                                                                                              . I
                                                                                         *    .           2E-5     .72-8 36 Krypton-S1          14knrs lan                                                 -        7-4       31-5 B-9                                                                 ODCM Rev. 25 B-9

II App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Table I Table 2 Table 3 Occupat1enal Values Effluent Releases to Concentratiens Sewers

                                                                    - Col. I          '.Cel. 2         Col. 3      Col. I        Col. 2 Oral                               .          -fnthly Ingestion           Inhalation                                      Average Atoeie     dlod1nuclide . Class                             ALI                   Xt   10'.              Air           Water    Concentration No                                                           (pCI)               I(pC)    (pCi/ml)    (pCI/0l)        (pCI/2l)   (OC/ul) t 36         Kstart-eL-2 Subterslgn                   .    -                               iE-Z         5E-5 38        Kryptorn,-3        Subeerslon],                                      -               2E-S         lE-7          -

36 Kryptont-4S Submers ioni IE-4 7E-7 - 2 36 Kiypt~cn-.7 Subeers ion]. - . SE-6 2t-8 - 38 Krypton-IS Submars ioni - 2E-6

• 9E-9 2

37 . Ruidfui.79 fl* all compounds 41.4 E1.5 #E-S 2E-7 -

                                                                 -       St. wall (6E14)                                         -                 IE4 eES7       UE-3
         .37         Ruhfdiuain-B1     0. all C, P- ds                   2E+S            3E.5             11-4        St-7          -
 ;                                                                       St. wall                                                              0.2 (3E1S)                                                           .-3     4E-2 2-5                        4.

3? Rs*Idust-6 O..all cempcunds 4E*4 SE+4 . 2E-S -4 41.4 7E-8 SE

         .37        Rubid~iuS82          0.llco"o"da                    .1E-4           2E+4             7E-6        2t8            2E-4      2E-3 11.4 D          37         "I*diu.83         0.11 Ccoimpud5

• 4E-7 lE-9 SE-6 9E-S SE*2 37 Ru*liusm,44 0. all compounds 3-7 11-9 7E-6 7E-S 37 Rubl*Iimr.06 0. all-cood SE+2 3E-7 11-9 71-6 7E-S 5:.4 37 subi*dai-87 0. all C~o~mF, 1E.3 2E#3 .6E-i 2E-9 If t-S l1-4 2

37 l.Idfgw88 ,. D. aul c~oupatd 21+4 $E+4 3E-t 't-! ' - St. wall (3E14) _ 4E.4 4E-3 2 37 mhldlue.89 0.* al caosmd 1E-5. U15 .2£-7 - St. well (6E*4). 91* .4 3E-3 38 Stvoastlm-act 0. all sol uble comcuids Imce~t StTiOs . 11.4 5E-6;, 21-S 61-'5 6E-4 Y. all inseluble Cos 1E+4 39 Stroatlinm-81 U. see 50 r, 3E14 81.4 3t-S 1E-7 .3E- 3E-3 Y: t see-S 2E14 8tE4 3E-5 1E-7 38 Stvonttsin S V.. M  ! Sr, 3t. V 4E12 2E-7 6E-IG Ul wall (2E+2) 31- -6 3E-S yv- see o. 2E12 9E1I 41-S IE-10 38 Stroatium83+ Di, got ;Sr 3E#3 7E13 3E-6 1E-8 31-'5 3E-4 T. see Sr 21.3 4E43 lE-ti SE-9 11- . . ..- 21.5 6 1.5 3t-4 9E-7 31? .3* 3E-2 4E-4 1E-6 38 Stranttsm-es . D :,see 50 5r-YPsee Sr. 3t-3 3E+3 lE-6 4E--9 4EjS 4E-4 38 Strontlum-SS- 0. see s - 2E-3 6E-7 2E-9 SE-4 115 SE-5 2E-7 6E- -4 6E-3 38 s* S 4E+4 2E15 6E-5 2E-7. B-10 ODCM Rev. 25

App. B

• App. B PART 20 STANDARDS FOR PROTECTION.AGAINST RADIATION Table 2 Table 3 Table I Octupstional Values r Etlluent Concentrations Sever to 05nlas Col,' I Col. Col 3 Cl. I Cl. 2 A veAwrage m

    --            tucRdeul Atomic    Radiomzcllde.

ts;^In~rstion class AL ALI Inhalation DAC ir Water Concentrattol

                                                           '          *'   *(pCI)           (pCI)      (PCt/l)      (PCi/ml)       (PCt/l)..(PCl/l)

1. , .lb. "

6(42 5E42 4E-7 E-9 38 Strontim-89 D. see 50Sr LLI wall E8-6 R1-5 e 7 Sr(6E+2) - 1E42 6E-B . 2E-10

                                           ,5E2 38        Strantfum-90     D. so* 'Sr                        3E41          21.1            OE-9   ..    -
                                                                   - -      Som surt       Sone surt (4E-1)         (2E1I)                        3E-12        5E-7      SE-6 4E1O           2E-9         GE-12 2E*3            61.3          2E-S6           E-9        2E-S      2E-4
                                              ':seeS SOS 4E.3           1E-6         5E-9 31,3           91.3           4E-6         1I-S         4E-5      4E-4 38        Strmntfmm-2z      0. %t s. r 7E13           3E-6         9E-9
              -39       Ynrr"_iuwu                        ds except                                                                 3E-4      3E-3 those givtfIr.                                  6E14           ZE-5         RE-$ .
                                                         'I 7.. oxides one                                  S1.4                        SE-4 hyrel. .'       -

I E*3 3E13 11-6 5E-9 2E-5 2E-4

              -39       vutriwa-56        V. to ce -

3E-3 11-6 5E-9 3E13 S1-9 3E-S 39 Yttrlow-57 V.0 see 86 2E#3 SE-$ 3E13 IE-6 21-7 I3E-10 11-S .3E-4 yttwim-88 V*st a 2143 3E#2 39' 2E#2 11-7 3E-10

 -I 21+4                        2f-G         IE-4   *1E-3 39        Yttisuw-90*-*    Vse B'6N                                                                     ZE-M Y:.se 86..,                                     11.4 IEO3 4142           71.2           3E-7         5'-10 39       Vttrium-90        v. so.all LIl wall                                                          JE-5
                                                                      *   (SE.2)                                        21-S         7E-6 Y. See, Mar                                      6E+2           30-2                                      11.            21.5           2E       3E-7         ,ZE-3     2E-2 39        Yttrium-916       W. s:

21.5 2E-7. 2E42 2E-lo Y. a.. S. wall. 7- I (SE642) ASE-S

                                                                 -          1E.2S          2E1.           51-S .       21-10 4E+4                                        2Elo
                                                                                                                      .11-11 W, eegg.6                                                     .41-6 ,                    .4-5       4E-4 31        Yttrim-92                                                          8143          3-1-S         1E-a W. S, 86                                                       21-6          4E-g                   2E-4 39        Yttriws93                                                                        U11-          3E-9 Y. see                           S*.L wall.
                                                                         ,t6E)2)
              .31       Yttviw94        ~ W. a".                                          81+4 ZE*5           31-S          1E-7 4E-4      4E-3 Y.Ma                                       81.4           '3E-S 1*45 Yttrimr95 2

sae' 21E" 61-i' 2E-7 39 2C-7 4E-4 7E-3 Y. st " GE-5 2E-7 B-ll  :- ODCM Rev. 25

App. B PART'20 STANDARDS FOR PROTECTION AGAINST RADIATION. II TableI , Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Cot. I 'Col. 2 Col. 3 Col.. 1 Col. 2 Oral monthly Ingestion lalatf Average Atawlc adlanuclide Class ALIl DF Air Water Concentratlin

                 .o.                   .                                            (,CI.             (PCI)          (YCl/al       (pCI/mI)           (PCI/01)  (MCI/al)

, 40 2irconiN86 D. all CMpoDa exeDpt. those given for V and Y IE13 4E13 2E-6 6E-9 2E-S 2E-4. V. oxIdes, hydroxfdes; halIdes, and nitrates 3E13 IE-6 4E-9 Y. carbide 2E13 E C6 3E-9 8 40 Zirconfi-88 0, se 6r 4E13 2E12 9E-8 3E-10 SE-5 SE-4 2t.C. , Z' 7t-su * - I 3E12 1E-7 4E-10 -

• 40 21Irconliiu-8S D. i,at i 2E-3 4E13 lE-6 SE-9 2E-S 2E-4 trc~nt: °.set Mi 21E3 IE-6 3E-9 -

21E3 1E-6 3E-9 - 40 ZV r: - E*3 6E10 3E-9 Bone surl so Surf .. (3E.3) (21'4) , 2E-11 4E-S 4E-4 9 _see 6Z, 2E#l . IE-S ~. Done smof ' 6E1) 2-8 9E- -i

                                         ;. i.se       UZI                                           6E#I     t-        2E-6,8 Bane- surf I.-                                         O'.sa        Zirr *     '            1- 3               11.2               SE-C          -                 2E-S     2E-4 F_

Bn surtf (3t.2) - 4E-10 - 40 U rconf a-9S 41.2 2E-VJ. .SE-lO - 3E*2 1E-?7> 4E-10 rnUZr to 6E#2 2t1 E-. - . 3E- 9f-6 915S I+1 61-7.. 2E-9 - 40 21ioi u-9 11.3 SE-' 2E1: - W.* all ctound5 except .. ." " those glveewfor Y 2E1. 9E-5 3E St. well (71E.4) 21. 9- 3 lE-I 1E-Z 2 ., oxides said *1drcmIdes 2E+S 9E 3E-7

            *41       NlobiUs.88 IE 4               41.4                             S1-a              IE-4

_4E*4 ..2E-S 51-8 SE-S; In miun) 5E43 2E.4 8E-6 71-4 ZE#4 31-8

41. Vloblii-90 Y. set no31 IE13 3E-3 11-6 . 7E-S.. 2E-4

_ ,. 2E13 41E 41 11180600-910 81:7I.- 31-9 7E-4 V.see 9E13 2E.3 IE-S,. LLI. well (IE4) -. ILSE . 2E-4' . ....

2E*2 31E-4 41 11ebtitme-94 9E.2 2E12 IE-S

                                                                                            -,2E#1                                    S1-12 41        Niebfum-95a                                               2E13               3E-3                              41-9 LLI well (21E3)               -                                                  3E-S
                                                                                -                 2E53                9(17           31-9 B-12                                                                                   ODCM Rev. 25

App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION . App. B. Table 2 Table 2 Table 3 Occupatfona1 Values -Etflunt Releaes to Concentrations Severs Col. I Col. 2 tCl. 3 Col. I Col. 2 Oral Mlonthly Ingestion InMAUltias Average Atomic Radionuclida Class ..- - ALI Air vater Concentration nt. i, (WI) (PC0.) (PC,/ml) (0Ci/l) (pCf/al) (Pci/el) 41 Nieblia-95 W. set 2E13 14E3 SE-7 2E-9 3E-S 3E-4 V. see -E13 SE-7 2E-9 - 41 Niobium-96 W. SeeM 1E13 3E13 IE-6 41-i 2t-S -2t-4 Y. see ~ 2E*3 1E.6 3E-9 41 *Mio hism-972 V. se 2E-4 BE44 .3141 3E-4 31-3 7E44 3E-S lt-7 2 61 41 N~iob~iuo 98 W, Se" 1144* 5.E+4 2E-5 SE-8 2E-4 2E-3 SE+4 2E-5 7E-S 42 'Holybdenvw-90 D. all csompuds aescopt those given for Y 4143 7E*3

                                                                                       .11.4           3E-6        lt-8           3E-5       3E-4 Y. oxidei. hydroxides.
                                                                        *2143            SE*3         2E-6         6E-9           -

42 Molybdenumn-93a D 110- 4E*3 .2E*4 7E-6 21-B 6E-S 1E43 6t-4 tE-6 2E-8 - 42 polybdens-93 D' to ~~ SE3 2t-6 SE-9 SE-5 SE-4

                                                                       .2E*4            2E.2 21.4           WE-S       2E-10 42             Molybenum-"        D. see 90Mo                   2E*3           3E13

• 1145 .1E-6 41-9 -

LLI well. (2E*3) 21. - - 2E-5 2E-4 2E3. 6E-7 2E-9 2 3143 42 "DI dema-01 V2. W mN 4E*4 1E+S 6E-5 2E-7 - St. wall (51.4) _ S 71-4 7E-3 6E-S 2E-7 - 2 43 Techntmatum-93m 0, all coempowmds eacept those. given for W 7E*4 21.5 W S- 21Z-7 i1-3 1E-2 W, ouides, hydroxides, halides, aad nitrates ,3E+S 1E-4 4E-7 43 'Fechmotilaw'93 0, . 31.4 341-5 1E-7 4t-4 4t-3 1E-7 43 lechnatiew91462 I. ~C MP 1.+4 SE-S - 3E-4 3E-3 21-5 E1-S 21.4 43 lectsietis -94 U, ea ~ C 9E.3 21.4 6143 3E-S 3E-S 11-4 IE-3 41.3 SE24 IE-9 - 5E-5 5t-4 3 2E+3 RE-7 .3E-9 43 lechnetiimi-ss 3 e TC _ *5 2E+1 IE-3 2144 - 51-6 3E-S

43. Techenetium-US6 Da' 'T 43E-i W:, see c . 2E*4 2IlE-3 2E-2 1.3. 11-4 4t-7 35-7 4

13 leclinetiwus E U. s -eTC 2E#3 3E*3 IE-6 5E-9 3E-S 3t-4 2E*3 3E-9 13 A lechnsetiumrS78 D. so*e9-`TC SE-3 7E43 :3E-6 6E-S 6E-4 St. well 71.3) 11-s W. see 93Tc 27+3 SE-7 21-9 I 13 '-- ODCM Rev. 25

I I Al )P.B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION 4pp. B Tabe I T- Tabl* 2 Table 1 Occupattonal values Effluent Relnases to Concentrations Sewers Col 1 Col. 2 Col. 3 Col. I Col. 2 Oral. Monthly lnsetion lnbalation Aaerege Ato It ladionuclfde Class _I -ALI OJ Air Water Concentra ab(PCI) (Ptl) tpt1f/ol) (PcIS l) (PCIM") (PCI/ l) 9 43 Tecbetium-97 D ne C 41E-4 SE44 2E-S 7E-B SE-4 SE-3 W

                                                 -      to 'Pc                        -             6E43          2E-6        BE-9 43       Tech"Mtium-98        D see "933C                          113          2E14           7E-7        2E-9            11-S           1E-4 Y    *n       3Tc                   -             3M42           1E-7        41-0'          -

9 43 Technetfwur-9  ; nee Tc SE*4 2E'S CE-S 2E-7 11-3 IE-2 V, se "tc 2E'S 1E-4 3E-7 43 Tchnotlum-" D* e 93eTC 4i.3 SE43 2E-6 - 6E-S 6E-4 St. wall se -E6*3) BE1-9 - -

                                                -     see                           -3%le          7142         .E-7         9E-10 2

43 Technotiu_-101 D. see Tc 9E.4 3E+S 1E-4 5E-7 - - St. wall (lE ) - E-3 2E-2 Y x* s 93-"- 4E+S 2E-4 SE-7 - - 2 43 Technotiwf104 D, ' 9ha* 2E14 7E-4 3E-S 1E-7 P St. wall 41-3 i I.1 W."a '3Tc (3E;4) 9E-4 4E-S.. E-7 4E-4 4-C1 44 Euthenium-94 0. all compounds xtept "I these give, for v and v 2E+4 4E-4 2E-S 6E-B 2E-4 2E-3 VYhalides 6E-4 , 3E-S 9E-8 r oxides nd ydrmxtd. - 6E(4 2E-S S-S 44 Authlani-S7 D s"e %V 3 7E+4 SE-6 3E-1 1E-4 IE-3 W: 4"ol - 1E-4 SE-t 2E-S - - Y. s lu -1144 1-S 21E-6 - 44 44 Ruthenium-103 auUthi-1arlS D* sn 94e V. bee 94 Y. *n #lu O, *ne Ru 21E3 SE.3 2E*3 1E+3 6E12 E44 7Ei7 4E-7 31-7. 6E-6 2E-9. 11.9 9E-10 21E-S 3t-S 71-S 3E-4 E-4 0* Wa"ne - 1E+4 61-6 21E-91 t - 1144 SE-6 ZE-B1 44 RutheriuarI6 0.Dxn R ZE.2 9(41 4d. E-6 IE-I0 ( 2142) - - - 3E-6 31-S W e 4 . 5E.1 2E-S E-I -- - r, 1- . U1. SE-. 2E-13  : - 45 thodiur9a 0. all Compomis except those give for W and Y 2E14 644 2E-S SE-S 2E-4 2E-3 V halides - 5144 3E-S 1E-7 - - YT,oxides and h)droxhids - 7E14 3E-S 9-4 - 45 thodiawr" see se .9i24E3 3E*3 IE- 4E-9 3E-5 3E-4 W: see9 - 21-3 SE-7 3E-9 . - Y see i - 2E.3 E-7 3E-9 - - .1 B-14 ODCM Rev. 25

                                                                                                                                          .App..B App.~1     -PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION
                                                                                ;Tablel                     Table 2           Table 3 Effluent       . lel*ssS to Occiattional Values                                     Seters Concentrations Col. 1         Cl. 2          Col. 3   ColoI        Cel.'2 ontLhly
                                                             'Oral                                                            Average InAstiin C           Inhalation ALI             -ALI          ZDC      Air          Water      Concntration Atomic       adinuclide    Classs                                  (PC)       (pCi/ml)  (pCi/l)       (pCI/ml)    (pCi/ml)
                                                            ; (PCI) o.

SE+3 2E-6 7E-9 2E-S 2E-4 Rhodil_-100 D* see a 2E+3

           .45                                                                4t*J          ztlb     61t-V* S*'"ii~               -              43E+           2E-6     St-9 0 a.ee.9m~               61.3-
                                                              -         1    1E'4
                                                                            .B              SE-6     2E-S      -  SI-S        8E-4 45         *Rhadium-10la                                                         4E-6     IE-8 Y. soee~                 -              51.3
                                                                            '!K43           3E-6 2E.3           5E12           2E-7     7E-10        3E-S        3E-4 4S'       thodium-101,    0 sme  e_"               -               51.2 2E+2 3E-7     1E-9'
                                                              -              2E-2           6E-S     2E-10 1.E3           5E12          2E-7     7E-10 45                        D    S"ee h 0,hdu-2i' LEI wall                                                        2E-4 VIC1£-1                                               2E-5 2E-7     SE-1O.-

Y. See 3h 2E.412 SE-S. 2E-20 4E-8. lE-10 S 5E-6 SE-S 7E-8 2E-102' 2 99 41E4 2E-S SE-ll 45 thodiun-103 0 see 3h , .U. SE-4 2E-6 6E-3 6E-2 SE-4 2E-6 SE-4 2E-6 SE-6 2E-8 Y: 'Se . 1E61 SE-5 SE-4 U- 9E-9 3E-6 9 4E-3 2E1.5 2tE-6 SE-g. 45 Rhodiumi-107 a0. Sa. 3*h 1E-4 1E-3 St.L wall 4E-8' SE-S 2E-S E1-S SE-S 1E-4 31-7 1E-3 1E-2 see 314 GE#1.4 11-4 4E-7 4W.ald~mO 1E-4 3E-7 45Rofn 0 see V0m -"&R -K* 31.4 6E-7 2E1- 2E-4 St. wall SE-7 2E-9 6E-7 21-9 lE-5 SE-S 2E-4 2E-3 lE-S SE-S 31.4xce2pt lE-S 4E-S 46 Falladlim1207 . ill 0ow 3E-6 9E-9 hyr V. aiesan Lids,wal 2iEey 1E-4 1E-3 00 Pd 3E71. 2E-6 6E-9 v see SE-21-6 1 643. raldim24 .set 00d 6-3 9EG

                                                                                             -        3E-S        SE-4        SE-3 3E-6     11-8 2E-7     6-10 I

B-15 _, ODCM Rev. 25

Apr- B. PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION. App. B

                                                                                                                                                                        ' W Table I                             Table 2         Table 3 Ocuepationel vaalus                         Effluent        Releases to concantratlon.         Sewers Cl1            CO.2-Cl3Col.

I Col. 2 Oral Wothly. Ingestion inhalation Average Atouic Radlonuclide Class ALI Air. Vater Concentretien No. (pC0) I(NCI) (0C/) (RCi/al) (pCI/a) (pCf/al) 46 Palladfum-109 0. seteOf -2E43 51.3 3E-6 91-9 3E-5. iE-4

                                            ~         1 f ew OPd                                       SEW            ZEi-6     .      8-9

• Y, see s.E+3 2E-6 6E-9 2 ZE-S 2E-9 47 Sliver-c02 0, all compewds except 815. 21-those given for V and 'f SE*4 St. well (6E4) 6ES ZE-7 9E-4 9E-3 W nitratte and sulfides 2E#5 BE-5 3E-7 oideas and hroldes.

ox 21.5 4? Silver-13Z o, see 2:A SE-S lE-7 51-4 SE-3 SIS 2E-7' V. ee. A# lE-S SE-S 2E-7 2 47 Silver_-1O4 Da" -3E+4 4E-S - 1E-7 4E-4 41-3 IE+5 5E-5 21-' 21.4 1E+S C 0. o. 3E-S 1E-7 3E-4 it-3 I. W, se I 6E-S 2E-7 T. se. Ve. 10 2A AS lE*S V. see.AV, 3E*3 6E-7 .E1. 41-S 41-4 1E-3 2(*5. 71-7 -2E1 4? Silver-lOS D. see 102 ZE*3 7E-7 21-9 47Slve-IC. *..W. see ICi" 7E;'2 3E-7 1E-9 1f-S 1E-4 W42 3 0 2 9(s2 41-7 1E-9 47Siver-OS Y. see C A Vase 6E-S 3E-7 St. wall 1145 2-i (tSE4) 9E-4 SE-3 47 Silvir-1C0in C. see 2Aq 9C-S 3E-7 6E-S 3E-7 9F21. Y. :*ael Ag SIE- *. 31-10 9E-6 9E-S 43E+2 1E-7 4E-10 VS102h SE42i 2E I 1E-S 3E-11 47 SJiler-Ill D. seeIAg IE-2 SEI- 2E-10 61-5 *E-S Zf+2 6sE-S 3E-10 W.see St 901 4E-9 0 lE.rw 6E-7

47. Silier-110a D. 102 Ll1 wall (113) (2E*3) - 2E-9 2£-S 2E-4 4E-7 1E-5 9E+2 .4E-7 1E-1 31E3 BE*3 3£-6 lE-8 4I-5 4E-4 3E*4 4E-6 1E-6 9E+3 4E-6 11E-0*

B-16 ODCM Rev. 25

App. B App.pB PART 20 STANDARDS FOR PROTECTION'AGAINST RADIATION Table I Table 2 Table 3

                                                     -.                       CIccupetlonal Values                     Effluent        Releases to Concentrati0ns        Sewers Col. I          ol. 2          Col. 3      Col. 1        Col 2 Oral                                                              Imonthly Inrstfon            Inhaletion                                    Average Cltss                                          ,tT1'F                     Air          Water   tConcentration Atontc Radlonuclide                                                                                                         (PCif/l)

NOe. NOi) 102 47 Siler-lS 2

o. Se. Ag 3E64 91.4 4E-S 1E-7 St. well 2~02 *(3E*4) 4E-3

                                                               .Ag                   9E14            4E-5        IE-7 V. to*           AoO2                         3E+4            3E-5        1E-7 2

4S Cadmium-104 U. all coounds except 31E-3 those given for W and Y 2E*4 7E14 3E-5 S 3E-4 W. eulfidus. halids. and nitrates 1E+5 5E-S 2E-7 - Y. oxides and hydrexides 1E1S 5E-5 2E-7 - D. so 204Cd 21.4 SE4 2E-S 6E-I 3E-4 3E-3 48 Cadmlm-lO7 ItE-S - v 1 04 Cd 6E#4 2E-5 , V: e. Cd 5E14 2E-S 7E-S. - 48 Cadem-S109 .D. set 104a4 3E-2 4E#I lEt4 lKidneys 104 (4E12) [5E*1] , 7E-1S *E-C SE-S W. oe, C4 C 2E4Z SE-D& £- (51.1) .- 2E-lO I Kidneys II 12 SE-9 E-l - It I7 a 2.1, 41t Cadeim-1 Y. see 104Cd 241. Ki dnys SE-I (4E*l) (11.) - SE-10 SE-

• 204 V. see Cd

                                            ~*f104w se'.

xidneys (E+l) 2E*Q -9E-91 - E-l2 - 4E-8E+4 3E-9 2E *E91-10 -E 4 Cad lml3 0. see 104C lidneys 4E-6 (3E*1) *E(40) Kidneys - 5E-12 4E-7 W. a" 104Cd SE (11.1) 3E-9 6 -it21to - (11) - 2E-11 - 1 04 1.1 SE6-9 2E-11 Y. see Cd 04 3E.2 5141 21E- - 4E-6 4E-S 4U Cafti wo-l25l 0. se 1 Cd (E14) - lE-10 6 1st 204 Cd 1E42 SE-6 2E-1O - Cd 1E42 6E-- 2E-I0 - I Y. ee S1.2 1143 6E-7 2E-9 - 45 Cami a-~US D.6# 104 Cd IL! wall 1E-4 _ - 3E-S (11.3) 104 11-3 SE-t .22E-t - v m Cd 1143 6E-7 2E-9 - 2E-6 6E-S 6E-4 E+13 lE 4 SE-I 48 Ctdoi1m-17a . am 104 ICd 2144 7E-6 21-S -

                                         .see 104.1                                 11.4            6E-S-       2E-S          -

Y. se Cd B-17 ODCM Rev. 25

.. App, .B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App,.B. 7able I lTable 2 Table 3 Occupational Values- Effluent Roleases to Concentrations Soee Col.1 Col. 2 Col. 3 Cl. I Col. 2 oral Monthly Atomic Radifnuclide Class - ifnhalatfon Aft- Wster Conentret10n

                              .   .    ,*(PCl)                                                       (PCO)      (PCf/-1)    (Pcl/al)          tvcf/ol)    (ISCI/m1) 48           Cadmium-117              0, He INCd                            E43          IE14            SC-f       2E-t             6E-5        6E-4 W. gt 104Cd                                       2E14             7E-f       2E-              -           -

V. sae Cd - 114 1E-6 2E-S - - 49 IndifuID9 D. all corpounds except those given for W 21E4 4E-4 2E-S 6E-Sf 3E-4 3E-3 W, oxides, hydroxides, halides. and nitrates - 6E.4 3E-S 9E-S 2 49 Indiu 110 0. see lID9n 2E14 4E+4 2E-S 6E-8 2E-4 2E-3 (69.1 min) W. see In - 6E14 2E-5 E-S 49 Indlu_7flO 0 s In SE13 2E.4 7E-6 2E-8 7E-S 7E-4 (4.9 h) v. see In *2E.4 W16 31-S6 49 Indi10_-l 0D ee:91m 4E13 6E13 3E-6 9-9 6E-S 6E-4 VWsee In. - 6E3 31-6 9E-9 - - 2 49 IndiUlSUZ O se 2E1S E+In 6E+5 31-4 9E-7 21-3 21-2 W s3 In - 71 1E-4 11-6 - 49 Indiur-113Z 0, see l 0ogn SE4 lE-S 6E-S 2E-7 7E-4 7E-3 V W- s" In - 2E45 SE-S 3E-7 - - 109 49 IrdiLm-114e 0. so. In 3E12 61*1 3E-8 9E-11 LLI wall

                                                                                                  -                -          -                SE-1f       SE-S 309                   (1)               112             41-S       IE-10           -.           -

49 Indium-tSx s' 09 n 14 4.4E 21-S 61- 2E-4 2E-3 4S IndfulSS D Iw5°~n 4El E+6 E-10 2E-S2 SE-7 SE-6 W se In - SEfO 2E-S 7-12 49 Indum-S162 0 see 1 4In 2E4 1E4 3E1-1 11-7 3E-4 3E-3 W: sne 1In - gSE+ SE-5 2E-7

         *9         JndluoSU7.j2               D xn :      5099ln                 11+4          3E;4             lE-S        SE-8           2E-4         2E-3 W. se          In                  -             4[+4             2E-S        S-I            -

2 49 nd1fu-3357 D xnsee Iln 2E44 24S 71-S 2E-7 1 8E-4 3E-3 W *s In - 21+S ff-5 31-7 2 49 NfIm12179. P. se SIn 1E.4 S1E4 15-S 2E-7 St. wall

                                                               .               (SE+4)           -               -            *t-4                       7E-3 W. se  u1n                        -E5                             61-S        2E-7 So         Tim-12                     0. all c            qomds
                                                                      *XcVpt those given for v                 4E13           1E+4             SE-6        2E-f           S-S         5E-4 W, sulfiden. oide.

hlydrexides halides nitrates, nd stannic phosphate - 1E4 5E 2E8 - - 2 so Tin-i. D s*n l10Sn 717E4 21.S 9E-S 3E-7 1E-3 1E-2

                                                ,   sec      Sd                 -               3E1S             11-4        4E-7          -

B-18 ODCM Rev. 25

App -PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Table I Table 2 Table 3 Occuoattonsl Values EfflueVt Releases to Concentrations Sewers I Col.-1 Col. 2 Col. 3 Col. 1 Col. 2 Or-aI . monthly I Vestion Inhalatin Average Atomic Radionuelide tiCss AU ALIE 7 'r I Water Concentration NO. - (UMI (PC1} (C1

                                                                                                            )      (VCl/SlY        (pCi/a l) (Mp/l).Cfth).

110 sC Tin-113 0 see S 23 1f+3 #-7 2E-9 LUI wail M(212) - JIE-S E 3E-4 Y* n n - - SE.2 2E-7 SE-10 10 C 71n-117* D. sea 2 Sn 21.37 1£.3 SE-7 ULI wall Sone iurf 0 50 Tin-Ull% Y. 0 tee n

                                              . 210
                                                  'n                        (2t13) 31.3 LtL wall (2E-3) 11.3 21E3 6E-7 1£-6 3E-9 2E-9 3U-9 3E-S          3E-4 1 10                        (4wt)               -             *                      .        6E-5          6E-4 W, see         5n                       -                 21E3         4E-7           I-

50. Tin-il. O set 21OSn SE-Z 11.2 4E-7 -. lE-9 WU well- - -5 14 0 (4W3) -E-4

v. see 22 Sn - SE*2 2E-7 SE-10 so Tln-iUl D. sea110 6E13 2E-4 EE-6 2E-8

                                                                             * - wall -Ul                                   -     .                           .

(6E13) - _5 E-4 Y se Sn - 1E-4 SE-6 2- S 2 Dn so Tln-123% D S-n SE+4 11+5 5E-5 2E-7 7-4 TE-3 W. OR O Sn 1E1S 6E-S 2E-7 110 so Tln-123 0. swee Sn SE;2 61E2 3E-7 9E-10 LLUwall an ** (61Z)- -E1. 7- -fS V. s e -So 22 7£ 2E-10 - - so Tn-1US 0. set W1U 4E*.2 9HQ2 4E-7 lE-S LWwell 1 (1SE2) - - - 6£-6 6E-5 V. see Sn 4E2 1£-7 SE-10 - - so *Tin-U26 aD i n o' 3E12 6E+1 2E-S SE-11 41-6 4E-5

                                    -     a e 11 05SR                      -                   7E1o          3E-S    :       E-11          -

sC Tin-127 0 se . - 7E+3' 2E+4 E-6 3E-S SE-5 9E-4 V. se Sn: - E2E4 SE-6 3E-5 2 s5 Tin-S . ae 1 0Sn 9E.3 3E14 l 1-S 4E-S IE-4 IE-3 VWe Sn - 4E14 1E-5 SE-5 51 Antimny 2 0. all e h axiept those given tfor W #E4 2E S 1£-4 3E-7 IE-3 1E-2 o.axides, h.droxids.- , halides, sulfide$; - Sulftes. t nItrates - 3t15 1E-4 4E-7 51 Antial o-2162 'D. Si. Sb 2E14 71.4 3-S 1E-7 3E-4 3E-3 WV.s. " S - - 1E-S 6E-S 2E-7 - - 51 - Atil11 2 *D see 11S5 b . 7E+4 31.S 1£-4 4E-7

                                                                        -       ,al4)l                     -       -                     1-             1E-2
                                   -V      115si,                         -.                 3E+5          1E-4         SE-7             -             -

51I Ant ionoy117 0D se1f5M 7E14 .2E*S 9E-S 3-7 SE-4 SE-3 V Sea Sb .

• 3E1S 1E-4 4E-7 - -

                                                                                  ;...                                                             ODCM Rev. 25 B-19

I I App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION Pp. B. table I Tabli 2 Table 3 Occupational Values Effluent Releases to Concentrations Sevrs Col 1 Col 2 Cel. 3 CoL I Col. 2 oral .Monthly Ingestion Inhalation Average Atomic ladionuclidc Class ALI ALL DAC Air Vater Concentration No. (PCI). (PC0) 00~/0I] (IjCf/m) (pCi/ml) (PCI/Ml) 51 Antlmony-11" O sco 12S5b 6E13 2E14 tE-6 3E-8 7E-5 7E-4 SE+3 2E-4 9E-6 3E-8 - - W: set 1155b 51 Antimo1ny-1 0e115e 2E-4 SE-4 2E-S 6E-8 2E-4 2E-3 21E4 3E4 lIE-S 4E-8 - - 11E5 4E+S 2E-4 6E-7 - - (16 min)

                          ..              ':Vs.. "        i           S2tE vIl       _               _                          2E-3          2E-2 W. see. lSsb              -                 51S            2E-4        7E-7          -

51 Antimony In w115 1E-3 2E+3 9E-7 3E-9 lE-S 1E-4 9E+2 1E13 SE-7. 2E-9 - - SI Antimony-124st 0. sm 51 Ant1 122 0. 'sit 21SI 9E*Z 2E13 1E-6 3E-9 - - LLI well - - (8aE2) - 1E-5 1E-4 W. V.W**.. 115Sb Sb 7E12 11.3 4E-7 2E-9 - _ 51 Antimony 115b 31S . 81E 4E-4 lE-6 3E-3 3E-2 2E15 6E+S 2E-4 tE-7 - - 115 Vws. Sb 61E2 91.2 4E-7 1E-9 7E-6 7E-S it C' 5E12 21.2 1E-7 3E-10 - 1 15 D51 ntimony-125 U 5 3b 2E13 2E13 1E-E 3E-9 3-S 31E-4 .DL

                                                                   -                 SE12               1-7     7E-10.          -

115 51 Anty1 2 D. 3155b SE-44 2ES 8E^S 3E-7 - - 3W. 115 (71.4)  : 1E-4 9E-3

                                                                    -                2E1S            #-S        31-7          -              -

51 Antimony-126 DV*se 3 usSb Sb 6E12 11+3 . SE-7 2E-9 7E-6 7E-5 SE2 SE+2 2E-7 7E-10 - 51 Antim127 D0.St Sb 81*2 21E3 9E-7 3E-9 _ _ Wi wall' (5E12) - - - 1E-S 1E-4 W; Sam lIslb 71.2 91.2 4E-7 lE-9 - - 51 Ant1mr ol22 D. see 11S&b SE4 41E- 2E-4 SE-7 - -

                  -      (10.4 mIn)

• St wll (lE+S) -I E-3 1E-2 51.Ant mony-. W.0 see", U. S Sb - 4E-S 2E-4 6E-7 -

1 V m , Sb 11E3 4E13. 2E-56 .SE- 2E-S 2E-4 (9.Z1 h) W,". 11555bs - 3E13 lE-6 SE-9 - 51 Antimony-129 W 113 31-3 9E.3 4E-6 1E-8 4E-S 4E-4 Wo Sb - 91.3 4E-6 1E-S - - 51 Antimoyy-US D 11 Sb 2E14 6E14 3E-S 9E-8 3E-4 3E-3

                                                                 -                  5E14           3E-5        1E-7         -             -

2 11 51 Antimony-131 0. set SSb 1E+4 2E14 1E-5 - - Thyrold Thyroid (2E4) (4E14). - 6E-8 2E-4 2E-3 Y. set 115 5b - 2E.4 IE-S - - Thyroid

                                                                -                (4E14)            -           6E-S         -             -

B-20 ODCM Rev. 25

App. App. !i PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION: Tabl -Table 2 able 3

                                                                     --                    Occupat ow4 Values                         Ifft~iunt        Release to Concentrations         Sowrs Cl.I              Col. 2       l .I.)           Col. I      Cal.

Oral I t Monthly IngestiM Inhalation Average Atomic iadisnucllde Class AI 1 ALItT*1_F Air Water Concentration (-P l, tpCI)) (PCI/el) (pCi/l) (pCi/ml) (pci/al) 52 Tellurium-l16 l. allI4o-eunds except those given ter - BE3 2E*4 9E-6 3E-8 1E-4 1E-3

• V. axides, hydroxid end nitrates 3E-4 lE-S 4E-8 SE'2 2E#2. eE-8 -
• o Surf Done surf 52 Tellurium-12221 .S (7E12) 4(42) - SE-10 11-5 IE-4 2E-7 61-10

                                                                        ;        3E*3                              2E-6        6E-9          (.04      4E-4 52        7.lluriem-12m              0. see     2a                                      3E#3             lE-6        4E-9 61,z            .2E12             9E-S        -

rt Dane surf. (1E.3) (SE+2). 11E-5 W. Sea11 61. 2E-7 OE-10 116 52 Telluritum,43 O..set 2 Te 2E*2 BE-& ,

                                                                                          'f      gone surf CI                                                                           * (1E.3)                                           7E-10        2E-S       21E-4 L1I                                               V.SeeaX                                                          2E-7        -

I Bangsurf (IE-3) 4E+2 - -2E19 (1E3) 41.2 2E-7 _ (IE*3) ,r newsurf S2X- zelliriqm125 . Sa U 15* (1(*3) _ IE-9 2E-S 2E-4 7E42 3E-7 IE-9 61.2 31.2 1E-7 _ W. see117 one urf 2E^4 (4E42) - CE-IO 52 'Fellarrtm-227. -. see 1Sa S" 1E-7 4E-10 21+4 31-6 SE-B. 7E-4 W:see 41e 7E-6 2E-O 52 Tollurimor"Sm 3E-7 9E-10 7E-6 7E-S IE-3 3.see te

                                    --             Sa  114;ee T                                     6E#4            1E-7         3E-10
        .52                                  ID: see ..                        SE14                             3SE-5          E-O         4E-i 5.4f,           3E-S -       1E-7 52        1e'll~wim-DU~ 3.S see - I                     m 2E-7
                                                                                              - tityreld (51.2)          (11E3)                       :2E-            9E-5      1E-5 W, See 6T                                          4E-Z -
                                                                                             *thyroid (3E-2)            2E-7        1E-52       lelrsm11                     . Sea U6T           . I                                                                     S-S 5IE*4]

(6E+3) 21-S E1-4. 2E-i 2E-C -

                                                                                                       .('              , ,2E-S (IEr4)

I B-21 _ ODCM Rev. 25

II App. B PART 20 STANDARDS T N A D FOR PROTECTION AGAINSTIRADIATION.. PP. B I Table 1 Tabli 2 Table 3 Occupatfonal Values Effluent Raleases to Concentrations Sewers Col. I Col. 2 to). 3 CtlI Col. 2 Oral Monthly Ingestion lltalstio Average Atomic Radionuclide Class ALI tptl ( Air 'Water Concentration Wi.(CI) NOC) (1CI/1mi)(pi/10l).(pCi/al) (PC RI/ ) 52 Telluriue-132 -D; &a. TO 2E12 2E12 - E-S - - - Thyroid Thyrold llo (7E12) (tE12)  ; E-9 9E-6 9E-5

                                                                              ,                Thyroid (6E*2)                          9Ew10 2             16 52          Tellurfum-133         D. ea i     Tv                 3E43             SE+3               2E-6 Thyroid          Thyroid
                                                                           ,(6SE3)            (1E14)               -         2E-B              9E-S        9E-4 W. se*   l1SE*                                   SE*3               2t-6                          -          -
                                                                                     -         Thyroid              .          2--
                                                                                             ,(1E4)               -              E-6          -

2 11t 52 Tellurimm-133 0. eta 6T 1E44 2E14 SE- - _ _ Thyrald Thyroid (1 3 E4) (6E14) SE-S B 4E-4 4E-3 W. seto 2E44 9E-6 Thyroid

                                                                             ,-              (6E-4)              -          *E-O              -

2 11 52 Tallu.lw_134 0. set 6Te 2E14 ZE4 lt-S - _

                                                                  .           Thyroid         Thyroid Tc                                                                          (21E4).           (2EE4).      .            5    7E-S              3E-4         3E-3

¢ V, set ZUTO 2144 11-5 -. Thyroid. (SEY), M- - 7E-S - - 2 51 lidimwIZ 0% 0. all' cc' ou da1E.4 ZE214 9146 3E- - .I 53 iodtne-hO 2

0. all ompounds Thyroid (11E4) 4.3 S1 4E-6 _

2E-4 2E-I

                                -Thyroid                                                      Thyroid                                    -14              1-(dE13)             (11+4)             -            2E-               1-4          E-3 U3           lodfne-121            D. all compounds              114               24 7"le              Thyroid

_ (3E.4) (51*4) - 7E-6 4E-4 4E-3 53 .ldine-hI O. all com d 31.3 CE3 C 31-4 ThyrId Thyroid (144) (2.4) 2f-8 1E-4 1-3 53 lodift-124 D. all compoumds SE41 E41 3E-S - - - Thyroid Thbtid (21E.) (3E*2) 4-10 2E 2E-S S3 lodIftn25 0. alI c mpuds , 4E. SE41 31-6 , - Thyroid Thyroid 111.2) (Z'2) - E3-1-3 21-C 2E-S S3 lodine-126 0, all copounds 21.d- d41 Thyroid Thyroid 1 -6 1-(71.3) (UE#2) 2- 1E 53 Iodie a . all cound 4144 1E+S SE-S 2E-7 - St.-Wull (6E14) - - 8E-4 AF-3 53 lodine-129 O. all c qmpouns SEd 91E4 4E-1 Thyroid Thrd (2E11) (3E11) 4E-1 2£-7 2E-6 B-22 ODCM- Rev. 25

App. B PART 20. STANDARDS FOR PROTECTION AGAINST RADIATION App.B; I

                                                                             -                      7able I                      Table 2         Tab)e 3 Occupational Values                 Effluent       Rtlaases to

• : Col.1 Oral Col. Col 3 Concentttons Col.1 Col. 2 Sewers olonthly Ingestion Inhalation Average..

Atomic Radionuclide Class ALI ALI *DAC Air Wdater Cencentration 53° l- all co(wcd/a) - (dci - (OCf)_ (1c,17.) 00/ 82) S3 lodinr-130 D, all compounds 4E*2 7E42 3E-7 - Thyroid Thyroid (1E*3) (21E3) - 3E-9 2E-5 2E-4 53 lodine-131 , 0. all compounds 31.1 SE1* 2E-E Thyroid Thyroid (91.1) (2E*2) - 2E-10 11-6 1 S 53 ldir132 ' D. a)) copoads 4E.3 &E13 4E-6 Thyroid Thyroid r(11.4). (2E+4) 3E-E IE-4 IE-3 53 lodin.-132 0. all campounds 4E*3 .E*3 3E-6 Thyrold Thyroid. (91.3) (SE4) - SE-4 53 Iodine-133 11.2

0. all compounds 31.2 3E-7 T@hyroid Thyreid (SE12) 1E-9 7E-C 7E-5 2 OEM4 53 lod1neL34 . D. all comoumds 2714 SE+4 2E-S 6E-8 6-a (31.4) 4E-4 4E-3 53 Iodine-US 0. all ca ounds 2E13 7E-7 Thyroid Thyroid 11E.11 (41.3) . CE-9 3E-S 3E-4 54 lenoneir2 2 _ IE-S 4E-E 2

54 tenon-3l .-. Suersioni - 2E-6 2E-8

                                               *Ssuuersionx 54         XSeon122                                                                  -              7E-5'     3E-7 54        Xvnon-523,                                                                -              6E-6       3E-8 54         R0non-125                 Siarsion                                       -              2E-S       7E-8 54         Xernotru2                 Subtersloo                                      -              lE-S       6E-8
        *54          Xenon-32.2                                                                                          9E-7 54         lvlnon-133                Sumaersioai                                                    2£-4       2E-6
                                                                                               -              *E-4 1

54 lnon-13 Svbmersien 6E-7

                                                                                               -              11-4 54          NCIr fl33                Sutuallmnp S.Sliersis.n                                                   1E-4      SE-7                                     f 54        S non-135.

2 SE-6 4E-S . i i 54 ladnon-IS SE-5 2E-8 I 2 54 lrenon-135 - 4E-S tE-7 1 55 Cesiim-125S a. aellen 5E44 1E+S 6E-S 2E-7

0. -l.... ... St. wall (91.4)

• 1E-3 1E-2 55 Cesiu m-327 e, allc _.ond .4 . 9E-4 4E-S 1E-7 9E-4 91-3 B-23 ODCM Rev. 25

I L jApp. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION: App. B. 4 I .I Table I Occupational Values able 2 Effluent Table 3 Release$ to Concentrations Sewers Col. I Col. 2 Col. 3 Cl. 1 Cal. 2 Oral. monthly Ingestion Inhalation Average Atomic Radlenuclide Class ALI AKl OrC Air Water Concentration NO. (PC,) (PCO) (c - pt ) p§Z)(t/l 55 Cesium-129 O al compounds 2E14 314 1E-S SE-I 3E-4 3E13 2 55 *Ceslum130 D. ll compaound 6E-4 2E-S PE-5 iE-7 - St. wall

                                                                                                             -               -              1E-3          1E-2 55        Csi~tm-it             0. all compoundis             21E4           3E4             lE-5            4E-6           31-4          3E-3 55        Ceslun132              0. all Copounds              3E13           4E*3            ZE-6.           6E-9          4E-5           41-4 55         Cesiuaw34m            0. all crpound5              1S             lES              SE-S           2E-7            -             -

St. wall (11.5) _ - - 2E-3 2E-2 55 Ceim134 P. all c1mounds 7E41 1E2 4E-t 2E-10 9E-7 91-6 2 55 Cesfum-135 D.eall onpo 1E-5 2E1S oE-S 3E-7 1E-3 1E-2 55 Cesl-35 0. *all cO ounds 7E12 1E13 SE-7 2E-9 5 1E-4 55 Cesfu-136 0. all compunds 4*2 7E12 3E-7 9E-1D 6E-6 6E-S 55 Cesi a-337 0Dall cowounds 11E2 2E12. 6E-8 2E-10 11-C lE-S 55 Cosluor.1382 0 all. coponds .21E4 61.4 21-5 81-St - ' - r St.wall (3 1.4). - - 4E-4 4E-3 2 56

• rlw126 D. all copounds 6143 2E*4 6E-6 2E-S SE-S tE-4 56 Sarfuu-128 0 all coounds SE2 2E-3 7E-7 2E-9 7E-6 71E-S 56 Uiju_13102 0D all cmounds 4E1S 1E14- *E-4 2E-6 - -  :

St. wall (5.ES) - - 7E-3 7E-2

            ;56          Sarifu-131           D. all compounds              3E13          8E43            3E-6            11E-           4E-S           4E-4 S6           eriuw-133a         0. all cmounds                 2E13          9E13           4E-6          .11--             -             -

LLI wall (3E13) - - - 4E-5 4E-4 56 Urium-133 0. all compnds 2E13. 7JE2 3E-i 9E-10 2E-S 2E-4 S6 Sariia-135m 0. all compounds 3E13 1E-4 SE-6 2E-S 4E-5 4E-4 2 56 *arium132 0, all Ceomods 1E-4 31E4 1E-5 4E-8 2E-4 2E-3 SC Barfu-140 0. all compounds SE2 1E43 6E-7 2E-9 - - LI wall (6.E2) - aE-s aE-5 2 56 Oirlum-241 D. all compounds 2E14 7E-4 3E-S 1E-7 3E-4 3E-3 2 56 Barium-142 0 all copounds SE14 1E14 61-S 2E-7 7E-4 7E-3 2 57 Lanthanue-131 0. all ceoe_ except those given for V SE44 1E+S SE-S . 2E-7 6E-4 CE-3 W- oidors and ydrovides - 21.5 7E-5 2E-7 - - B-24 ODCM Rev. 25

App. B App. B' PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION .: TabIt I Table 2 - Table 3

                           .           .                                   .                 i.:cutatonal Values                        Effluent             Releases to Concentratlons                 Sewers
                                                                                - teci. I            Col. 2.          to). 3       Col. 1        Col. 2
                                                                  *, .              Oral                                                                        Monthly
                                                                       - -          Inoestlor             Inhalation                                            Awerage Atomc Radiomuclide                 class                                                ALi            OAC           Air          Vater       toncentration Noc.                                                                                  O M~i       O W(NI/m) pCI/mi).          (PcI 1.)       (PcI tel) 57         LAnthanm_132            0, **r 131                          3E+3             1E-4           4E-6           1E-8          4E-S-         4E-4 11.4            SE-6          2E-8                -       -

S7 Lanthainm-135 e. " 3231.1. 2E1S  ; 1E-7 SE-4 SE-3 W: .5. 1312.8 9E+4 4E-5 1E-7 - - 57 Lanthanm-f137 .. ea, 132L 5E41 3E-8 - 2E-4 2E-3 Liver (7M1) - E-10 - - Wi.set 3a 3E12 11-7 - - - St V Liver -4 -1 -- W: a" 131L, (3E*2) - -E-l - 57 LantUhanm-M1 D' iae 131 4E10 IE-9 -SE-12 1E-5 IE-4 St s La . 4E+4 11E1 6E-9 2E-U -U Lnu' 131 (41*4) 57- .entUa w-140 0. set 3 La 1E.3 . 6E-7 2E-9 9E-6 93-5 21.3 51-7 2E-9 - 57

                    ~

s7 L

                                ~ ~

hanlr141 eV~ V. see 131ja L11a

                                                            .atnz                   11.4 4E+3               91.3           4E-S         l-S..         SES            SE-4 4.3            1E14          SE-6          2E-S           -              -

II II 57 Lanthanu-4i 0, see 13. I1E4 9*-6 3E-8 21-4 IE-3 (6E+2)

0. .see 131 La W

2 3E+4 IE-S SE-S - - 57 LsUiawMV-143 II 91.2 1E+5 4E-S 1E-7 - * - I1 'St. wall (4E-4) - - - 5E-4 51-3

                             -*                     s at 1312.                                    9E.4            4E-S           1E-7:         -             -

U Ctrim-134 VWel cIapaosds except

                                         -    thse given for Y                                     71E2            3E-7          lE-i          -             -
                                                                                                                                              -E-6            tE-5

r. oxides, htdruxudes, and tleorldes- 71E2 3E-7 9f-10 - -

Uo Cerium-135 W seo 3Ce *(CE-2) .4E13 2E-6 51-9 2E-5 2E-4 4E13 IE-6 SE-9 - - 38 t rin-137 e V. see tC 4E13 2E1- 6E-9 - - Ul wall L.11mall _ . - 3E-S 3E-4 V. see 1341. 4E 3 2E-6 1-E9 - - 134 5J teri-137 W. a" MC 11.5 E,-5 2E-7 7E-4 7E-3 2.143 E+S SE- 2E-7 - - 58 Cat eim139 - . ** se 1 33Ce (21.33 8E12 3E-7 2E-J .7E-S 7E-4 4 lllee C; 5 7E-2 - 3E7 S-10 - - 58 Car -1J3 V. see 3Ce 71.2

• 3E-7 lE-l -

                                                                                                                          .,                 3E-s           3E-4 Y. Ses 134C.                         Ul. _all       6E52            2E-7            8E-I0         -             -
          -SI       Cariumn-143              W. as. 15Ce                        (11.3)           21')3           6f-7           3E-3           -             -
                                                                                                  -              -

• 21-5 21-4 Y. 5se2.4c ZE23 7E-7 2E-9 -

B3-25 ODCM Rev. 25

II i II App.!.13' App. B .PART 20 STANDARDS .FOR PROTECTION AGAINST RADIATION.

                                                                                                                                                                .. 4 I Table 2            Tdle 3 Occupational Values                  Effluent                     to teleoses Concentrations           Sewer, Col. 1       Col. 2         Col. 3     Col.         Col. 2 Oral                                                             Motnthly Ingeston          tnhalatfon                                   Average Atminc   Radfonsucltde .       Class                          ALI                                     Air.         watar   Concentration NO.                                                         (COl          N(Oa)      (eCI/al)   I(pyi/al)     (pyi/mi)    (pCi/el) so       Cerruem-U            W. SeaU4C                       ZE42          3E1.           XE-S       4E-11        -          -

LI wall O3E42) - _ 3E-6 31-S Y.* Sa 134Ce -1 1 6E-9 2E-11 - - 19 PraseodymiLam-362 W. all compounds eacept tb@Ss givenfoY SE*4 St. wa11 ZE IE-4 j3E - (7E-4) IE-3 IE-Z Y. oxides, hydroxides, carbide.. and fluorides - 2E*5 9E-S 3E-7 2 13 eody

                             ;9ParUe17 W.Sea          6Pr                41'4          2E-5           6E-S       2E-7         SE14        SE-3 19        Prassodyasa-1daIl Sea            P                 -              11E5           6E15       2E-7 1E+4           5E44          2E-S        BE-8         IE-4        1E-3 Y.s.          Pr              -              4E14           ZE-S       6E-8 59       PreseodymiuvM13       Wu , t     36'                4E14          IE1S            SE-S                    6E-4       6E-3 IE*S           SE-S        2E-7 2E-7 59       FraseadymiuvrMA242         seeam~                   SE-4          2E.5           7E-5                      1E-3      1E-2
                                                                        -             1E45           6E5         2E-7 3E-5 59       Prastodymuhu-142 W. act            Pr               IE3           2E13            9E-7       31-9         l1E-S      21-4 Y. see 13pr                                  21E3           eE-7

• 3E-9 I.

          .59       Pressodywitu-143     W. too Iipr                   9E42           tE12           3E-7        lE-9 LLUwall 136 (1143)                                                  2E-S        2E-4 Y.ise,     pir             -              71.2           3E-t        9E-10 2

SO Fraasdy'luera-14 W. see iPr 3144 1+15 5E-S ZE-7 St. well (4E4) - - 6E-4 T. see L15S1. - 11.5 SE-S 2E-7 59 Frauuodywha-w14 W. see Pr 3E43 9E*I 4E-6 l1-9 4E-S 2E-3

                                                                       -              8E.3           3E-6        1E-o 2

59 Preseaodyalu.-141 W., iS"L r SE-4 2f*S SE-S W 3E-7 St. wall (SE-4) - IE-3 'E-4 Y. B"ea sp - 2E SE-S 3E-7 1E-2 60 Neodysim-LUZ1& W. all ceapourida except thoe ivn for Y 11.4 6E-4 2t-5 SE-S 2E-4 Y.. Oxidjes. hydrexides, 2EI carbideig. NWdfluorides - 5E-4 2E-5

• 6E-O 60 11todyium-Iw- W. S"a Nd 21.3 61*3 31-6 iE-9 31-S 3E-4

                                                                     -               SE+3            2E       7E-9 60        mwe*Uiums13f          W.S ...                     SE.3            2144            7E-4       21-S        7E-S       n-4
                                                                      -                 144         6E-6        ZE-         -          -

B-26 ODCM Rev. 25

App. B App. B-PART 20. STANDARDS FOR PROTECTION - AGAINST RADIATION

                                                                                                     -table I                                  labie.2                    Table 3
                                                                                      -    Occopational Values                                Effluent                 Releas to Concentrations                   Stwers
                                                          *,     ,              Col . I             Col. 2             Col. 3           Col. 1            Col. 2 Oral                                                                                     honthly Ingestion                 Inhalation                                                     Average Atomic    Radionuclide        Class                   .          ALI                 -IT                 W               Air           ,WYtur          Concentration No.        ...                              .. '            '(PCi),                (yClt)       (fiial)          (pC/al)                WCilel)      (lCs/al).

2 35 60 Neodyr um-239 V* 2e N1d 1 36 9E.4 3E.5 1E-i SE-7 1E-3 -1E-2 Y. rn Md - 3E-S IE-4 4E-7 - 60 Peodymilu24 Y ae 136 - 7E15 3E-4 1E-6 2E-3 2E-2 1361d 2E#5 6E15 3E-4 9E-7 - W. Ste 236d 7-t 60 60 Peodyniun-247 uodyiWi-149 2 V. aee 2lijgdi 3 W. aer 2 6gd O Mt

                                                                               -d+

9E12 8E+2 3£-4 41-7 4E-7

lE-S lE-9 l1E-9 4E-t' 2t-5 1E-4 2E-4 IE-3 2E-4 lE-5 3E-S - -

60 INeodye r-2512S - W 236N  :: - St 7E44 w1 2E.5 8E-S 3E-7 9E-4 9E-3 2E.5 8E-5 .3E-7 - - 2 61 Prometfivii-141 . W, all comps,ds except the. given la? Y -S[44 2E+.5 5-5 3E-7 - -

                                                                         .St.       wall (61.4)                                                  -                 JE-4             8E-3.

Y. oxides, Itydrouides. carbidus. and fluorides 2E+S 7E-S 2E-7 - - I I-en 61 Promethigm-143 v: se. 341FI 6E12 I 2E-7 .E-10 7E1S 7E-4 71E2 3E-7. lE-9 - - co in 61 7ra metlei-34 YY see 1E43 11E2 SE-9 2E-10 2E-S 2E-4 E* pr> j y 141, 1E42 SE-$ 2 0-1 - - 61, Pr..etim-145 V. see p1.2 , 7E-S - 1E-4 1E-3 one surf

                                                                                                 !E#2).           -                 3E-10 Y. set 141p                                            2E42             JE-8              3E          -               -

I

            "         Proemethm-146       y, a.4 14 r,                                 2E* 3              IiE#2             2E-8              7E-31            2E-S           2E-4
                                                                                       -4                 4E1     2E-S             UE-11, 61       Pramethism-147      W. Sete4                           4E.3               I1        au2 SE-6                 -                -               -

tLI watl ISon surf (SE13) (2142) - 3E-10 7E-S 7E-4 Y; see _ 1 1.2 6E-S -2E-10 - - 61 Promtblui-24&a -' sue1414 71.2 3 1E2 1E-7 41-20 3E-5 11-4 1*V:set 141., -3 142 11-7 5E-10 - - 61 Prometh~um-143 W, see 141tw 4E12 5 1E2 21-7 tE-10 - ,- LULwall SE#2) Y. s 141., 1E2 2E-7 7E-,0 7.-S 61 Pr-othium-149 W. g 241 1E.3 , 2 1E3 JE-7 . 3E-9. - - L1I wal (1E#3) - - - 2E-5 2E-4

• Y. got 21 E#3 6E-7 2E-9 - -

61 Proinethfun-250 se 2. SE13 21 E.4 , E-6 31-6 7E-S .71-4

                                                                                     -21         E.4             7E-6             2E-S            -               -

1 41 61 P amethum-151 j.ac p 21.3 41E#3 E1-6 5E-9 2E-S 2E-4 31E43 1E-6 4E-9 - _ B-27 ODCM Rev. 25

App PART20 STANDARDS FOR PROtECTION AGAINST RADIATION . A I Tablel T able 2 Table 3 Occupetlonul Values Efluent Releams to Concentrati ons Sewers Col. 1 Col. 2 Col. 3 Col. 1 Col. 2 Oral monthly Ingestion Inhalation Average Atoie Radlonuc1fdo Class ALI Al DA6 Air Water concentration 62 SaMarlua141a 2 W. all cocmpuds 3E64 1E6S 4E-S 1E-7 .4E-4 4E-3. 62 Samarfim-141 2 W. all comounda SE-4 2E6S SE-S 2E-7 - - St. wall (6E+4) - - - -4 E-62 Smarfun-1422 W. all, comounda 8E.3 3E+4 lE-5 4E-I 1E-4 1E-3 i2 Smariu-145 W, all compounds 66E3 56E2 2E-7 7E-10 BE-S . E-4 62 -Samarfun-146 W. all cowponds 16.1 4E-2 IE-11 - - son surt Done aurt (3EI6) (6E-2) - 9E-14 36-7 3E-6 62 Simmarla-147 W. all c~omond5 ZE+21 4E-2 2E-11 1 - - Bone sur t sr - (3E6,) (7E-2) - E-13 4E-7 4E-C 62 Smawita-151 W. all coounds .1E-4 16.2 . 4t- - - LI wall Bonesurf (16*4) (26.2) - 2E-10 2E-4 2E-3 62 Samariim-153 V, all c~oqponds 2E+3 36.3 lE-6 4E-9. - L*I wall (2E#3) - - - 3E-5 3E-4 Smariu-155 2 .I

       *62                          W$all compounds           6E+4          2E*S         .95-S           3E-7          -            -

St. wall (86-4) - - - 1E-3 1E-2 62 Samarlim-1SI W. all compoands SE#3 9643 4E-6 1E68 7E-S 7E-4

6) Euro 45 uiiaw W. all comoanda 2E63 2E63 8E-7 3E-9 2E-S 2E-4 63 Europiumr-146 W. All comounda 1E63 16.3 56-7 2E-9 1E-S 1E-4

       *63         Europfur-147     W. all comeunds          3E63           2.E3            76-7        2E-9           46-S         4E-4 63         Europiw:-148     W. all compoirml         16E3           41.2            1E-7        SE-10         1E-5          1E-4 63          Europiaun-149   W. all comounds           2E64          3E63            lE-6       4E-9           2E-4          2E-3 63        Evrogiterl50      V;all      ow d          3E63            56.3          4-6          16E-S         4E-5         4E-4 (12. 62 h) 63        Europlw-ISO       V. all compounds          8E62         26*1              8E-9      36-11         1E         1E-4

• 34 .2 Y) 63 turopfiw-15ft W, all comounids 3E63 CE3 3E6- 9E-9 4*-S 4E-4

      *63.   '.Europiu-152          V, all compounds        -E#Z           2t+1            16-5        3E-U          1E-S          1E-4 63         Europim-154       W.,all c~ompouds        5E6Z           2E41           BE-9         3E-U          7E-6          7E-S 63         Ewroptiul-15      v. alIacompuds,        4E63            9E#1           4W-S        -              SE-S         SE-4 Boeasurf.
                                                            -             (31E2)          -           2E-10         -             -

63 Eurogium-156 W;all comounda 6E62 56Z 2E-7 66-I0 BE-6 eE-5 B-28 ODCM Rev. 25

2 App. B App. B PART 20. STANDARDS F6R PROTECTION AGAINST RADIATION Table 1 Table 2 Table 3

                                                                             *Occtwetional                       Vaile               Effluent          bett Concentrations            Sewrs Col. 1          Coll 2          Col. 3    Col. 1        Cl .2 Oral                                                               Nontlaly Ingestion            Inhalation                                    Average AltookI Radionuclide            Class                                  ALI          -XLT                M       :Air           Water    ConcentratIon "a.                                                                   (YU)            (jaC,) '(MCI/ai) (iSctI/p)(MCi/al). '(PCI/ml) 63         Europiuw-157      .W,& al      'compounds                   7E43            5E*3            Z215      7E19          U15        J3E4 2

63 Eurojptum-1sg W. all compouads 2E.4 6(44 2E-5 L8E-8 3E-4 3E-3 64 Gadol iniuw-1452 D. all Caupowads except those given for.W 5E. 2E+5 6E-5 26-7 (SE+4) - 6E-4 6E-3 W. oxides. hydroxides. and fluorides, 2E+5 7(-5 2E-7. . -- 64 Gadol inisw-146 'ZE-1O 2(Z-5 2E-4

                                                                                                                              .4E-10           -. -
         * .64           Ca&doIfafu-147.        ase 14S~d                            -E3             4E41            16'       S6-9          3E-5       3E-4 W:5. set114164                                         4E*3            3E-Il 64         CadolInium-i4i 14                        (2E.1)'           (2-2)           -          26-24          if-7      3E-6 W. se         5u                       -                3E-2      . 16-11 Bone surf (56-2)                     86-34       -.

45 64 *C'adolinaum-249 2 w. XI* 2E+3 9E-7 3E-9g 464S 4E-4 Wf G 2E43 - 36-9 -

            .64         .cadoliniam-I15      **se 15 Gd                             *(3              4E.Z           2E-7                                9E-4 I                                                                                                     so"e surf

y. ,. -. - (6E*Z) .SE-IO

W. ts, 459dJE*3 51-7 26-9

           -64          Cadolnf un-IS2 lBan surf        -lone surf, 46-2)           2-1       3E-14         4E-7        4E-6' W. set     S~dBoa-                              s          E-w1                              4-S 36E-23 64     -      ado~tln R7i3u-1                         24*w(86-2)                                 .                                         6E-4 W. see                                                   6E+2           2E-7       8E-10.

• 64U :o 24I 3E.3 86.3 M(4 46-5 4E-4 SE-I0

          .65                                              W.al Cmouds9                           .336.4           IE-5                    '(-4        1E-3 65                             d. all C~OW"d                          SE-3    ~.        762        . 3-7                     7i-5        7E-4 65          Terbtum-145         f. all Compouds                        1(43       .-      E244          S-                      IE-5       7E-4
                     -Terbium-150          i. all comounda                       4E43              9E+3           46-6       £6-S          SE-s       5E-4 65                              I. all eomounds                       16E1              76.3          73-         iE-S        .76-5        7E-4 65         lerbsum-154                 T.alC~am                       2643              4E.3           2E(6       41-S                     2E4
                                                                                                                                          .2i-S-55                              1. all comaund5                       6(43           K 863             3E-6                                *E-4 Terbiaun-I55 U5.         TebiM-O            V)'. .1c~oamouda                       26.4              3E44           1645.                   26-4        2E-3 B-29               I-     I                                                 ODCM Rev. 25

I I App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Table 1 Table 2 Table 3 Occupational.Values Ifflumnt Iuleasia to Contentratfons Sewrs Col. 1 Col. 2 Col. 3 Col. I Col. 2 Oral Mlonthly Ingestlon InValstlon Average Atnefc Iadfanuclfde Class AL !. - Atir Water Cociettratfan (CPC) 00t) (MCI/0l) 00csi ) Wpi/11l) (pAt/al).' Ss ler-blum-1564 W. all caompouds 7E13 aE13 3E1- 1E-8 1E-4 IE-3 (24.4 P.) 65 Terblui-l156 V. all campoufds 1E.3 E1.3 6E-7 2E-9 IE-5 1E-4

           .65        Terblue.1S7      W. all co@mpouds           SE+4            31.2             1E-7        -              -

LI.wall go" surf (51E4) (6E12) - bE-lo 7E-4 7E-3 65' Tewblus-ISS W. all c~owpods. 11E3 2E11 8E-9 3E-11 2E-S 2E-4 65 7ertfI.-260 v. all coap.unds Bi.2 2E12 9E-3 3E-10 1E-5 1E-4 65 Tertbu-216 W. all compunds 2E.3 21E3 7E-7 2E-9 - LLI wall (2E*3) - - ^ 3E-S 3E-4 64 Dysprostia-15S W. all Comounds 9E13 3E-4 lE-S. 4E-8 1E-4 1E-3 64 Dysprosium-157 W. all cow"I'mdi 2E14 6E.4 3E-S 9E-8 3E-4 3E-3 i 96 DYSPr'uwae-19 W.- all campoundsd 11E4 2E13 1E-6 3E-9 2E-4 2E-3 66 Oysprostuirl6i. W. all co~mpuds 1E4 51E4 2E-5 .6E-S 2E-4 2E-3 II 66 Dysproatrm-165 V, all cotmpuds 6E42 71.2 3E-7 1E-9 LLI.wall M1'42) - - - 1E-5 1E-4 2 67 14lmara.-155 V, all cowosnda 4E14 2E15 6E-S 2E-7 SE-4 6E-3

          £7         "0xlahw137'2     v. all compouds           3E+5             114            SE-4         iE-5            4E-3        4E-2 2

67 Hlitel rlw19 W, all c*aound 2E1S' 11E6 4E-4 ,E-4 3E-3 3E-2 is 67 Holef..r161 V, all c~o~mpds lE S1 415 2E-4 6E-7 1E-3 1E-2 67 Holohma-262oz V. all cao~mpon

• SE4 314S 1E-4 4E-7 7E-4 7E-3 2

67 Nalsllu.-262 v.all C."wund, 5SE5 2E4 11-1 3E-6 - St. wall - - - 1E-2 lE-67 H410olrt~u.-6 v. All Co~mpuds lEd5 3E45 11-4 4E-7 1E-3 1E-2 2 67 Holsfull164 V. all c 2E+5 5*5 3E-4 9E-7 - St. all (ZE*S), , 3E-3 3E-2 67 ltolialum-166a W. all comounds 6E*2 7E10 3-S 91-12 9-f 9E-S

         £7        Holehur-16S        V. all c~osmpds          9E12             2E13            7--7        2E-9            -           -

LLI! wall (9E*2) - - - lE-S 11-4 67 Holsius-167 V. all compouds '2E-4 6E.4 2E-S SE-8 2E-4 21-3 68 Irtaium-161 W. a1l comounds 2E-4 6E*4 3E-S 9E-1 2E-4 2E-3 68 Erbiua-16.S W. all Co~mpuds. 6E14 2E.S CE-S 3E-7 9E-4 9E-3 B-30 ODCM Rev. 25

AppO. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION lable I Occupational Value4 table 2 Effluent Concentrations lable 3 Releases to Sewers

                                                                                                                                                                    *1%

Co-tl1 Col. 2 Col. 3 Col. I Col. 2 Oral Il"onthly' 1niQstSfn Inhalation Average Atomic Radionuclide Class AL - D C Air Water toncentnratlc 60 Irblu.i69 w. all comounds - 3E13 3E13 11-6 4E-9 - - (41.3) 5E-4 68 Erbfummlfl W. all copounds 4*E3 1E.4 4E-6 lE-ts SE-S SE-4 68 rbimun-172 W. all cmwourids 11.3 1 6E-7 2E-9 2E.3

                                                                    .    .-           W wall .

(11. UE3) .t _ ZE-S 2E-4 69 Thwli.-U22 31E*5 S .1E-4 4E-7 l a11 C1oundS; 7E14-

,St. wall - 1E-3 (71.4)

                                                                                                               .. _C.

69 Thulimm166 v, all compounds 4E'3 U144 6E-6 2E-t *E-5 6E-4 69 Ttiali~mw167 W. all compcumds Z1.J 3E-9 _ 2E13 8-7 U1 wall (2E13) - i3E-5 3E-4 69 Tutiuali1m70 v, all compounds W2 *2E42 9E-8 31-10 - LIU wat 11-4 (1E*3) I E-S 2E-7 in 69 Thulfum-171 V. all c~omonids . I 11.4 LLI wall 3E#Z Renasurf (U24) (61E2) IIE-10 2E-4 2E-3 U- 69 Thuli~m172 W. -all compouds 11E3 5E-7 LUI wall 2- I-5 (812) 2E-t E IE-4 69 . Thullsm-173 W. all comcwVs- 4E13 1E*4 5E-6 21- . £ *E-4 2 4E-7 5 69 Thuliwo-175 V, ll comouida 7E14 31.5 1E-4 St. .wall (9E+4) 1E-3 1E-2 2 70 yttlteiwlu162 W. a1lCaq§Paunds eacat -W c.I-sw- 41 71.4 .;-, I-7 1E-3 1E-2

• 7, ewildina iwy r aid es.-

audO fl1uowlides 3E15 3E-4 4E-7 70 Ytterbimin-16 . aa 1143 2E+3 E8-7 3E1-9 2E-S 2E-4 2E+3

                                                                                                 *21.3            BE-7        3E-9 2

70 Ytterbiuar-167 'W sew Y 3E*S 3E-4 11E- 4E-3. 4-2 3E-4 1E-S 3.3. 11.2 70 yttarbitem-169 . V sat f 2E-3 7E.2 4t-7 IE-9. 2E-S 2E-4 3E-7 lE-9 IW: see -4E*3 1E-6 SE-9 21-5 LLI wall 70 2

• ae 2ta62n~

T (31E3) 4E-S 4E-4 I. 3E#3 1E-6 SE-9 V. see 1S2 2E+ 4 5E644 2E-S 7E-t 2E-4 2E-3 70 5E#4 2E-5 6E-I 2E-4 ytterbwl27S' W. aeaU 1..4 11.4 2E-S . 61-8 ZE-3 Y.. ea 4E*4 2E-5 51-t B-31 ODCM Rev. 25

ii PP PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App., B Table I Table 2 Table 3 Occupational Values Effluent Relesses to Cencentrattons Sne" Col. I Col. 2 Col. J. Cot. I Col. 2 Oral monthly Atomic Radionucide Class Ingestion Inhalation Mr Vatar Average Aoi laiwcdo CasALI AIUw - i ae Concentratiofl No. (PCO (loci) - Ocf/ol) (PU/01ll. 61CI/MI) (pCiful) 71 Lutetiu. 169 W. all CompouWda except those given for Y 3E13 4E*3 2E-6 6E-9 3E-5 3E-4 Y. oxides hydroxides, and fluorides - 4E13 2E-6 6E-9 71 tutctu_170 W. set Lu IE.3 2E13 9E-? 3E-9 2E-5 2E-4 Y. See 16 9 Lu 21.3 SE-?. 31-9 71 atiat-17u W. see 169e 2E13 2E13 E1- 31E-9 31-5 3E-4 T. let 6 Lu - 2143 31-7 31-9 71 Lutattu-172 v* se 1 69Lu 1E3 1E+3 SE-7 '2E-9 1E-5 1E-4 V. 'Se Lu - 113 51-7 '21-9 - 1 9 71 tutattu-173- W. s" 6 Lu  ; SE3 3E12 1E-7 - 71-5 iE-4 eonr surt

                                                                             -     -(51.2)               -             5-20,        -          -

Y, see 1Lu - 3E+2 1E-7 4E-10 - 169 71 tutetiu174 W. See LU ' 23 22 1E-7 - - - LLI wall Do Su" E-169 llE 3) 132E*2 l E2) 3E-6 SE-10 3E-10

                                                                                                                                     *E-5 4E-4 Y. see     6 La I.-

IL 71 n tetfe-74 Eutatfun-176m W. I Y. see YW 1 69 169 Lu Lo 6 Lu 169 5E+3 6E43 1E12 (2E12) 21E2 3E*4 5E-i 6E-IE S 3E-10 2E-100 3E-7E-5 _1-7E-4 1E-3 71 uttt1i66 V see 9u 71.2 .E+4 2E- - 13-5 1E Senoses"? Y. ro"Lu - 8E1.0 3E-9 lE-11 - 71 Lutetls-w.17 W. "e L 7E12 E+.2 2E-9 - 1E-5 1E-4 anea ur? 16- (12) - 2E-10 - -

                                   *. see.       Lo                 * .                                3E-9           1E-10        -1 LUW                "ll 71       Lutetiuam177        W         16uoLu2                Ul*2 .                             SE-9         31-i              -         E-1931,3)                                                           41-S
                                                                                                                                   *E        *E-4 Y. see                                      .    . 2E3             9-7          3E-9           -

71 Lutailum-271m y. Se

                                       .       9      ,            SE44                2.E5            6t-5         3E-7           -

• SL wall r, J(64 4) 2E- -7 i-39-3 2 169 St. wall -- - 6- 1 71 Lutett.u-178 W se ,'L 41E4 (6E*4) -1ES 5E-S 2E-7 - E-4 -

6E-3 r, see 169 (.4). 1E5 SE-5 2E-7 - 71 Lutotlum-173 W, see 169 Lu 6E13 21E4 oE-6 32E- 9E-5 SE-4 Y: sl Lu - 2E+4 6E-6 31-S - - 0 B-32 ODCM Rev. 25

App. B PART 20 STANDARDS.FOR PROTECTION AGAINST RADIATION ApTB Table 1 Table 2 Table 2 Occujpationasl Values, -Effluesst ieIesses to coflcsntrstlons S4;4Vrs Ora. Col. 2 Col.3* Cot. I..Col. 2 01st . Mnthly

                                                                       -Ingpstiofi                   Inhalatlon                                           Average Atomic Radionuclide            Class                               ALI!             ALIt---DAr                     Air          Water       Concentration No.'                                                            -(PCI             (PCI)         (PC,/e,)       (PCI/81)       (pCI/mI)     (PCI/Ml) 72       Rainlimm-1TO    *V.al c~ogjonds exc-p those given for W                   21.3           6E*3                 2E-6        81E-9        4E-S           [5 W. oxides. tydrozides.

carbides. and nlitrates SE3 2E-E 3E-2 72 Matoimw-172 ..D set 70

                                                                ~               j,               E40 Bone sort 4E-3
                                                                                                                              .    -11 2E-5         2E-4 (21I)               -

4E11 2U-1 Done surt (6E1) - ,3E-U 2_ 8 170 72 H&fI'Mu-173 D: :.. Hf SE*3 1E.4 5E-6 71-s 7E-4 1E4 SE-C 79 72 Maffliumi-17S 0. Sea ; m' 31.3 9E*2 4E-7 2E--1 4E-5 4E-4 gone surs. (1.3) - .21-9 1E13 5.E-7 2 -1E-3 2t-S i2 Y~aInIsm-177. 170 2Ee+4p 6E+4 2E-S SE-11 3E-4 9E+4 4E-S 3E-22 .a 72 Hafnilm-178m. 0. set 2 Ht 70 21.2 l 'd5S-10 31-S I' twoe curt (2EO) - SE+O ZE-S 1E-70 .31-6

                                                                                             .Bome surt (1E40)              -             SE-i U-0 1 70 72        Hefliawm-171.      D. S"                  k              *113            3E.2                1E-7         3E-12         3E-S        1E-3 one surf 1                                   (6EE2)               -

W. set 7%f. 6E.2 3E-7 1 .21-4 72 Nhatniaw.-1B 0 eg m; 0N 71.3 ~2t 3P.4 4 91S 1E-5 SE-3 41-i 7 IHsfnti'r8 3, see 14 lE*3. 21.2 7E-2 2E-S 2E-4 Bos surf (4E+2) - lE-10 4E+2 2E-7 , 72 rstnifeM-1*2 0*ee~y 4144' S1E4 4E5 _.-7 5E-4 SE3-3 1E5 '6SE-S 2E-22 1 0 72 Msam12 0. S"g ?ot 21.2 #-1 3E-10 ieee suit so" surf (2E4) .- 3E-10 SE-S W.s, "S t . . *.20 31-9 Sone suit H-2 F2 H"affimm313 1O'S"27 21.44 51E4 2E-5 SI-i 3E-4 *3-4 1 6E44 2E-5 11-i P2 Nstnlam-2jN .se 1 T t 2t-3 81.3 3E-S ,E-1 31-S 3EM-4 V.w a t . . ; 9SE3 3X-Vv B-33 ODCM Rev. 25

App B App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION i pp. 1 B. le Table 1 JATWb2 Tablo 3 Occpatlanal Values Effluent Releases to- ' z i Concentrations Sewers Col. 1 Col, 2. Col. 3 Cel. I Col. 2 Oral ."nthly

                                                                        - Ingestion                Inhalation                                            Awrage Atomic Utdlaonclide              Class                            A ALI,                  I           W                   Alr            water   Conentration 73        TpC                      al(PCI/n)                                      (Pana        (PCl/a)           00/0.)                       WCcd) 73         Tantalu;_1722        W. all com~ouds xxttpt

_. those gten for Y 4E-4 1E5 SE-S -. 2E-7 SE-4 SE-I Y,elemental Ta oxi dei, hydrexldes. halides,

                                           =arb1ds nitrates.

and nitrides 4E-S ;11~-7 - 172 73 Tant~slam-173 W se Ta. 71.! 2E-4 BE-6 3E-S 91 5.5 S9-4

                                                                         -                  29-4            71-6               ZE-8 2

73 Tan'talu.-174 y: in~ W' 1721` 4E-5 4E-S i1-7 1E-7 4E-4 4E-3. 73 Tantalui-175 6E43 .214 4E-6 2E-8 BE 8E-4 _ 1E+4 6E-6 i 2E-. - To 73 Taaat~al-- 176 V seele172T 4E+3 1E-4 SE-S 2E-S SE .SE-4  :.. sev17e21` SE-C 2E-S -- 73 Tatatulm-177 V 72se IE.4 2E.4 5E-li i3E-S 2E-4 2E-3

w. s. 372r - 2144 7E-6 i 2E-S -

73 - Tantauhm-78 2E*4 9E.4 4E-S i .1E-7 2f -4 2E-3

                                               ...                                         71.4            3E-S                1E-7 Y* sew        To 73         l'antulium-179        Wj       1721'                  21E4               SE3             2f-C                8iE-           3E        .3E-3

se .l2Ta2 91.2 4E-7 tE-9 -

73 . Tantalum-law 172j 2E.4 7E-4 3E-5 9E;S 3E-4 3E-3 2 - - 26E.4

                                                                                       . .                 2E-5                BE-S           -

7.se17 Ta 73 Taantalum-180 1143 41.2 2E-7 61-10 2E .S5 2E-4 W. sitee1` - 21.1 tE-S 3f-ul 2

*73              lantalUM-829ft                                        2E+S                               41-S 5E-5            21-4                       31- 3E-7   1 St. wll                              31-S              _

1 72 (2E45) 21-6 '3 3E-2 Y. see Te 4t-S 2E-4 6E-7 - 73 laststaun-M.,- see 17 .~T SE12

                                            ~:se   2                                     3E.2             1E-7               SE-10           1E--5      1E-4.

Ta 2E-10 - Tantalum-283 v. see InTs 9142 11.3 LLI wall {311.3) .-. 3LE-4 IE-9 _2f- - 5 2E-4

• 73 lurtl~m-54 V see 172 Ta - 11+3 3E-S

v. see 7t 5-JE-

*73            Uaetalmir-252         V. se 72 Ta 2_4                 5E3                                3E-9           3E- -S       3E-4 71-9           -

1 31.4 7f.4 11-5 IE-7 4E-' 4 4E-3 73 TantalwrlhS W.se 7Ta 2E-C

           .                                                                                             31-5              9E-a
                                                                    -E4           -     2f*

51-SE4 SC. 1. 2E -S 3E-7

                                                                 'St.         ati (7E+4)                                                  -                1E-3I       1LE-2
                                                                                       -2E...                              IE-7 74        Tamgstan-176           0. all c'ampuds                                                    ff(-5 IE.4                5SE4              2E-S             7E-C             31E-4  I    1E-3 74        Tcmgtaft-177.         D, all compounds               2E14                9f+4                               1E-7             3E-4I4      3E-3 B-34                                                                                  ODCM Rev. 25

j I .- App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION

                                                        . . I .             .

App. R. I.

                                                                              -                        tablt l Vclu

• Table 2 Table 3

                                                                                    -    .    -   occupational Values                  ,Effluent         Releases to Concentrations          Sso~

Col. I Col. 2 Col. 3 Col. 3 Col. 2

                                                                                        'oral                                                            NPonthly Inge~ti~n             Inhalatien                                  Average AtMIc        AadfmnoClide          Class                               AlI                                        Ar          Water    Concentratlon

-N. 00t) , 00/0lI) '(PCU/l) (P01/01) .(vi~/ml) 74 eTngston-178 , 'all capo:nds SE-3 2E.4 &E-6 3E-S 7E-S 7E-4 2

74 .' tungs tan-79 O. all civoundx 5E+5 21E6 7E-4 2E-6 3E-3 .7E-2 74 lungastn-101 D. all coounds 2E14 3E14 - I-5 2E-4 2E-3

                                                   'D 411 comopunds                     2E+3                                      5E t 74          lungstan-lMS            D, oll ce~opunds              -     21.3.           7t13           3E-6 LLI wall (31+3)                                                   *E-S 74         Tungsten-2h7             0. all Compounds                    ZE-3            9E43           4E-6       IE-8        3E-S       3E-4 74          lungstan-lbJ            0. all copevinds                    24E1Z            12E3           SE-7       2E-9
                                                                               .. . LU we1 Ia-(51.Z)                                                   9E1-       7E-5 75          lRh       lw-2772     'O0. a 1 CoMPOU; I  ~One given        IciY                   .4   :      31.5            1t-4      4E1-2E-i W. ex*ides.      hydi ad     ni1trate$                                                     1E-4       SE-7 (11.5)            4E+5 2

75 Abeniem,178 -, aae 377'A 7E 4 1E-4 4E-7

                                                                        . .           St. well         3E#S 1E-2 31.4       4t-7.
                                                            "1771                                                   . 4E-C                    2E-3 73 t~Se ftimw-181           ,    5a     7 lahn11 DIn
                                                                               -      51.3.                           al-I       11-S 1E-6                   7E-4                  .  ,

9SE' 41-I 71.3 IE-3 7E-S 75 Rbvnhs32 " O,177b SE-6 2E-8 IE-1 9E-4 (12.7 h) W " 17 3143 2E+4 6E- 2E-S 91-4 75 lhen1-282 s eeW. 2E#3 3E-9 7E-S 21-4 (64.0 Ii) W:.-&8 1w 2E.I 1E-7 31-9 2E-7 73 bhe"IM-14a3," , 177 11.-6 31-5 3E-4 4E+2 6E-7 ZE*3 2E#3 75 thenap-384 0 se) 2443 *E-9

                                                                                      ;.3              ZE13                      2E-9         2E1-S nUben1156                         .. 177k'                                                     7E-7 75         thsnim-36a               0. set 177k                                                         71-7 St. .wl          St. wall                              31-S (214,3)         (2E43)                       3E-9                   2E-4
                                       ,-.. Ya. I" 27                                           2E1Z                      .2E-10
             *75        Uthen~i     -106         0. s       177                      213              XE43                                             3E-4 21*3            7-7        2E-9 75        therniso-147            D,     tec 1                                          9E#5       - 4E-4                        E-3     hE-2 Z2E*3) 11._              St.-wall (9E S)            41.        IE-6 1E1S                       lE-t 71..3.                           4E-5 75        ktoiumal-liz                        1.7 277                                       1E*S           61-S       2E-7         11-3      IE-2
                                                                                                     .1E+S           6E-S       2E-7 75         khonlnm-188             0 a         I77                                       3E*3           1E-6       4E-9         2t-S      2E-4 WI sa            go                                    3E43            11-6      4E-9 B3-35                                                                 ODCM Rev. 25

II App. B ' App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION Table -I Table 2 table 3 Occupational Values Coflultn Releases to Concentations Sewers Col,.I Cel.2 Col. 3 CCol.I tl. 2 Oral Monthly Ingestion Inhalation Average Ataefc Radionultfde Class ALI ALI - C Afr Water Concentration No. (PCi) (PM) (PCI/ml) (pCi/ml) (pCi/al) OpCi/al) 75 Rhienlwrlag a. see 177' W. set 17710 31.3 1SE3 2E-6 7E-9 4E-S 4E-4

                                                                             -      .        4*E3           2E-6      6E-9       -            -

2 76 Osali-1B0 0. all coepowns except those given for V and Y 1E*5 4E#S 2E-4 51-7 IE-3 IE-2 W. ha ides and nitrates 'SE+S 2E-4 7E-7 Y. oxides tnd hydroxides SE#S 2E-4 6E-7 2 76 0salm-181 . .see IsO 1E-4 4E-4 2E-5 6E-8 2E-4 2E-3 r: se n SE.4 2E-S SE-t 4E14. 2E-S 6E-8 76 Ossitum-12 0° seeWO 2E13 2E-6 bE-9 3_-5 _14 4E*3 ZE-6 6E-9 4E#31 2E-6 6E-9 2E*3 2E-7 3E-S 3E-4 7C~t 0"1u_. . .IE* aE*2 3E-7 lE-g 76 Osaium-189 0 see Y. set O

                                                          -Os                                              3E-7       6E-9 ZE+2 81s4            6Ef2           1E-4      31-7       1E-3          1E-2 i                                                                                                           SE-S      3E-7 II 3I                                                                                        2E+S           7E-S      2E-7 II               76         Osaitm-ig        Y. seesee       0                               3E*4           11-S      4E-8       2E-4         2E-3 C                76 76 0 Os I 9I n-0 Osisfa-ISI * ~

0

                                               . see lad-os e la.s                   1LI.41 2E-4 2F44 OE-6 7E-6 3E-6 2E-D 2E43 4.1.            9E-7      3E-9 7 6Gsaki n-193       W. see 8 ,                      (1.E3)           51.2                               3E1S         3E-4 r1: see                                         2E-1                      2E-9 LLUwellI     -1E43             71-7 nIrld1XIA22         D. *1set          unsexp SE-3                     SE-9 LLI well W. seets           ~ttx          (21.3) 11-6                 21E-S        2E-4 Y: see                                         3E#3           ZE-C 7E-6     4E-9 Y. see Iwo$  O                               3E+3           CE-6     4E-9

• Vsfm19

                                               . see 0 76         OsIrin-194       Dr see MO$                      4E.2            JEIM           lE-C      6E-U LU wall 616E2)          4.3                                8E-6         8E-S U-U
                                           *. *t        Igo                                              3E-9       IE-ll 77         !vridiaa-*       0. all cwvowmd except
                                          *those given f or Wand Y        4E14            114S           6E-S       ZE-7 St. well (4E14)           -                                           4..-   61-3 W.' halides. nitrates.

and metallic Iridua' - 2E1S 2E-S T. oxides and hydroxides - . 1145 11-S 2E-7 6E-5

              .77        Iridiun-18M       0 see92r                        8E13          2E.4             11-S      31-S       IE-4          1E-3
                                                                          -              3E.4            IE-S       SE-8 v: se          Ir               -             .3E14                       4E-6 B-36                                                                 ODCM Rev. 25

App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION Appila. , Table 2 Table 3 Occupation~al Valuai fflusnt hRleaws to Conc antratiani Sewers Col. I Col. 2 Col. 3 Col. . Col: 2 oral aonthly Averae Atomic kadilnucll dt Clogs ALI AU DAG Air Vater Cornctration I) (PO/91) (WC111) 77

                           'Ir diumM          D.

18 2 1r 51.3 IE*4 SE-6 2E-8 7E-5 7E-4 IE+4 SE-S 2E-B W. 1" 1E.4 4E-6 -lE-8 61+3 3E-C6 It-1e 3E-S 3E-4 6E#3 3E-C 9E-5 77 Irfidfa-187 D. See mlr 36E3 2E-6 eE-3 W, see 2Wit 2E+4 3E14 1E-5 51-5 IE-4 1t.1 lE-S 4E-C If.see Jr 3E+4 2E-5 4E-S SE*4 77 Iridlsa-287 D.'$" JIt 21+3 2E-6 6E-9 3E-5 3E-4 11.4 4E*3 1E-6 SE-S 1s e.2 21 3E#3 2E-6 51-5 SE43 2E-t9 7t 1 77 Iridlim-2119 P. so*. T. seer (SE43) 41.3 71-5 71-4 21.3 4E13 2t-6 5E-3 4E43 If-6 5E-5 77' Iridium-M '.seOI 2E*5 tE-5 3E-7 21-3. 2.-2 9E-5 3E-7

                                                                                      -2E#S 77          IrtdiIu1,29OZ        0:

D o 21 r 2E S 9E*2 t1-1 4*-7 3E-7 1E-S 7E-S 7E-4 Ii I Vsee I9tr IE*I 4E-7 lE-5 9f*Z 4E-7 11-5 77 It dluim-190 b. 1k 11 (5143) 31.E 3 *4E- 11-10l 41-5 4E-4

• 2f*2

                                                                              ,14. 2E#I           IE-8           31-10 Jzr                                    6E-9           2E-ll                                          I 6142         IE*2         1E-7           4E-10         2E-S.       2E-4 tE*z         2E-7           61-10 77    -. Iriditum-13        D. see       Wit            2144                       3E-6           3E-10 31..3         3E.2 IE42 2E t       *4E-C    .       il-la W-11 41-S          2E-IO
                                                                                                             , 11-10 77          Irfillsm-155      'D            82i            81E3
                                                                   *11.3                          4E-e          lE-lO lt+4                        f-76         4E3            11-5       21-4 2E*I           WE-          3E-1
           .77         Irfdlam-1SS           .I"        O          ,,      3        21*1         6E-7           3E-3 14sea        WIT                                       1 1-5           E31-U        11-4      .1E-3 11-5 "        .41-2E 4         lE-6           34E-J 78         Platln w-1950      0 , sel co so nd                          41.2          21 2f-5      .      1-C6E-5.          2E-4       2E-3 2E+4       2E-5          7t-8 7 8 la tm -1 8       v a l lm c1      T 1                      4EO4         2t-S            WIE 78Plttmm-19           0 al           Jr
                                                                   - 11.4           2E+2         ZC-S 2E.    .

SE-61 51- 2f-4 2E-3 7E-7 2E1- 21-5 21-4 78 Platinsw151,8 0. allI caqsounds 41.3 3E-4 4E-S 4E-I 11-;4 11-i 31.3 41-6 11-5 SE-S lf-4 B-37 ODCM Rev. 25

II App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B

                                                                 -                        Table 1                            Table 2                 Table 3 Occupationul Values                        Effluent              Releases to ConcentratIons                 Sneers Coll           Col. 2           Col 3        Cole1         .Col    2 Oral                 n aothly lvestion           Inhalation                                     -       Average Atomic    Radionuclide         Class                             ALI             ALI             DAC          Air              Water     Concentration No.                                                             (pC0)         (pCI)        (pCI/e1)    (pCI/al)           (pCi/el)    (pCi/el) 5       Pltiamnea-               aa1Icowounds                 3EtJ.           bt-3              E-6        agy             _-

LLI wall (3144) _ 4E-5 4E-4 78 Platins-193 0. all cowounds 4E14 2E#4 iE-S 3E-S - LW wall (SE14) - - - 6E-4 6E-3 78 Platinam-9Ss a allcowmunds ZEf3 4E13 2E-5 6E-9 - - (2E13) - - - 3E-5 3E-4 78 Platinma-197m2 0 all compounds 2E44 4E14 2E-5 6E-8 2E-4 2E-3 78 Plainm-i97 0, all compounds 3E13 1E-4 aE-S iE-S 4E-S 4E-4 2 7E-3. 78 Dlatfres-199 0. all compounds 5E+4 11E5 6E-5 2E-7 7E-4 75 Platine200 - D; all cuds 11E3 3E13 iE-6 SE-9 2E-5 2E-4 79 Gold-193 Jr. al compunds except thsesegieie forW and V 91.3 31+4 1E-5 4E-S 1E-4 1E-3 W. halides and nitrates - 2E14 9E-6 3E-S

                                  -Y      oxides and hydroxides.       -             24                8E1          3E-S            -             -

79 told-194 D.- 1s31b 3E13 81.3 3E-S 11-8 4E-S 4E-4 W. an I;?,,. SE13 2E-6 .E- - -

                                        .se4e       A                                51E3             2E-6         7E-9        -                  -

79 Cold-ll" 0, cc M w 51-3 11E4 SE-S . 2E-S9 7E-S 7E-4 W,S- 1 - . 13 4 6E-7 2E-9 - -

                                      . see         u                -               4E12             2E-7            110          -             -

79 Cold-19M o s . 11.3 3119 .E-S 41-9 lE-S 11-4 Y."19 _i 1E*' SE-7. ZE-9 . - 7CldlS asoe 11.3 41*3 I1-5 WS- 21-S 11-4

                                     . san        A                 -              21E3             SE-7.        3E-9             -

Y. s l Au - 2E13 7E-7. 2E-9 - - 1 93 79 Gold-In o ago k 3E+13 13 41-S 1- - - LU wall

                                                                          +3)                        2E-S                 -              S      4E-4 4A43               -          6E-91 1E-an 132:-413                                              21-4         5-79         .old-20           0      Si     193                  iE+3           4E13              11-6         SE-9           21-S          2E-4 0             19a                                 3E+13             11-5         4E-9           -             -

rE*, -1 214 1-6 3E-9 - 79 Cold-2002 19933a 34 4 3E-S 9E-i 4E-4 4E-3 V. se -E+ 74 3E-S 11-7 - 17E4 3E-S 1E-7 79 Cold-201' 0, ste 1sAis. 21+5 SEl-S . 3E-7 . - 7E+4' - St. well T. see "AI (9E*4) - .- - 1E-3 11-2 w "193 - 2E1S 1E-4 . 3E-7 - Y:.. " Asil 2E1S 9E-S 31-7 - B-38 ODCM Rev. 25

App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION --AV .B - fable I - Table 2 Table 3 Occupwetional Values Effluent Releases to Concentrations Sewers Col. ' Col. 2 Col 3 Col.1 Col. 2 Oral NIonthly Inostion Inhalation Average Atomic Radionuclide Class 'ALT L-X -o .AAir Water Concentration No. tPCI) (vCI) (pCi/ml) (pt/0l) - PtI/0l) lf)iC/8l) 80 , ercury-193a Vapor 83

• 4E-S6 1E-8  ; -

Organic 0 4E[3 1E*4 SE16 2E-1B 6E-S - 6E-4 0, sulfates 33 3 9E43 4E-6 1E-8 4E-S 4E-4 W. oxides, Iiydrn)X ides. ' halides, nil tnrtsIs s; and sulffdas BE+3 31-6 11; aE+d 8o Wercury-193 Vapor 3E44 4S-8 - 31-5 Organic 0 6E+4 9E-E 3E 3E-3 3E2-S ,.-4

• 0.... Ng 4E+4 i E-8 2E 2E-3 2E-4 64-8 -

so Vercury-194 Vapor W. see 0 Organic 3E[. 3E11 IE-8 1 41-11 4E-11 2E 7-7 , 2E-6 4E31 41.1 21-8 ,E-11 1E 1E-4 1E+2 2E-10 - I,. r., 1,1 2E43

                                                                                     ..3 4*.3 6[.3 5E+3 IE-6    ;

61-9 8E-9 71-9t 4E-

                                                                                                                                                  .5 3E*.5 41-4 3E-4 I-0                                         Organic 0                                                           -.2f-6          U-9 4.3E*

IE . . i0 Mercury-iS ' Vapor W. see,2)3S; 2E7I E34 4E54 4E-6 61-S 21z.4 W, Gn s.see e U1-S i2E-3

                                                                                                                .21-1                             .4 3E*4 SE-4                           SES            21'          2E-3 s0       liercuy-197         Vapor                                 (11'S)                                        7E-9 Organic 0                                            9E*3             21-6          U1-S           SE..5 E5)                                                                     4E-4 7E.3                           1-           ,4E- .5        ,E-4 SE+3                           7E-9           -

21-6 to Percury-197 Vapor. 8E+J

• W,.see Organic isO 2E#3 ZE-8 9E-.5, 1144 4E-6 21-S SE-*1 4E-4 91.3 11-8 2 (7.4) i80 Nercury-19 Vapor 5144 71-5 1E-7 -

Oranic D 21.5 2E-7 - St. wall 1E-'3

                                            . see 6E.4 ZE S 11.5

• 21-7 8E- ~4. 11-2SE-3 7E-5' 2E-7 - .3_

         .8           lherery-203         Vapor                                               81.2             4E-7            E-S          7-Organic a                                                            3E-7          lE-9           ?E- 6         'E-'

19 3 :1. 11.3,

                                        - D, *o            iig                                                 51-7          2E-9           3E-1E-1            *5E-7          2E1          .

84 all eounds 1144 6E11 2E-7,

                                                                   *           *St. well
                                                                                                                             -              Ii-3          1E-2 B-39                                                                        ODCM Rev. 25

                                -I PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION                                                                                   pP:

• App. B Table I Table 2 Table 3 Occupational Values Effluent release$ to Concentrations Sawri Col.1 Col. 2 Col.3 Col.1I Col. 2 IngrStion Inhalation A ve on Class. Atl ETI D^r Air Water Concentration Atomic Radionuclide lpti/al) (pCa/l) (RCi/0l) (PCI/sl)

NO. I (pCt) . (pfi) Thalliu.-194 2 0D al1 cowounds 3E-S 6E+S 2E-4 tE-7 - 8l St. wall 4E-3 *E-2 E (3145) - 2 0D all conpounds 6E14 1E15 5E-5 2E-7 9E-4 9E-3 Sl Thallium195 0D aIl compunds 7E14 lE5 SE-5 2E-7 1E-3 1E-2 tl8 Thallium-197 2 all co s 31-4 5E-4 2E5S 1 44E4 4E-3 E1 Thallium-1am a De all compounds 2E14 3E14 E-S 5E-B 3E-4. 3E-3 S1 Thallium-19S

0. all compouods 6E*4 8E*4 4E-S - '1E7 9E-4 9E-3 tl Thalliumw199 0D all C.oIF-ds 8E3 1£.4 SE-6 2E-B 1E-4 1E-3 81 Thailfiur200 D. all coepounds 2E14 21E4 9E-6 3E-t. 2E-4 2E-3 SI Thallium-201

0. all c ounds 4E+3 SE43 2E-6 7E-9 .51S5 5E-4 81 T Illt-M202 -

D. all emounds 2E13 2E+3 i9-7 3 E9 21-5 2E-4 si Thalliiu204

                                           . -all co.ounds              6E44             2ES                S E-1     3E-7           E-4. BE-3 v-          U         Lea4d-9Sm2

'aL D *all compounds 3E-4 6E+4 3E-5 9E-t 4E-4 4E-3 U Lead-li a 2 D,a*1 caoll ds 2E44 7E*4 3E-S 1E-7 3E-4 . 3E-3 U* Load-299 U eied-200 D; -a1 eopounds E*3 6E*1. 3E-6  : 9E-9 4E-S 4E-4 82 D. ll cmpounds 7E3 2E+4 8-6 3I-t 1E-4 1E-3 W 32 itrd-201 .

0. all coounds 9513 3E14 1E-5N .41-S 1E-4 1E-3

82. Lead-202m D.*all cmounds 1+2 5E+1 2E-8 71-11 2E-6 2E-S 82 Lood-202 0, all comounds SE3 9E43 4E-6 1E-B 7£-S 7£-4 82 L'ad-203

0. all cosauds 4E13. 11E3 6E-J 2E-9 5E-S SE-4.

82 tead-20S

                                           . all Cooouds                  244             6E4            2-1      8E-t         3E-4       3E-3 82       Lead-209 6E-1            2E-1              10      -            -          -

82 Lead-210 0Dall counds Sanesurf Sonesurf (l t (4f-1) 6E-13 1E-8 1E-7 2 D. all compounds 1E-4 1.2 3E-7 9E-10 2E-4 2E-3 82 Lead-211 . 82 Ltad-212 0. all covm'ds 8AM1 31.1 11-8 SE-l- - - lone surf (1E 2) - - _ 2t-6 21-S 82 83 Load-214 Bismuthb200 2 2

0. all cowmpad D; aitrates W, all other coMPouns 9E.3 3144 51.2 1E+4 1S 31-7 4E-S 4E-5 1E-9 11-7 1E7 1E-4 4E-4 1E-3 4E-3 II se 20 1E-4 3E-4 lE-S 4E-8 2E-4 2E-3 83 Sismuth-201; 3I 41*4 2E-5 5E-8 - -

V see i C: "'::o- 2R ,

                                         °                              1E-4             4E64          2E-S       6SE-          2E-4       2E-3 t3       Blsautt 2022                                                                                   1E-7         -           -

_ E*4 3E-S B-40 ODCM Rev. 25

A 2T . App. B App.A"' PART 20 STANDARDS.FOR PROTECTIO'N AGANST'RAD"IATION

                                                                                      -       Table I                        Table 2            Table 3 Occupational VIblues                   Effluent         Releases to Concentrations            Sewers Col. I             Col.

• Col. 3 Col. 1 Col. 2 Oral Monthly Atmi adoucl.

lasIngestion nal inAverage Atomc' CassALI Rdiouclie' ADXCn Water. - Concentration Air,' No. (p~~CO) (C) OpCI/01) (pCI/01) -(PCi/Mi) (JjCi/M1) 83 Bissauth-203 2E13 YE+3 3E-6. 9E-9 3E-5 3E-4 _ E+3 3E-6 9E-9 - - v 20%0 S3 lisouth-2OS W: 1-Rl - 1E.3 3E+3 1E+3 IE-6 SE-7 3E-9 2E-9 2E-S 2E-4 83 Slsauith-206 6E42 1E*3 6E-7 2E-9 9E-6 9E5

                                                                                 -9E-2                    4E-7         1 U-9            -          -

8.1 Bifsuth-2O7 1143 ZE .3 7E-7 2E-9 lE-5 1E-4

                                                                       -              -    4E42             1E-7     5E-l0          -          -

0. :0* RI, 83 3 ismutI5-210i' 4E~l SE00 2E-9 - . -

v. see f Kidneys KIdnays (UE41) (61E0) - 9E-12 BE-7 6E-6 7E-l 3E-10 9E-13 - -

W. Ste 20%i 53 ellimth-210 81.2 1Z42 IE-7 - 1-5 1E-4. Kidneys,

                                                                      -                  (4E1Z)                     SE-O0.          -         -

aI 3E+I 1E-5 4E-1l - - II 2 W.ato 20I i6 o 83 63 Wsm atl-222 Blsautfr213 2 W. see 20%f SE-3, 7 1,3 3E*2 iE2 21.2 11-7 11-7 1E-7 3E-10 4E-10 4E-10 7E-5 IE-4 7E-4 U-3

                                                                     -                   '4E.2            U1-7     5E-10           -          -

2 83 liamuth-n14 21-4  : E42 3E-7 31-9 - -

St. wil (21E4) - - 3E-4 3E-3

                                                                     --                   9E-2            41-7      I3-9             -

2 V. Sami gi by d ual e 84 Poloniuam-203 3E*4 6E14 3E-S 9E-S 3E-4 3E-3 9E,4 4E-5 iE-7 - - 203 2E14 4E+4 2E-5 SE-0 3E-4 3E-3 64 polonlsar-205 2 - 7E+4 3E-5 1E-7 - - 8 143 . 31 4 3E-5 3E-B IE-4 1E-3 34 FPotnlum-21fl 0: meIk - 3E+4 lE-S 3E-10 4E-S

                                                                                                                   *E-23 4*-8
                                                                                                                                             *E-7 31E-                  6E-l
                                                                   -                      6E-1          3E-10      #-13            *        -

2 as Astati nbb-20~ *_613 3E-3 lE-S 4E-1 E-5 DE-4

                                                                               -21.3 2E-              31-7. 3-4            -         -

as Asetaint-2UI U. ha]lIdes 11 2 1.+1 31-a lE-lD 21-6 2E-5

                                                                   -                 - SE+              2E-U      S-lI            -         -

86 Radeon-220 With dauightars

                                                                               -2E+4                    7E-6       2E-5           -         -

With daughters present t -.\- F .E+1 4E-9 3E-1l - (or 12 working (or 1.0 level nthss) working lee)? B-41 t ODCM Rev. 25

II . App. B PART 20 -STANDARDS FOR PROTECTION AGAINST RADIATION - pp Table I Tabie t. 7ab e Occupational Values Ef luent Releases to

                                                                                                             -          Concentratlins           Swers Col. 1            Col. 2            Col. 3        Col. 1         Col 2 Oral                                                                       monthly Inoastion               Inhalation                                         Average Atomic    Radionuclide     Class                        ALT                ALI             °             A              Var    Coneantration Nn.                                                    (IStI)            (VCI)       (pCI/m)       (PCI/1)          (PCI/1l)  (PC1/al) 86       Radon-222         ViLth daughters reeed                       -                  1.E4             4E-6          1E-            -         -

With daughters present - 1E*2 31E- L- l10 -- I.-A -14... I..fl51 level) 2 8? Framcum-_22 2 0. all copound 2E13 SE42 2E-? 6E-10 3E-5 3E.4 2 8? Fransciim223 D. all cotmpnds 6E.2 8E+2 3E-7 1E-9 tE-6 tE-5 go Radiur-223 W. all comounds SE1O 7E-1 3E-10 .9E-3 - -

                                                                -#one        r       -                 -             -               11-7      IE-C as        Radfum-224      VWall ceoexds                   1E+0            2E-0             7E-10         2E-12 go" surf (2E11)                -                -             -              2E-7      2E-6 8S       ladiun-22S       V. all c1 ounds              8E10              7E-1              3E-10        9S-1            -1 oson surf (21E1)               -                 -             -              2E-7      2E-6 I.         8J       Radui-226        Wall W                e           2i-0              6E-1              3E-10         9E113                    -

saoe surf (5140) - -8 6 6-E1-7 I1i 2 88 tediuw-227 W, all Coo11d5 2E-4 11.4 6E-6 - so" surf Ion, surf I (21.4) (2E14) - 3E-t 3E-4 3E-3 . . 0 88 RadItu-228 W. al cpouunds 2Eo0 1E40 SE-10 2E-2 - Bon surf (4E*O) 6 E-8 WE? 89 Actinisr224 0. all coounds *xc pt those ulvea for V and t 21.3 3E-1 lE-S-LLI wallI Bone surf (4E - 5E-1 3E- 3E-4' W. halides and aitrates ' 5141 2E-6 7E-ll - - Y. ouxide and hyriida -. 5E1 2E-6. 6E-U - - 89 Actinltu-225. 0, set 224AC 5141 3E-1 IE-10 - ULLwall Bond surf (51*1) (5E-1) - . 7E-U 7E-7 7E-6 W. 2 2 4Ac- 6E-1 3E-10 5E-13 - - Y. soe A4 - 6E-1 3E-10 9#-1Z - 22 t9 Actfnium-226 0. see. Ac 1E2. 3E+0 1E-9 - _ LLIa

• B"o surft -

224 (12) (41) - SE-12 2E-6 2E-S W "S 2 24AC SE10 2E-9 7E-1Z - - Y: *a Ac - . 1O 2E-5 6E-12 - - 89 Actiniu-227 O. sea 224AC 2E-1 4E-4 2E-13 Bone surt Bonoaurf 22 (4E-1) . (B-4)  ; 1l-lS SE-9 SE-8 W. lee 4AC 2E-3 71-13 - - - B" savr 22 Y (3E-3) - 4E-15 Y. see 4AC- 4 1-3 2E1-U 61 15 - B42 ODCM Rev. 25

App.B PART 20 STANDARDS FOh PROTECTION AGAINST RADIATION Tabil 2 lab)c 2 Table 3 Occupational values Effluent Releases to Concentrations Sewers Col.1 Col. 2 Col. 3 Col. Col. 2 Oral onthly Ingestion Inhalation Average. AtomIc Radlonuclide Class -ALI All DAC Air Iater Concentration (vC') Wc. (VCI) (MCI/ml) (MCI/m1) (vCI/"l) (MCi/ml) 89 ActinIu-228 0 set 224Ac 2E13 9140 4E - 3E-5 3E-4 loan surf (2E11) 2E-ll1 - V set 224AC 41 2E-8 U.see 22Ac aEox surf (61.1) &E-11 - - 4E11 2E-S *E-ll - 2 90 Threm-226 W. all cO XIard those given for 2E2 6E-8 2E-10 St. wall tSE*3) 7E-S *7E-4 Y. oxides and t l(droxides - - E112 *E-S 2E-10 I . E-. 0 thor1 m-227 VI see -^. 2E+Z1*1 SC. 3E-1 lE-10 SE-13

• 2E-6 2E-S 3E-1 l1-10 1E-13 SOtht1_2f V gote 2 21 b I aae 6Th 6E+8 1E-2 4E-12 gone surf Bone surf 22 (2E-2) 3E-14 2E-7 2E-6 YSx STh 2E-2 7E-12 2E-14 90 tbor1_r22S W. See 225Th 6E-1 9E-4 4E-13 II $ one sorf IMonesurf 2E-7 0I t(E+0) (2E-3) 3E-14 2E-e Y. u 22^b 2E-3 lE-12 gone surf (3E-3) 41.0 3-t 90 Thorlu-230 W. See 226Th $E-3 3E-12
• lone surt Sane surt 22 (9E 0) (2E-2) 2E-14 -1E-7 1E-6 Y se Th 2E-Z SE-12 San surf 3-5 t2E-2) 3E-14 5-s 90 Thorsem-232 22

                                                              *4E+3                    SW.3            3E-S

• 9E-9 SE-S SE-4 Y. *. 6T 6E-3 '3E-S 9E-9 226 90 7hrlm-232 Y. s Th 7E-1 1E-3 SE-13 lon surf gone surf (21E0) (3E-3) 4E-IS 3EM-8 3E-7 rse 221j 3E-3 2E-12 Bone surf (40-3) 6E-1S 90 7hor1_ie234 I ,W 226Th 3E-2 2E+2 BE-4 . 3E-10 LLI wall -

22 (4E+2) . . SE-6 SE-S Y. set "Th - 2E#2 SE-8 *'2E-10 2 91 Protactinier-227 V. all coounds excpt those given for V I 4E+3 11E42 2E-10 SE-S SE-4 T. xide and bydromidei l- IE+2 *1-8 91 Protactinium-228 V. se 227 Pa -E3 1E*1 lonesurf t(2E11)

                                                                                                .'. 5E-9           lE-IO 3E-11 2E-5      2E-4 Y. se     226p                     -                  1141           SE-9          2E-II B-43                                                                                ODCM Rev. 25

App. B PART 20; STANDARDS FOR PROTECTION AGAINST RADIATION App.I.Bei

                                                                                           .                                  T . le 2             Table 3

atistionl volu n Efflutt. Release to  % Conetntrstians Sewer S Cal 1. Cal. 2 Cal. 3 Cal. 1 Cal. 2

                                         ._    _0                                             Intaletibn                       -            \jnthln

__ ___ __ ___ionv re v ALZ1C XAaanucijd* class Ar - ALI. - - Air -- ite . Concontration

               .*                                                       C).             (UaM)        t td/l)        (Ci/0i)        (Ytl/nl)     (jaCi/ul) 2 Sl       Protactinium-230 V. see 2Pa                    6EZ                St+.               2E-i:       7E-12         -             -

SBam surf 1 ^ 227k 91.2) E11-S IE-4 4E O IE-9 5E-12 91 PF tsctinfr-231 y.. 227p 2E-1 2E-3 6E-L7 - Bone surf Soft surf 1'5E-1) (41-3) _ 6E-1S. 6E-9 .6E-8 r,.ee 226so 2E-I1 . - sane surf (6E-3)

• 6E-15 91 . Protactniua-232 V. sre 227t? 11+3 2E+I 9E-9. - 2E-S 2E-4 Y.* 227 S0ne ruurt (1.E1) 21-8 _

Bana soarf

                                                                                       £1£.

(71.1) 7 -i 91 Protactinitor233 V; a" 22P 71.2 3-7 lE-Ii U1Iwell 22 2i-5 1-4 II Y. se 1pa - !E+3 23-7 BE-" I 91 Protctilnum-234 WV 2se7 21 £43 3E-6 1E-6 3£-5 2E-4 71.3 3E-4 *E-9.

         -92        Uranim-230         0 uFO-. UW*Fj Ws,(NCS)2       1E.0

( tone 4E-1 2E-10 - surf . 8aen surf 16E-1) _ BE-13 8E-6. 8E-7 V. U03. UfN UC14 4E-1 11-10 5E-13 T. Wt. 11303 U3-1 1E-10 4E-13 92 Uranir-231 D sse230U , SE43 8E+3 3E-6 IE-8

                     * ~        ~,     Y, O~t,230 uo B4E'3)

(4 6E-5 6E-4 6E+3 .2E-6 6E-9

                                      .r: .'        0..
                                                    %                                 5E#3              2E-6          6E-I 92        Uranlu_-232       0. A, 23                                       ,2E-               SE-11
                                                                    ---

• surtf 80onosurf

r. "t 23%.t  ; ' (E-I) _ 6E-13 61-8 61-e 4E-I 2E-10 SE-13 6E- 3 3E-12 IE-14 (2

E+1 lE#0 5E-10

• ae sull Bone surf 9Z Urfu_2 Dled 230 2Eel) (2E+Ol I 3E-12 3E-7 -3E-6 lt-10 lE-12

Y: See3 -4E-,2 2E-1 5E-14

                                                                                                                                                                     -i 92        Urentue-2343      a. see z3u                  11.E1            .lE0                 5E-10        -

3 one aurf tw* surf II Y:Urn V, 0,X23 s2e 2see Z (2 E*1) (24E4) 3E-12 3-? 3E-6 i 7E-1 .3E-10 lE-12 4E-2 2E-1I '.#114 B-44 ODCM Rev. 25

App. B App. B PART 20 ,STANDARDS FOR PROTECTION AGAINST RADIATION.. Table-I Table 2 Table 3 Occupational Values Effluent Release% to Concentrations S5wers Col. 1 Col..2 tol. 3 Cal; 1 Cal. 2 Orcl 1onthly Iigestlon Inhalation Average ALI Air Water Concertrat1on Atomic Radionuclide Class 23OOWNl)

                                                                 -sC        1V1)

t. ( .Pc-/-l) tPCI) 1pC0/.l) 9e I1:4I lE*0 6E-10 - - -

                '92      Uraniui23S3                 -lsee230U-Senc surf Sone surf (2E1)            (2E#.)                       3E-12       3E-7    3E-6 S E-1          3E-10      lE-12 23 0              _-                 4E-2           2E-3 I     6E-14 Ve. U 92.:     (ranhim-23S           Os.  .se              '      sn. surf          Sone surf SE-10 (2E*1)           (2E40)                       3E-12       3E-7    3E-6 3-0 V

W " 230 eE-l 3E-10 lE-12 2E-1 SE-14 - _4E-Z 0 .set E213 3E-3 - 1E-6 4E-9 92 IJra nii-237 LLI wall 3E-4. (2143) 3E-S 71-7 2E-9

                                                                                        .21*3 2E+3           2E-7       2E-9 92  ~ ~        ~     ~   . a      23nim23 1.1               l1EO       :   6E-10 Bone surf- Bone surf (2E1.)           (2E14)                       3E-12       3E-7    3E-6 Ulraniwm-238 2

0. 2ee - eE-l 6E-10 z-n 1E-32 92

_ 4E-2 6E-14 7E14 2E+5- tE-5 3E-7 9E-4 9E-3 D, VYset 23U - 2E45 2E-S 21-11 2E-7

                                                                            -.                2E+5           SE-S       2E-7 6.1                                                                        11E3             .4E13                      5E-9        ZE-S    2E-4
               '92       Upanioma-23tra2              ee 2.

_3E-3 4E-9

                                                                           -                   21.3           1E-6       3E-5 2

33Nptensm-22 V all cooU SE-10 Sone serf Bone itur (2E11) (2E+O) 3E-12 3E-7 3E-6 _ 1-1 3E-10 9E-13

                                                                            -                 SE-2            2E-11      9E-14
            . I                                                            1E-5              2E13            iE-7                  2E-3    2E-2 Sone surf
                                                                            -              (5E12)                        -E 4E-3 2                     .wmds       SE*5              3E+6            IE-3                  1E-2    lE-1 93     - kotzanlue-i33          W. all c 2E13               3E+3           11-6       4E-9       31-S    3E-4 93         Neptunius-234        W. all c'ompds W. all com        ds  .      21.4               BE*2           3E-7 93            tunis.-235 -

Ne16o tLl wall Bofe surf (21.4) (lE.3) 2E-9 3E-4 3E-3 V, all c~o~mpors 3E O 2E-2 9E12 93 IlP tonigum236 Sane surf Sone surf (1.15 'SMY). (6E1O) (5E-2) 2E-14 9E-S 91-7 93 Ipeiu2B

                                                                      -     3t.3               3E#1           11-S W -all cwounds
                         .(22.5 h)                                          Bone srf          Se surf (4E.3)            (7E.1)                       11-10      SE-S     SE-4 ao.
                                              'W. all cpud                  SE-1.              4E-3           2E-12 I93        Meptanla-237                                                          Bone surf Sone surf (lE.O)            t1E-2)                        lE-14      2E-S I  .

B45 ODCM Rev. 25

I I App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B , Table1 Table 2 Table 3 Occavatioml Values Effluent eleases to Cencuntrations' Severs Col. 1 Col. 2 Col. 3 "Col. 1 Col. 2 oral Monthly livestion Inhalation Average Atomic Radlonuclide Class ALI ^ETP D' Air Water Concentratioi I WO (~) ON1 tvC/4l) 00C/8l), (1uCi/ol) ,pCf/ml) 93 Meptunhua-238 V. aull counds 1E3 lo41 31-S 2E-5 2E-4

                                                                                  -(21.2)                                  2E-10        -

93 Weotunlum-239 V. all c hoind 21E3 21.3 9E-7 3E-9 - LLS wall (21#3). - 21-S 2E-4 93 Meptunium240Z V. all comounds 2E+14 4.4 3E-S 1E-7 3E-4 3E-3 94 Plutoniu.-234 W. all ceounos eatxct Pt U 3 21.2 9ES- 3E-10 1E-4 IE-3 6E-6 31-10 23 4 94 Plutonh-232 W, sit h 1.5 146 11-3 4i-6 1E-2 IE-1

                                                                                                           . IE-3.       3E-6          _

2E-0 2E-2 lE-12 Boat surf ton surf O.- (4E*O) .(4-2) SE-14 6E1 6E-7 In - 4E-2 1E-11 1-5 6E-14 - 94 plutonuw-23J7 " 2Pu 1E-4 31*I 1E-6 SE-9 2E-4 2E-3 _ 3E#3 1E-' 4E-9 - uI. .U 94 94 lutflflh238

                       . ?4toluipb7          *n 23' Pu Y1 sea                  .I    9#-1 bon surf (21E0) 7E-3 toer surf-(E-2) iE-12 2E-14        2E-S
                                                                                                                                                . U-2E-7 234                                                         tE-12       2E-14        --

tE 1 E-Z 2 34 94 PlutoniuN-239 V. ae . hp SE-1 61-3. Sone surt Don Sult llEdG) llE-25 2E-14 2E-8 ZE-7 Y. see 231tp P - 2E-2 3E-12

                                                                                            *of. surf (1-E2)                        2E-14        -              . A-54         Plutonitm-240  W. S 234h                       E-I                 6E-3             7E-12
                                                                  ,ow surf tSom %srf (lE*O)            . UE-2)                          -  Z£-14        ZE-5     .2E-7 Y, aSeenp                : -                        2E-2            IE-12
                                                                        . I                   ofe surf
                                                                  *  -                   (1E-2)'                                                                        . I 94         Plutoal  e241   V seq 2 PU W.                           4E-1                   8 3E-1             7E-10
                                                                                                                                                   .- 3 2 34             -{7E*1)                 16E-1)             lE-14         31-13'       1-6 Y.l*n Y. Se       h U     .
                                                                                 -            E-1                                                                       . I}

Son. surf .

                                                                               -         (11.0)                          1E-12       -

B-46 ODCM Rev. 25

• App.B App.B I 0 PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION -

I Tabi 2 Tablv 3 1..

• Atemic Radionuclide io. - -

                                                -Class
                                                                                . . cDall.

rClj ITJStial AL (iiCi) Occupational Values Col. 2 Inhalation Col. 3

                                                                                                                                    ,Air 0(pCi/l)

Effluent Concsntratiens aCol.1 Col. 2 water

                                                                                                                                               ,lpC1h/l)

Releases to sers oenthly Averae9 Concentrat10n (Pci/mi) 94 Plutonium-242 see 234pu 7E-3i 3E-12 Sogm surf Bon, surf

                                                                          ,,       (11.0).          (IE-2)       -      -            2E-14         2E-8       2E-7 Y.StZ341`                                       2E-2               7E-12 B0nc surf (2E-2)               -            2E-14. --

2 94 Pluton1mr-243 W. as. 3'P .2E 4 4E14 21-5 SE-a 2t-4 2E-3 234 Y. e Pu 4E14 2E-5 SE-8 i4 Plutonium-244 V Se 234pu 'BE-1 7E-3 3E-12 S6ne surf See. surf V. Se 2 34 pu. * (21.0) (1E-2) - 2E-14 .2E-S 2E-7 2E-2 7E-12 ge surf (2E-2) - 2E-14 - 94 Plutonium-245 V, sa 234Pu 51E3 2E6 6E-9 3E-5 3E-4 Y.a.. Pu 4E13  ; 2E-6 6E-9 - 94 Plutonim-24 -W. Sao23 4 Pu 4 12-3 3E12 IE-7 41-10 - _ ILI %all

                            ;V     ...                         u      ..        (4142),                                            -              6E-6      6E-5 a                                        -   .se    234f        ;                               3E12               IE-7        4E-lo        -

II 95 Amriciui-237 2 W. li toped Of.4 3E15 1E-4 4E-7 lE-3 1E-2 2 a 95 Amercism-238 W. all co4p e 3E13 tE-6 - SE-4 5E-3 oenc iurf (6EI3) 9E-9 - _ 95 Aaridti239 Y. all e.mounds SE43 1E14 -5E-6 2E-S 7E-5 7t-4

                                                                            *       .2.

95 . rici m-240 W. all c eomd5 .2E 3 3E13 1E-6 4-9 3E-5 3E-4

                .l5          Aricium-241        V. all cespounds                    E-i            tE-3               3E-12               .-

Bee, surf. BSne surt (1E-2) - 2E-14 2t-0 2E-7 7. s Amriciu-242- . E-1

                                                                             - (LE4a)              5E-3              3E-12
                                                                                 'one surf         Ree      surf
                                                      .           Id           tlE10)             (1E-2)             -            2E-14        2-8        .2E-7                  I tS          Aer!c2 242        v. all ceomds

• 41.3 *1'1 4E-0 - SE-5 5E-4 gone surf r (9El) - 1t-10 -

                                                                                                                                                                              .z I 95           IAmricium-243     W.ill    coeds                   3E-i               *t-3             3E-12                                                   ..5 SE      sr        Se surf (11.0)             (tE-2)                          2-,14         2E-d        2E-7                  I z

95 Alvriclmu-244 2 W. all Compounds- 61.4 4E13 2E-6 St. bell See, surf I (SE.4) (7E'3) l-0 -E . E-2 II Ss Amrkclm-244 W. all topounds 31.3 21.2 Ue- - 4E-5 4E-4 Raw surf (3E+2) - 4E-10 - i 3.. I 95 Auericimr-24S W. all compounds SE44 3E-5 1E-7 4E-4 4E-3 B-47 ODCM Rev. 25

App.B .

                                                                                                                                                                     'AO ' -;      0.
                    .PART 20 STANDARDS FOR PROTECTION AGAINST RADiATION.%                                                                                                ?I, 7abia 1                          Vabi 2           table 3 Occuwatonoal values                         fttluset         Releasm to Concentrations           Sewers.

Col. I Col: 2. Col.3 Col., I Col. 2 Oral rb0thly

                                                          *-      .                 asetlcc               InhalationAvrg Atomic    Radfiouclfdo         Class                             ALL                                          Airv          Water     Concentration "a             .                                               (PC().            2I.

(Pca4 (loCual) (ioCl/4od (sc i/ (1C. lS17 95 Aurc+ 4_o y.c1E*n d- E;2ES E-5 3E-2 St. wall

• 23 (4E 4)
• Ss Aotrituirm246 3E*4 1E*S. 1E-7 4E-4 4E-3 V. all compounds

               .96        cuitus-fi8       2    l. All compounds                 2E*4               11.3           a2-l       2E-9          2E-4        2E-3 36      -cvrlums-?490                                            SEJ1                WI-1 Bont surf (82.1)

Sonssurf (6 1-1 ) 9E-13 If-A lt-S I 96 Curium-241 V. all compoupws 1243 3E41 21-10 *

• 2f5- 21-4 So" surt

                                                                                -                  (141+)                     SE1-I1        -

96 Cur~ium-24i W, all cpounds. 3E11 3E-1 4E-12 Son. surf Sotn sarf (5 2.1) (3 1-1) *E-13 *-7 7E-6 1110. 96 CaaiLiin-243 W. all comounds a 12-10

                                                                                "       surf         "     surf t
     .         96 (2E40)              (ZE-2)                       ZE-14         31-8
                                                                                                                                                     .,M-31E-turliar244          V. all comounds                   1240         .      1E-2            3E-12 CI                                                                              Soors surt          lkam surf-13.-0) -                62-2                      32314          35-9     - 3E-7

'41I. 96 Curlum-24% .. . oll.calwounds to.,a surf lone suit ( E1240). (12-2) 12. 22-14 22-6 2E-7 96 Cutlimw-246 W. all undo 71. 7E 6WS S-3 So.,. surf sonesurl (110) (12-2) 31-12 2E-14 2E-B 2E-7 0 Songe srf Bone utow-96 Curtuir247. V. all c .ounda 2E-I ) (1240 (12-2 ) 6E-3 2E-14 2E2- 2E-7 94 Curlsm?248 W; all counds Dom.4 $Wr 1304r surf O -) (3E-3) 245-13 2- .355 SE-3 96 Cr m23 52.4 22.4 12-13 - . . 72-3, Dom surf

                                                                              -§               (3E14) 95        Curl um-2SO          W. all compound           '   *42-2               13E-4 lone surf           Done' surf
                                                                         *(62-2)              (SE-4)              SE-i3      St-is         9E2-0       *E-9 97        Berkelloov-24S       V. all cowpounds                2243                 12.3                                                 3E-4 97         Sarkellim-    #4I    V. all c*)ouand                                  .3E13               1E-6       42E-9        4E-S         4E-4                   I d

97 Serftelfiar'2ol W. all coands 4E-2 4E-3 . i Bonesurf 11ofe surf (2120) (E-3). 22-12 2E-14 2E-a 2E-7 97 Sark.)la-249 V. all ctownds 2 1' 2 1.0. Bowssot otor surf 15E42) - (42D) E-12 5t-6 SE-$ B-48 ODCM Rev. 25

                                                                                                                                                                    . I,'. .:L -- .

App. B App. 3B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION table l Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Savers

                                                                                  -ol. 2            Col. 2           Col. 3       Col.2         o1. 2 oral                                                                   monthly Inrgestion             lrtdalati0                                      Aver Atomic RdiOnuclide          . Class                             ALS                                              Air         wator   Concentration P)

NCO (CIO U 0p0u) (vt1/^1) OWN)h) (v""18) 9i Br.lnr250 W. allcewouds -SE41 3E12 1E-7 - 11-4 IE-3 gome surf

                                                                                -               I7E.,1              -            ,g-s         -         -

2 98 CalJtorr1Ue y, W44 all cepounds excete those givenfor Y 3E14 -4 42 2E-7 Bt-lo St. wall (3E14) 4E-4 4E-3 Y. oxides and Iiydrexdes 6E*2 2E-7 81-10 41.2

• So90E 4E-3 SE-6 SE-S

                                                                     *         -,,,          .9E4O .                             XE-11 so       Callfsrnluin-248      W. see 24411       ,              8E10                6E-2             3E-ll tons surf Done surf E2E*I)       (tE-l)                        .2E-13          21-7'     2E-6 Y.sa244 ct                   -                  lE l             4£1-U.

N Cslifornisw-249 W. ae " f 5E-1 4E-3 2E-12 .11-14

                                                                                   -W Suf        Bone surf 2                 AI110)                 (9E-3)                                        21-0      2E-7 V. a"      "4Cf,           , _                       1E-2            41-12        21U-14 Sone surf II I                                                                           _                (1E-2)

II XC 98 Calltor,,Ig.230 W. see 244l

• 140 S1-3 4E-12 Son, surf Sor surf U21'0) (21-2) U-li .31-14 3E-8 Jt-7

_ t£3-2 7E-ll 41124 3 Callfornm-f251s W. see Cf, SE-1 4E-3 E-12 - SAonsurf lone surf (lEi0) (3E-3) - .. U14 2£-7 Y. see 244Cf - 11-2 Same surf -2 2EE-t

                                                                                              -(1-2).

• 21-14 98* Californmf-252 W,se2 Cf 21*0 2E-2 goe surf loiw surf (SEI0 (4M-2) 71-S 7E-7

                                          -Y, 'See 24411                     -                   3E-2             1-U          5E-14 36 Californium-253                                         2E.2               21+0              lE-1ll       3E-14 30",     surf 244C                 14142)                -                                                     5E-5
                                                                         -                      21.0              71-10        3E-12 96 Calf forium-254          Va..       C14f..

• 21.0 - 2E-2 SE-U2 3E-14 651-4 3E-7

                                                                             -                  2E-2             7E-U2         2E-14 96   1 staimism-250                                        4E.4                5E.2             2E-7         -                       6E-3 low su~rf W. all compodsas                 _                (11.3)              -            2E-9          11-4 96"       Eisstainiea-251                                        71E3               9E.2              4E-7         _                       1E-3

• Sonweurf W, all Compounds' - (1E.3) _ 2E-9 21-4 E6istlaa.uiwam-2S3 21+2 I10 4E-10 2E-12 2E-5 B-49 ODCM Rev. 25

II App. App. B App. B PART 20 STANJDARDS FOR PROTECTION AGAINST RADiATION. . YA; Table I Table 2 Table 3 Occupatfonsl Values Effluent Releases to Concentratlons Srs Col. I Col. 2. Col. 3 Col. I.. Col. 2 Oral . Iiatoly Ingestion Inhalatibn Average Atoa1c. Radioruclide Class ALI ALI Aer Watr t Concentration li-. -(Pi) CPU,) . (jit1) (YCI/M1) (PCf/xI) (ICI/1) 99 Efnsteins24e W.all ce pounds 3E42- 1E.1 4E-9 l-11l - - LLI wall (3E12) - - - 4E-6 4E-S 99 Einstelniu-254 'w all compounds 8E0O 7E-2 3E-11 - - Bone surf Bone surf (2E1) (IE-I) - 2E-13 2E-7 2E-c 100 Fersiu -252 W.all comouend SE2 1E+1 SE-9, 2E-11 614 6E-S 100 Fermfun-253  ; all cpounds 1E3 1E1I 4E-9 iE-21 1E-S 1E-4 100 fer1u-254 W.all cOePOwedS 3E13 9E+1 4E-1 1E-10 4E-S 4E-4 100 Fervium-255 W; ali compounds. 51E2 21.1 9E-9 3E-11 7E-6 . 7E S 100 Fer1iu-257 W, all cospounds 2E11 2E-1 7E-11 flaSensurf Bone. surf-(411) (2-) - 3E-13 5E-7 5E-6 I S 101 Mmndelevfue-257 W. all tso~mpuds 713 81.1 . E-rn - 1-4, IE-3 I D

                                                           *            -            (9f.1).            -;             lE -10       -           -

101 Nendlesviue-258 W. all cmounds 3E11 2E-1 E-10 Son. surf Bone surf (5E#1) (3E-1) - SE-13 6E-7 6E-6

          -         Any single radionuclide oat IIsted abeve with decay soe ther tban alph mission or so               1s..
                    *leo mndwith radioactive half-life- lss thag 2-hours              .Sgnen.      -              2E42.            IE-7            2E-9
         -          Any sinleradfoenuclide not listad w1tbd y softother than alphx seission or spontaneous f I-lo1e and iUth radioactive half-IIfegraat., than 2 how                 ....      -               2-I             IE-O            lE-12         IE-S        1E-7
                                                                                                                                                                         '.i
                   -   y single radionuclide not listed'.

above that decays by alpha eamssion or spontaneos fission, or any. mx- I i tun for which either the: Identity 4 or the concentration of any radio- 'i nuclide in the.uixturu is not r bo .... - 4E-4 2E-13 1E-1S 2E-9 2E-9 i B-50 ODCM Rev. 25

App. B -PART 20. STANDARDS FOR PROTECTION AGAINST RADIATION App. B FOOTltTTESS 1

                   'Submersion'   mons that values given are forsubnmerson in a hemispherical seai-infinite cloud of airborne materiel 2

lhuS radionuclide$ have radiological half-livs oflecs than 2 hours. The total effective dose equrvalent rnceived during operations with these radionuclides might include a significant contribution from externl expo-sure. The DACvalues for otl radfonuclides. other than those designated Class Suboersion. are based upon the comitted effective dose equivalent due to the intake of the radionuclide into the body and do NOYT include poten-tially significant contributions to dose equivalent from external exposures. The licensee may iistituta IE vC'Iml for the listed DACto account for the submersion dose prospectively, but should use individual monitoring devices or other radiation meaturing instruments that measure external exposure to *monstrate:cooIixance with the lilt$. (See 1 20.1203.) - 3 For soluble mixtures of U-238, J-234..and U1-235ia air, chemical toxicity say be the lisfting ftitor (see 5 20.1201(e)). If the percent by weight (enrichment.) of U-235 Is not greater then 5. the concentration vlue for a 40-hour workweek Is0.2 milligrem uranium per cubic meter of air average. For any enrichment, the product of the average concentration and time of exposure during a 40-hour workweek shall not exceed bE-3 (SA) pCi-hr2al, where SA Is the specifIc activity of the uraniux inhaled. The specific activity for natural uron1_. Is 6.77E-7 curies per grem U. The specific activity for other mixtures of u--238. 1-235, end t-234. it not known, shall be: SA

• 3.6E-7 curins/gra U U-depleted SA * [0.4
• 0.38 enrichment) + 0.0034 (enrichment)'2 E-6, enrichment 3 0.72 where onrichlent is the percentage by weight of U-235, xpressed as percent. ^; '

NOTE: I. lt.the identity of each radionuclide in a mixture Is known but the concentration of ne or more bf the radionuclides in the mixture Is not kn, the OAc for the mixture 'sha l be th most restrictive DC of any radionuclide in the mixtre.

2. If the identity of each radionuclide in the mixture Is not known, but It Is known that certain redionuclides specified in this appendix are not present in the mixture, the Inhalation ALI, SAC. and effient end sewage concentrations for the mixture ore the lowest values specified In this eppendis for any radionuclide that is not -known to be absent fromi the mixture; or C1 a table I V Table 2 lab I 1.

Occupationl Valus,, ' Effluent ' eSoless to I Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. I Col. 2 Oral)otl Ingestion Inhalation Average ALI ALL AC Air. - Water Concentration Redlonuclide (PCi) a(RC) (pci/sl) (RCi/01) (NCi/I) (OCI/Mi) If it Is known that Ac-227-D and Ct-25O-W are not present 7E-4 3i-13. - - If, in addition, it is known that Ac-227-VWY. TW-229-W.Yr Th-230-V, Th-232-W.Y. Pa-231-V.VY Ihp-237-W. Pu-239-W. Pu-240W. Pu-242-Y A_-24i-U. Am-242m-W, A*-243-VW Cm-241-W, Ca-246-Y. C_-247-W' C_24o-1t. I-247-V, Cf-249-W, and Cf-251-V xne not presnnt: - 7E-3 If. In addition, it is known that S-14S-V. *a SW-147-V, Gd-I48,V0i,.Cd-152-0DV, Th-228-.Y. Th-230-Y U-232-Y,. U-233-Y, 6 U-234-Y. u-235-r. U-2361. t-23a-Y, tP-23 -WV Psr-236-V Y *1 Pu-238-WY.' Pv-239-Y, Pu-240-Y, u-242 Y, Por244V,, Co-243-Y Cm244-V, Cf-24e-v Cf-249-Y. Cf-250-V.Y. Cf-21-Y tt-252-V.Y and Cf-254-VY are not present 7E-2 il-li - - - If, in addition, it Is known that Pb-20-0. Oi-210miI, Fo-210-0,v, Ra-223-. R2a-25-1-, Ma-226-i. -Ac-225-DWV,i, Th-227-V.Y. U-230-0.VY. U-232-D.CV Pu-241-W. Ca-240-W. C_-Z42-W* Cf-248-i Er-254-V. fo-257-Y,.and d-25O-V are not present 71-1 31-10 - B3-51 ODCM Rev. 25

I f I App. B PS 4 F.O PRTCTO PART 20 STANDARDS'FOR PROTECTION.AGAINST RADIATION

                                                                                 *:.*ATI         AGANS            R                                Apji~.B.

Table 1 , . Table 2 Table 3 Occupational Values Effluent Release$ to. Concentrations Sewers Col.1 Col.z . Col. 3 Col. 1 Col. 2 Oral monthly Ingestion Inhalatfon Average ALJ A O Air Water Concentration Radionuclide (PCI) (pC1) (st/mi) (pCi/el) (pCi/al) (pCi/mi) If. in addition~ It is known that Si-32-Y. Ui44-Y. Fe-60-0. Sr-gO-Y; Zr-93-D. Cd-113*-D. Cd-113-D, ln'11SOW. ta'138-Di Ra-224-W. Ra-2211-W. Ac226-DW.Y. ?a-230-W.Y. U-233-O.V. U-234-0,1, U-235-O.W. U-236-O.W. U-238-O.V. Pu-241-Y. Sk.249-W. Cf-2Ss-W.Y; and Es-2S3-W are n~t plQ5e~t ?E-0 - 3E-3 It It is known that Ac-227-D.W.Y. Th-229-W.Y. Th-232-W.1. Ps-231-V.Y. Cm-243-W, and Ca-250-W are not present IE-14

• It. in addition, it is known that Sm-146-W, Gd-14"-.W. Gd-152-D. Th-228-Y.Y. Th-230-W.Y.

11-238-Y. U-Nat-?. Np-236-V, Np237-V. Pu-236-W.Y, Pu-23"-.Y. Pu-229-W.Y. Pu-240-W.Y. Pu-242-V.Y. Pu-244-W.Y. Am-241-11, PAm-2rW.,Am-243-W.. Cm-243-V,-Cm-244-W; Cm-24S-W, Cm-246-W.- Cm-247-V. gk-247-W. Cf-249-W;Y. Cf-2WO-,Y. Cf251-VW.Y. CV-252V.Y. and Cf-254.-VY are not present. id-li - If. In addition. it is known that Sm-147-4. D Gd-ISZ-W. Pb-210-D, 31-210-V. Po-210-D.W. Ra-223-W. la-22S-V. Rta-226-V. Ac-225-OVW.YT Th-227?W.Y, U-230-O.W.Y. U-232-D.V. U-Nat-W. PUr2421V. Cir240-V. Dre242W. Cf-24"-.Y. Es-Z54-W. F.-2$7-W. end 9*-2WW- amt not present - IE-12 - If.-to additiont It Is known that Fe-BC. Sr-90. Cd-113.'Cd113. lnIrIS, 1-125. Cs-134. So-145. Sm-147..CGd148; Gd-ISZ., Ng.194 (organic). ISi-ZlOE. ka'Z223. Ra-224. Ra-22S. Ac-Z25. Th-US. Th-230. U-233. 11-234.

3-235. U1-236. U-238~ U-Nat. Car242. Cr-248, Es-254. Fo-257. and Md-258 aea nott present- IE-6 IE-5

3. If a mixture of radionuclidet consists of urani0 and Its daughters ln ore dust (10 p AImAD particle distribution assued) prior to chemical separation of the uraniu frox7 the ore, the following values may be used for the DAC of the mixture: SE-1l pCI of gross alpha activity fros uraniun-230. urani10-234 thoriumx230.

and radfu-226 per milliliter of air; 3E-2 pCi. of natural urani1u per milliliter of aIr. or 45 micrograms ot naturol uraniuo per cuhic meter of air.

4. ir th identity and concentration ofe'ach radionuclide In a mixture art know, the l11iting.valuos should be derived as follows: determine, for each radionuclide in the miature. the ratio between the concentration present In the mixture and the concentration otherwise established in Appendix _ for Ive specific rmdionuclide when not in a mixture. The sum of, suck ratios for all of the radionuclides to the mixture may not exceed 1I (i.e.. 'unity).

dxapIt: If radionuclides 'A. o. and C *ar present in concentrations CA. C. and C.. and if the applicable DACsare DA 4 DACB,and DAdC. respectively. then the concentrations shall be limited so that the following relationship exists: CA C1 CC 1 MA - aq B-52 ODCM Rev. 25 J_

APPENDIX C EMS SOFTWARE DOCUMENTATION C ODCM Rev. 22

I I APPENDIX C EMS SOFTWARE DOCUMENTATION TABLE OF CONTENTS CONTENTS PAGES Effluent Management System Software Test Report for C-3 Seabrook Station, May 1994 Cover ii 1-11 Resolutions of EMS Software Test Report Discrepancies C-4 1-2 Software Requirements Specification for North Atlantic C-S Energy Service Corporation, Seabrook Station, 1-35 Effluent Management Systems, Revision 04, FP 75486 Technical Reference Manual, Effluent Management System C-6 NAESCO Seabrook Station, July 1994, FP 75486 36-93 0 C-2 ODCM Rev. 22

APPENDIX C: EMS SOFTWARE DOCUMENTATION ATTACHMENT 1: EFFLUENT MANAGEMENT SYSTEM SOFTWARE TEST REPORT FOR SEABROOK STATION, MAY 1994 C-3 ODCM Rev. 16

APPENDIX C., EMS SOFTWARE DOCUMENTATION ATTACHMENT 2: RESOLUTIONS OF EMS SOFTWARE TEST REPORT DISCREPANCIES C-4 ODCM Rev. 16

I Attachment 2

2. Resolution of EMS Software Test Report Discrepancies The following discrepancy resolutions apply to the findings contained in the "Effluent Management System Test Report for Seabrook Station, May 1994" as noted on pages 9 and 10 (see Attachment #1 of Appendix C of the ODCM). With the positive resolution of the discrepancies identified'in the EMS dose-code, use of EMS as a computerized alternative-approach (designated as Method IA in the ODCM) to determine compliance with the radioactive'effluent dose and dose rate limits is acceptable since the results are comparable with'-the currently approved dose methods.

Discrepancy: Mixing ratio for shoreline activity used in EMS Program not equal to the value used in the ODCM Method I (Mp - 1.0). Resolution: The mixing ratio for the shoreline activity pathway in the EMS is consistent with the ODCM Method II approved value of 0.025, and therefore does provide for a, calculated dose that is within the parameters already approved in the ODCM. The'use of the EMS code (ODCM Method'IA) for calculating liquid doses is acceptable for , determining compliance with the dose limits -of the Technical Specifications without the need to modify the assumption used for the shoreline mixing ratio. Discrepancv: EMS dose factors based on Fp (fraction of year animals are on pasture) value which is not consistent with ODCM. l Resolution: ODCM Method I assumes that the pasture season in the North East is 6 months long each year (Fp +'0.5). Method II allows for the pasture fraction to be set equal 'to 0.0 for the first and fourth quarters which equates'the'non-growing period of the year. "The second and third quarters'corres'pond to the growing season where the pasture fraction is'assumed to be'1.0.` Thei'EMS software assumes an Fp value,of 1.0 for animal grazing (meat aind milk pathways) for-all conditions. This is a - moderately conservative approach compared to Method I and the off grazing season conditions modeled in Method II. It is equal to the grazing season assumptions of Method II as applied in the second and third quarters. As a result, the added conservatism in the EMS calculations for doses via milk and meat pathways are within acceptable margins and guidance provided in NRC NUREG-0133 for demonstrating compliance with Technical Specification'dosezlimits. No changes to the EMS software are necessary. Discrelancv: l Shielding factors (SF) applied uniformly to dose rates and doses -in the EMS program. 1

Attachment 2 l 2. Resolution of EMS Software Test Report Discrepancies (Continued) Resolution: The EMS program for gaseous releases incorporates a shielding factor (SF) equal to 1.0 for both- dose rate and total dose determinations. In contrast, both Method I and II use a SF value of 1.0 instantaneous dose rate calculations, but a value of 0.7 for integrated doses based-on assumptions in NRC Reg. Guide 1.109. The use of a SF equal to 1.0 for the external ground plane exposure pathway. for both dose rate and total dose is a moderately conservative assumption that is within the bounds already assumed in the ODCM dose modeling. As a result, no modification to the EMS code as an acceptable approach (Method IA) for demonstrating compliance with Technical Specification dose/dose rate limits is required for SF: I Discreuancy: Incorrect receptor location identified on EMS printout for ground level release lpoint. I Resolution: Incorrect name is identified on report with no impact on dose or dose rate calculations which were verified to be correct. Discrepancy: Assumed fraction of elemental iodine used in EMS program differs from ODCH Methods I and II. Resolution: For ODCM Methods I and II, the fraction of elemental iodine assumed for gaseous releases-in 0.5 based on the guidance in NRC Reg. Guide 1.109. The EMS code assumes an elemental iodine fraction of 1.0-based on the guidance in NUREG-0133. Consequently, the EMS program (Method IA) will produce a moderately conservative estimate of dose impact (factor of 2); for iodine radionuclides if present in the release estimations when compared to existing approved methods. As a result, no modification to the EMS code is necessary for use in the ODCM for determining compliance with Technical Specification dose limits. I Discrepancy: Missing organ dose rate information on EMS printout for effluent discharges containing I-131, I-133, H-3, and particulates. l Resolution: This required information is easily obtainable from the permit closure process with flashing indication if any dose or dose rate limits are exceeded. 2

APPENDIX C: EMS SOFTWARE DOCUMENTATION ATTACHMENT 3: SOFTWARE REQUIREMENTS SPECIFICATION FOR NORTH ATLANTIC ENERGY SERVICE CORPORATION, SEABROOK STATION, EFFLUENT MANAGEMENT SYSTEMS, REVISION 04, FP 75486 C-5 ODCM Rev. 16

EFFLUENT MANAGEMENT SYSTEM: SOFTWARE TEST REPORT FOR SEABROOK STATION MAY 1994 Prepared by -5/f Late to/4 -tL . /' Date Reviewed by . A+.Jf Dat Approved by / I/ Date Yankee Atomic Electric Company Nuclear Services Division 580 Main Street Bolton, Massachusetts 01740

Table of Contents

1.0 INTRODUCTION

. . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.1 Background . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.2 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . 1 2.0 

SUMMARY

OF FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . 2 2.1 ENS Dose and Dose Rate Conversion Factors . . . . . . . . . . . 2 2.2 Liquid Release Testing . . . . . . . . ... . . . . . . . . . . 4 2.3 Gaseous Release Testing . . . . . .. . . . . . . . . . . . . . 4 3.0 TEST CONCLUSIONS . . . . . . . . . . . . . . . . . . . . . . . . . . 7 4.0

SUMMARY

OF DISCREPANCIES . . . . . . . . . . . . . . . . . . . . . . 9 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 i i* 1

1.0 INTRODUCTION

Software testing as described in Reference [1] has been conducted for the Seabrook Station version of the Canberra Effluent Management System (EMS). The results and conclusions -are presented in this report.

1.1 Background

Canberra Industries Inc. developed the EMS software to assist nuclear power plant personnel track effluent emissions and perform associated dose calculations.' North Atlantic Energy Service Corporation purchased a Seabrook-specific version the Canberra EMS software which must meet specific requirements and incorporate site-specific information -'provided in the Offsite Dose Calculation Manual (ODCM) [2]. Software testing was conducted to provide assurances that the Seabrook EMS program produces results which are consistent with current ODCH assumptions and iethods. All executions of the EMS program were performed at Seabrook Station on the target software. All executions of ODCM Method II were conducted at Yankee Atomic Electric Company in Bolton, Massachusetts. - 1.2 Acceptance Criteria The operability of the EMS software will be accepted if (i) information contained in the EMS data files is consistent with'the ODCM, (ii) test results from the EMS program are consistent'with -results from ODCM methods, (iii) Technical Specifications requirements are met by the'EMS software, and (iv) the EMS software meets design specifications.' Final user (Seabrook) 'acceptance- is contingent on-Seabrook approval of verification testing results and'criteria'established by user needs. 1

I I 2.0

SUMMARY

OF OBSERVATIONS The EMS software testing included (i) identifying appropriate meteorological set up data, (ii) review of dose and dose rate conversion factor development, (iii) assessments for liquid releases, and (iv) assessments for gaseous releases. ODCM Method I was used initially to confirm dose results from the EMS program.. However, the simplified nature of ODCM Method I made it difficult to change the values of various parameters or obtain meaningful comparisons (other than .bottom line" comparisons). The more adaptable ODCH method, Method II, was then used to confirm EMS doses. Observations made during the software testing are summarized below. 2.1 EMS Dose and. Dose Rate Conversion Factors The EMS software uses precalculated conversion factors which are contained in a data file. The dose conversion factors for both liquid and gaseous effluent releases were developed for-four age groups (adult, teen, child and infant), and for specific organs (bone, liver,total body, kidney, lung, GI tract and skin). The liquid release dose conversion factors in the EMS program are the summation of the components for water recreation and ingestion of aquatic foods. The gaseous release dose conversion factors are exposure pathway-specific (e.g., inhalation, ground plane, milk ingestion, etc.). Dose conversion factors are provided in the EMS program for all exposure pathways addressed in the ODCH. The development of all dose conversion factors in the EMS program followed the pathway-specific equations in the Effluent Management System Technical Reference Manual [3]. The EMS conversion factors for several radionuclides were-examined to determined that the development process was consistent to the Technical Reference Manual and the ODCM. 2

2.1.1 Liquid Release Dose Conversion Factors-Although the individual components for the ingestion of aquatic foods were found'to be consistent with the ODCM,. a discrepancy was discovered in the water recreation component. The mixing ratio. for shoreline activity. used in .the development of the EMS dose factors is 'equal to 0.025. While this value is inconsistent with ODCM Method I..(which.employs a mixing ratio of 0.1), it is consistent with ODCM Method II.- It' is identified as a discrepancy because it is unclear which set of ODCH assumptions (those for Method I or those for Method II) the EMS program is expected to adopt.- 2.1.2 Gaseous Release Dose Conversion Factors The EMS program uses dose conversion factors from Regulatory Guide 1.109 for assessment of noble gas releases. The dose factors in the EMS program were verified against and found to be consistent with Table B-1 of Regulatory Guide 1.109 [4]. The development methods for the other gaseous'dose factors (i.e., for inhalation, ground plane, milk -ingestion, meat ingestion, and ingestion of vegetables) were reviewed against applicable equations in the Technical Reference. Manual and information in the ODCM. It is noted that the dose factors for ingestion of milk and meat are based on the fraction of year that animals are allowed to graze on pasture land (Fp) equal to 1.0. This is not consistent with the ODCM which calls for the use of an Fp value equal to 0.5.-

     -The dose conversion    factors, in the EMS -.program for - gaseous releases incorporate a shielding factor- (SF) equal to 1.0. ;The EMS program is designed with a way of changing the value of SF (via use of the Options Table), but the factor is applied uniformly to both doses and dose rates.       In contrast, the ODCM calls for the use of different values for SF in the calculations for doses and 3

I I dose rates. 2.2 Liquid Release Testing Dose estimates from the EMS program for hypothetical liquid effluent discharges (containing single nuclide- and radionuclide mixtures) are nearly. identical to results from ODCM Method II when input data are based on the same mixing ratio value, indicating that. the calculation method used in the EMS program is consistent with- the: ODCM. Additionally, the EMS routine(s) responsible for liquid effluent concentrations comparisons to MPC values and monitor set point determinations was observed to be operating properly. 2.3 Gaseous Release Testing The agreement between estimates for total body dose rates, skin dose rates, and air (gamma and beta) doses due to emission of noble gases from the ODCM methods and the EMS program is-excellent, indicating that the EMS calculation method is consistent with the ODCH. There is also excellent agreement between inhalation doses from the EMS program 'and ODCM Method II indicating that, for the inhalation pathway, the calculational' method and assumptions in the EMS program are consistent with those in 'the ODCM. The evaluation of the dose estimates via inhalation pathway included both long and short release durations for an elevated (mixed mode) and a ground level release point. The excellent agreement between the EMS and ODCK Method II also confirms that the release duration adjustment term, t-a, is applied properly in the EMS program. However, an incorrect' receptor location was reported' on'the EMS printout in the tests (D-2c and D-2d) in which the Plant Vent was changed to be' recognized as a ground level release point. Also noted' during testing was that the EMS routine(s) responsible for calculating effluent concentration-to-MPC ratios and radionuclide release rates 4

appears to be operating properly for gaseous releases. The EMS program incorporates the assumption that the fraction of elemental iodine is equal to 1.0 (consistent with NUREG-0133 [5]). In contrast, the fraction of elemental iodine is assumed equal to 0.5 in the ODCM methods (consistent with Regulatory Guide 1.109). Consequently, the EMS program produces dose estimates due to radioiodine that are at least a factor of two greater than doses from the ODCH methods. This difference increases to about a factor of 4 when the current values for Fp and SF assumed in the EMS program and ODCM methods are used in the dose calculations. The different assumptions for elemental iodine fractions should not present a problem because each program is based on NRC guidance: the EMS is based on NUREG-0133, the ODCM methods are based on Regulatory Guide 1.109. The EMS program takes the more conservative approach for determining doses from radioiodine. Making appropriate adjustments for Fp, SF, and the fraction of elemental iodine (when radioiodine input was used) and comparing results for organ doses due to 1131, H3, Co6O and Csl37 revealed that the calculational methods used in the EMS program are consistent with the ODCM for all exposure pathways (i.e., ground plane, inhalation, milk ingestion, meat ingestion, and vegetables ingestion). Technical Specification 3.11.2.1 and the ODCM require the calculation of organ dose rates due to effluent discharges of 1131, 1133, H3 and particulates with a half-life greater than 8 days. However, in all test cases involving these types of nuclides, organ dose rate information did not appear on Page 4 of the EMS printout. Instead, the message 'No calculations performed - check Sample & Receptors" appeared. The EMS set up data and input were reviewed with no apparent error identified. Since the test cases included Csl37, Co6O, I131, and 5

I I H3, the missing dose rate information was.uriexpected. It is noted that organ. dose rate information' was provided on Page 4 of the EMS printout during a demonstration of the EMS program prior to testing. 6

3.0 TEST CONCLUSIONS Although the dose conversion factors are based on information which is not completely consistent with the assumptions in the ODCM, the calculational methods used to determine doses from liquid and gaseous effluent discharges are consistent with the ODCH methods.' - Other conclusions are:

1. As stated in Section 2.1.1, the development of the EMS liquid effluent dose factors is consistent with ODCM Method II, but not with Method I due to the mixing ratio value. If the EMS program is intended to be a hybrid method, the dose factors are consistent with the ODCM and are acceptable.

On the other hand, if the EMS program is intended to provide automated ODCM Method I calculations, then the dose factor should be recalculated using a mixing ratio for shoreline activity equal to 0.1.

2. Since the EMS program is not designed to support the use of two shielding factors (one for dose rates and one for doses), use of a shielding factor equal to 1.0 is acceptable with the understanding that, although the dose rates produced by the EMS program will be consistent with the ODCM, the doses from the EMS program will be based on a more conservative assumption than doses from the ODCM methods.

3. Under the normal ODCM assumption for elemental iodine, the results from the EMS program will be at least a factor of two greater than results from the ODCM methods. The different assumptions regarding the elemental iodine fraction do not present a problem because each program is based on NRC guidance: the EMS program is based on NUREG-0133, and the ODCM is based on Regulatory Guide 1.109. Of the two methods, the EMS program takes the more conservative approach toward estimating doses from 7

I I radioiodine in gaseous effluent.

4. The radiation monitor set point determination method for liquid releases produces a set point value that is consistent the ODCH set point method.

5. The EMS routine that is responsible for comparison of liquid effluent concentrations and MPC values is operating properly.

6. The release duration adjustment term, ta, is used consistently to the ODCM.

8

4.0 SUKHARY OF DISCREPANCIES Discrepancy Area of Impact, Potential Solution(s) Mixing ratio for Doses associated with Clarify whether the EMS shoreline activity liquid effluent program is expectedto used in EMS discharges. follow ODCH assumptions,,, program. for Method I or MethodII.. If determined to follow Method I, recalculate 'dose factors for liquid releases. EMS dose factors Doses due to ingestion of Recalculate EMS dose based on Fp value milk and meat. factors for milk and meat which is not ingestion pathways to consistent with incorporate Fp value ODCM. consistent with the ODCM. Accept added conservatism in EMS in calculations of doses via milk and meat ingestion pathways. Shielding factor Doses associated with Accept use of SF - 1.0 and (SF) applied gaseous effluent the added conservatism for uniformly to dose discharges. doses. rates and doses in EMS program. Modify EMS software to accommodate use of two values for SF (one for dose rates and one for doses). Incorrect receptor Potential assignment of Discuss with Canberra. location doses to the wrong identified on EMS receptor. printout for ground level release point. Assumed fraction Dose estimates due to Accept added conservatism of elemental iodine in gaseous in doses due to iodine. iodine used in EMS effluents. program differs Modify EMS software to use from ODCH methods. fraction for elemental iodine that is consistent with ODCM. 9

I I Discrepancy Area of Impact Potential Solution(s) Missing organ'dose, Technical Specification Discuss with Canberra. rate information- required dose rate not on EMS printout' calculated. for effluent discharges containing 1131, I133, H3, and particulates'. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ 10

References.

1. Yankee Atomic Electric Company, Effluent Monitoring System Software Test Plan for Seabrook Station, May, 1994.

2. NAESC, Station Offsite Dose Calculation Manual, Rev 13, 9/24/93.

3. Southern Nuclear Operating Comranv Effluent Management System Technical Reference Manual (07-0545), January 1993.

4. NRC Regulatoiy Guide 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purposes of Evaluating Compliance with IOCFR Part 50. Appendix I, Revision I, October 1977.

5. NRC NUREG-0133, Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants, October 1978.

11

Software Requirements Specification for North Atlantic Energy Services Corporation Seabrook Station Effluent Management Systems 48-8448 Revision 04 Nuclear Data Systems Division Software Product Originator: Date: De I Approved: St Y'D A", lX Engineering (CNDS) Approved: Date: Vg Approved: Date: 9______f Proj ger (abrok Station) 7c \ .9a.

II Software Requirements Specification RS-8448-04 Revision HistotY Initials Revision Date Description DJH 00 2/26193 initial version DJH 01 3/22/93 Updated incorrect dose equation DJH 02 4/30/93 Updated to Include all dose and dose rate equations DJH 03 8/3/93 Updated based on modifications to software and customer's requested modification to the use of the default nuclide for gaseous permit processing. DJH 04 9/14193 Updated based on customer's request to remove modification to the default nuclide for gaseous permit processing. i 7aze,

Software Requirements Specification RS 448-04 Page

1. Scope 1

2. Applicable Documents

3. Interfaces 1 3.1 Hardware 1 3.2 Software 1 3.3 Human 2

  -  3.4  Packaging                                  2

4. Definitions 2

5. Principal Changes from Existing Packages 2

6. EMS Functionality , --, i. I --, . . 4

   --6.1  Database Maintenance Transactions          4 6.2 Editing Values through INGRES OBF           7 6.3 -Uquld Pre-Release Processing               8 6.3.1 User Interface and Functionality     8 6.3.2 Associated Reports                   9 6.3.3. Underlying Calculations             10 6.4 Liquid Post-Release Processing              11 6.4.1., User Interface and Functionality   11 6.4.2 'Associated Reports                  12 6.4.3   Underlying Calculations            12 6.5  Uquid Permit Editing                       13 6.5.1 User Interface and Functionality     13 6.5.2 Associated Reports.                  13 6.5.3 Underlying Calculations              13 6.6  Liquid Permit Deletion                     13 6.7  Gaseous Pre-Release Processing             14 6.7.1 User Interface and Functionality     14 6.7.2 Associated Reports                   15 6.7.3 Underlying Calculations              15 6.8  Gaseous Post-Release Processing            21 6.8.1 User Interface and Functionality     21 6.8.2 Associated Reports                   21 6.8.3.. Underlying Calculations.          "22 6.9  Gaseous Permit Editing                     27 6.9.1 User Interface and Functionality     27 6.9.2 Associated Reports                   27 6.9.3 Underlying Calculations              27 6.10 Gaseous Permit Deletion                    27 6.11 Seml-Annual Reporting                      28 6.11.1 User Interface and Finctohality     28 6.11.2 EMS Trend Plots                     30 6.12  End-of-the-Year Data Archiving            30 6.12.1 User interface and Functionality    30

_ti3.

I I Software Requirements Specification RS-8448-04

1. Scope This document establishes the software requirements for the Effluent Management System (EMS) software to be Installed at North Atlantic Energy Services Corporation's Seabrook Station.

2. Applicable Documents 2.1 The following two documents are included as part of this SRS, and this SRS refers to specific sections of them:

2.1.1 Southern Nuclear Operating Company Effluent Management System Operator's Manual' (07-0544), Version 1, January 1993.' 2.1.2 'Southern Nuclear Operating Company Effluent Management System Technical Reference Manual" (07-0545), Version 2, 'January 1993. Note: The above documents contain material (including screens and report formats) imported from final manuals for other EMS packages. Utility and plant names shown on screens and reports in these' manuals are not significant, since they are determined by database data that will be custormized to fit the Seabrook Station's usage. 2.2 The following document is a reference source for calculation methods of the EMS software. This SRS may refer to specific sections. 2.2.1 wSeabrook Station OffsIte Dose Calculation Manual,' Revision 12, January 1993.

3. Interfaces 3.1 Hardware The EMS software shall run on the following CPU model: DEC Microvax 3100, Model 80.

3.2 Software The software shall be written under VMS version 5.4-2 or later, using INGRES version 6.4 or later. It shall be written in VAXIFORTRAN or VAX-DCL Utility programs provided by INGRES that are installed on the hardware configuration may be used if applicable. Software Requirements Specification RS4W48 04 3.3 Human The user may be expected to have received operator training from the system manager, CanberraINDS, or the plant training department prior to using any part of the EMS software. Knowledge of INGRES or VMS shall not be assumed. The menus of operations are intended to be self-explanatory, but an Operator's Manual shall be developed. The user may be expected to have'enough knowledge of USNRC-regulated nuclear power plant effluent management to provide accurate and appropriate inputs, and to determine the validity of the'software's results. 3.4 Packaging - A distribution kit will be produced for the customer. Any removable medium supported by the operating hardware delivered to the Seabrook Station Is an acceptable distribution medium.

4. Definitions EMS - Effluent Management System. Software for determining effluent monitor setpoints, tracking activity releases and dose Impacts of Individual releases, and generating semi-annual release reports.

SRS - Software Requirements Specification. SNC - Southern Nuclear Operating Company

5. Principal Changes from Existing Package The following paragraphs summarize the principal changes to the existing software that are required for the Seabrook Station system, and are Intended only as Introductory material.

Specifics of the required Seabrook Station EMS functionality are presented in the following sections. 5.1 The EMS software will be developed by customizing the generic EMS package. In general, the most important changes from previous versions are as follows: 5.1.1 Modification to Gaseous Permit Processing to allow scaling of nuclides for Plant Vent Spike release point 5.1.2 Modification of noble gas dose rate and dose calculation methods to use a third set of XIO values. I I Software Requirements Specification RS8448-04 5.1.3 Modification of noble gas dose rate and dose calculation methods to multiply X/Q and DIQ values by a factor depending on the release duration. 5.1.4 Modification to setpoint calculations to calculate setpoints for low gamma concentration releases. 5.1.5 Modification of Permit Processing to automatically correct the expected waste flow if it is greater than the calculated maximum waste flow. 5.1.6 Modification of Uquid Permit Processing to determine dilution flow rate based on the number of pumps operating. 5.1.7 Modification of the permit reports to Include Month-to-Date Cumulative Doses and Alert Setpoints. 5.1.8 Modification of Post-Release Permit Processing to update the monitor response. 5.1.9 Addition of data to database to support and control the above operations. Software Requirements Specification RS84804

6. EMS Functionality 6.1 Database Maintenance Transactions The functionality of the EMS Database Maintenance transactions shall be described in section 2 of the EMS Operator's Manual (Reference 2.1.1), with the following revisions:

6.1.1 On the Release Point Setpoint transaction [EM-DM-RP (Form 2)], and the Discharge Point Setpoint-transaction [EM-DM-DP (Form 2)], the following parameter shall be added to the list of those which can be entered, stored, and which appear on the printed report for these transactions:

• SCALNUC: For a gaseous release, a flag to denote that this release point will have nuciide concentrations scaled so that the total concentration matches a value entered by the user.

6.1.2 On the Release Point Setpoint transaction [EM-DM-RP (Form 2)] and the Discharge Point Setpolnt transaction [EM-DM-DP (Form 2)], the following parameter shall be added to the list of those which can be entered, stored, and which appear on the printed report for these transactions:

• DILOOKUP: For a liquid release, a flag to denote that permits for this release point will have a selection screen appear for the user to select the proper dilution flow for the release based on the number of pumps operating.

6.1.3 On the Release'Point Setpoint transaction [EM-DM-RP (Form 2)], and the Discharge Point Setpoint transaction [EM-DM-DP. (Form 2)], the following parameter shall be added to the list of those which can be entered, stored, and which appear on the printed report for these transactions:

• DEF..NUC: For a liquid or gaseous release, this parameter will contain the default nuclide that will be used In setpoint calculations for low gamma concentration releases. This parameter is used in conjunction with the DEF _CONC parameter.

6.1.4 On the Release Point Setpolnt transaction [EM-DM-RP (Form 2)], and the Discharge Point Setpolnt transaction [EM-DM-DP (Form 2)], the following parameter shall be added to the list of those which can be entered, stored, and which appear on the printed report for these transactions: a DEF CONC: For a liquid or gaseous release, this parameter will contain the default concentration that will be used in setpoint calculations for low gamma concentration releases. This parameter Is"used in conjunction with the DEFNUC parameter. 7DAbD,

1L Software Requirements Specification RS-8448-04 6.1.5 On the Release Point Setpoint transaction [EM-DM-RP (Form 2)], and the Discharge Point Setpoint transaction [EM-DM-DP (Form 2)],'the following parameter shall be added to the list of those which can be entered, stored, and which appear on the printed report for these transactions:' DEF_TYPE: For a. liquid or gaseous release, this parameter will contain the default nuclide type that will be used in setpoint calculations for low gamma concentratibn releases. This parameter is used In conjunction with the DEFNUC and DEFCONC parameters. (Note: For a gaseous release, the default nuclide type shall determine which monitor setpoint should use the default nuclide and concentration.) 6.1.6 On the Release Point Setpoint transaction [EM-DM-RP (Form 2)], and the Discharge Point Setpoint transaction [EM-DM-DP'(Form 2)], the following parameter shall be added to the list of those which can be entered, stored, and which appear on the printed report for these transactions:

• ALRT_SET: For a liquid or gaseous release, this parameter will contain the multiplier to be used In the calculation of Alert Alarm Setpoints for permit reports.

6.1.7 On the Release Point transaction [EM-DM-RP (Form 1)], the meaning of the Response Option will change. When set to AY, this option will denote the display of a Monitor Response window during the Post-Release Permit Processing, rather than during the Pre-Release Permit Processing. The Response Option parameter, Itself, will remain unchanged for this transaction, but the response entered should Include the monitor background values. 6.1.8 On the Dilution Streams transaction [EM-DM-DSJ, the following parameters will be removed: the number of extra dilution flow rates and the four dilution flow rates. These parameters will be replaced with two column fields. One column will contain the dilution flow rate, while another will contain the pump configuration description (such as Jockey Pumpoor 5'). In this transaction, the dilution flow rate for particular pump configuration 'can be added. 6.1.9 On the Meteorological Data transaction [EM-DM-ME (Form 1)], several menu options will added to the fist of MET DATA TABLES. These additional menu items are as follows: X10 - Noble Gases (Gamma)

                         'a' Factor-D/Q-Paritlodines waV Factor - Noble Gases a Factor- X/Q-Pa~rllodines a' Factor - Gamma Noble Gases
                                                             '                 itS0

Software Requirements Specification RS48448-04 6.1.1 0 On the Meteorological Data transaction [EM-DM-ME (Form 1)], the following menu items will be used to store short-term (1 hour) DIQ and X/Q values. D/O - Partics/Radioiodines X/IO- Partics/Radioiodines X/Q - Decayed Noble Gases XIQ - Noble Gases (Gamma) Note: This specification item only denotes a change in the meaning for the values on this transaction and requires no further changes to the software. 6.1.11 On the Meteorological Data transaction [EM-DM-ME (Form 1)], the X/Q, D/I, and Ca Factor values are defined for various elevations, distances, and directions from the plant vent or stack. This combination with the mode of release" parameter on the Release Point transaction [EM-DM-RP (Form 1)], and the receptor definition on the Gas Receptors transaction [EM-DM-GR], allow the X/Q, D/O, and jaw factors to be different for each receptor and/or release point. Note: This specification item is only for clarification and no additional code changes need to be made to this transaction.

                                               .6-

II Software Requirements Specification RS-8448-04 6.2 Editing Values through INGRES QBF In addition to the interactive forms-based EMS Database Maintenance transactions, certain flags and values must be edited through INGRES QBF on the database tables which contain data not accessible through the forms-based transactions. 6.2.1 Some columns of the Quarterly Dilution Volume table (QDVOL), which has no other use In the Seabrook Station version of EMS, will be used for recording monthly dilution volume for use In semi-annual reports. Once per month, an authorized user will use QBF to append a record to the QDVOL table as follows: sampleid (sample ID) 0 [not used] dvdate (dilution volume date) The first day of the month to which the volume applies (time not required). tvol (total volume) Dilution volume for the month, In user units. aflow (average fowrate) 0 [not used] f0Q_ \°

Software Requirements Specification RS-8448-04 6.3 Liquid Pre-Release Processing 6.3.1 User Interface and Functionality Uiquid Pre-Release Processing functionality for the EMS software shall be as described in section 3 of the 'EMS Operator's Manual (Reference 2.1.1), with the following revisions: 6.3.1.1 On the Liquid Perrmit Definition Screen (Screen 3.04): Upon entering the permit definition screen, If the DILOOKUP parameter is set to "Yi for the release point associated with the current permit being processed, the Dilution Flow Rate parameter will default to zero.

                     ,lf a user uses the "Tab or wReturn' key to exit the Dilution Flow Rate parameter on the Permit Definition Screen and the Dilution Flow Rate parameter has'a value of zero, a selection screen with two columns of data will appear. One column will contain the pump configuration description, while the other will contain the dilution flow rate for each associated pump configuration.

Upon selection of the Dilution Flow Rate, the selection screen will disappear and the selected dilution flow rate will appear in the Dilution Flow Rate parameter on the Permit Definition Screen; The cursor will then automatically advance to the Dilution Volume Parameter. 6.3.1.2 On the iUquid Permit Definition Screen (Screen 3.04): When a "Fill" (F14) or a "Saveg (F1 0) without a "FIJI" Is executed, if the DILOOKUP parameter Is set to "Y' for the release point associated with the current permit being processed and the Dilution Flow Rate parameter is set to zero,' a'selection screen, as described above will appear. Once a selection of the Dilution Flow Rate- Is complete, the selection screen will disappear and the 'Fill" operation will continue. Upon completion, the selected dilution flow rate will appear In the Dilution Flow Rate parameter on the Permit Definition Screen. If the Dilution Flow Rate parameter on the Permit Definition Screen Is not set to zero and the-DILOOKUP parameter Is set to "Y', the fill will proceed as normal without the dilution flow rate selection screen appearing.

                               ;B

II Software Requirements Specification RS-8448-04 6.3.1.3 Priorto entenng the Liquid Permit Approval Screen (Screen 3.09): If it is determined that the computed maximum waste flow is less than the anticipated waste flow, the anticipated waste flow will be changed to have the value of the computed maximum waste flow. If the anticipated waste flow is modified, setpoint, dose, and dose rate values will be recalculated based on the new value. 6.3.1.4 For releases with low or zero gamma emitter concentrations that result In a pre-diluted MPC ratio less than 10%, a default concentration will be used for setpoint calculations. This default concentration will not be used for updating curie, dose rates, or dose totals. The default nuclide will be attained from the DEFNUC parameter. The default concentration for this nuclide will be attained from the DEF_CONC parameter. The default type for this nucilde should be attained from the DEF._TYPE parameter. 6.3.1.5 The Monitor Response Screens for Release Points and Discharge Points (Screen 3.08) will no longer appear while processing a Pre-Release Permit when the Response Option Is set to "Y" on the Release Point transaction [EM-DM-RP (Form 1)]. 6.3.2 Associated Reports Liquid Pre-Release Permit Reports shall be as described In section 3 (pages 3-53 through 3-58) of the EMS Operators Manual (Reference 2.1.1), with the following revisions: 6.3.2.1 On the Pre-Release Permit Report (3.01), the Cumulative Month-to-Date Doses will appear on the page with the report category of Cumulative Maximum Individual Dose for Controlling Age Group at Controlling Location. The Month-to-Date dose values will contain the summation of the doses for all *Openn and Closed" permits Including the permit for which the report Is being generated. These dose values will appear immediately below the "This Release" row of doses. 6.3.2.2 On the Pre-Release Permit Report (3.01), an Alert Alarm Setpoint will appear below the Max Monitor Setpoint Value. The Alert Alarm Setpoint will be calculated by using the multiplying the release point setpoint value by a multiplier specified with the ALRT SET parameter mentioned above. 6.3.2.3 On the Liquid Special Report (3.02), an Alert Alarm Setpolnt will appear below the Release Point and Discharge Point Setpolnt values in the Radiation Monitor(s) portion of the report.

                                                 .9-

Software Requirements Specification RS448 6.3.2.4 On the Pre-Release Permit Report (3.01), the calculation of setpoint data for additional dilution flow rates (under Pre-Release Calculations) will use dilution flow rate values from the Dilution Streams transaction [EM-DM-DS] for a specific dilution stream. Up to four dilution flow rates which are lamer than the dilution flow rate parameter entered on the Uquid Permit Definition Screen (3.06) will be used. 6.3.3 Underlying Calculations The calculations performed by the EMS software for Uquid Pre-Release Permits shall produce the same results as those described in Chapter 2 (sections 2.1-2.6) of the EMS Technical Reference Manual (Reference 2.1.2), with no revisions.

                                               -1 0     z        \

It Software Requirements Specification RS-8448-04 6.4 Liquid Post-Release Processing,, 6.4.1 User Interface and Functionality Liquid Post-Release Processing functionality for the EMS software shall be as described In section 3 of the EMS Operator's Manual (Reference 2.1.1), with the following revisions: 6.4.1.1 On the Liquid Permit Definition Screen (Screen 3.13): If the DILOOKUP parameter is set to "Y' for the release point and a user uses the 'Tab" or 'Retumr key to exit the Dilution Flow Rate parameter on the Permit Definition Screen and the Dilution Flow Rate parameter has a value of zero, a selection screen with two columns of data will appear. One column will contain the pump configuration description, while the other will contain the dilution flow rate for each associated pump configuration. Upon selection of the Dilution Flow Rate, the selection screen will disappear and the selected dilution flow rate will appear in the Dilution Flow Rate parameter on the Permit Definition Screen. The cursor will then automatically advance to the Dilution Volume Parameter. 6.4.1.2 On the Uquid Permit Definition Screen (Screen 3.13): When a "ilr (Fl 4) or a "Save" (F1 0) without a Fill Is executed, if the DILOOKUP parameter is set to Y' for the release point associated with the current permit being processed and the Dilution Flow Rate parameter is set to zero, a selection screen, as described above will appear. Once a selection of the Dilution Flow Rate is complete, the selection screen will disappear and the "Fill' operation will continue. Upon completion, the selected dilution flow rate will appear in the Dilution Flow Rate parameter on the Permit Definition Screen. If the Dilution Flow Rate parameter on the Permit Definition Screen is not set to zero and the DILOOKUP parameter is set to "Y", the fill will proceed as normal without the dilution flow rate selection screen appearing. 6.4.1.3 (Item removed since actual waste flow is known at time of post release processing.)

                                              -1 1-    e

Software Requirements Specification RS-8448-04 6.4.1.4 The Monitor Response Screens for Release Points and Discharge Points (Screen 3.08) will appear while processing a Post-Release Permit when the Response Option Is set to 'Y" on the Release Point transaction'[EM-DM-RP (Form 1)]. These screens will appearfollowing the Nuclide Concentration Screen (Screen 3.15). The monitor response values entered should include the Monitor background values. 6.4.2 Associated Reports. Uquid Post-Release Permit Report shall be as described in section 3 (pages 3-59 through 3-62 of the EMS Operator's Manual (Reference 2.1.1), wIth the following revisions: 6.4.2.1 On the Post-Release Permit Report (3.03), the Cumulative Month-to-Date Doses will appear on the page with the report category of Cumulative Maximum Individual Dose for Controlling'Age Group at Controlling Location. 'The Month-to-Date dose values will 'coritain the summation of the doses for all Open' and 'Closed" permits Including the permit for which the report Is being generated. These dose values will appear immediately belowthe wThis Release row of doses. 6.4.3 Underlying Calculations The calculations performed by the EMS software for. Uquid Post-Release Permits shall produce the same results as those described In Chapter 2 (section 2.7) of the EMS Technical Reference Manual (Reference 2.1.2), with no revisions. I I Software Requirements Specification RS-8448-04 6.5 . Liquid Permit Editing 6.5.1 User Interface and Functionality Functionality for editing liquid permits through the EMS software shall be as described In section 3 of the EMS Operator's Manual (Reference 2.1.1), with the following revisions: The appearance and functionality of the liquid permit definition screen and the monitor response screen shall be modified as described for the Pre-Release stage In sections 6.3.1 and 6.4.1 above. 6.5.2 Associated Reports The permit report format and contents for edited open and closed liquid permits shall be as specified above for original permit 'reports', in sections 6.3.2 and 6.4.2, respecfively. 6.5.3 Underlying Calculations The calculation methods for editing open and closed liquid permits shall be as specified above for original calculations, in sections 6.3.3 and 6.4.3, respectively. 6.6 Liquid Permit Deletion Functionality for deleting liquid permits through the EMS software shall be described section 3 or the EMS operator's Manual (Reference 2.1.1).

                                                       -kzjo I --T- . I (O Software Requirements Specification                                             RS-8448-04 6.7   Gaseous Pre-Release Processing 6.7.1  User Interface and Functionality Gaseous Pre-Release Processing functionality for the EMS software shall be as described In section 4 of the EMS Operator's Manual (Reference 2.1.1), with the following revisions:

6.7.1.1 On the Gaseous Permit Definition Screen (Screen 4.05): The Initial Pressure'and Final Pressure parameters shall be deleted. 6.7.1.2 On the Gaseous Nuclide Concentration Screen (Screen 4.06): If the SCAL NUC parameter is set to "A, when exiting the Concentration Screen by hitting 'Process' (Do), the user will be prompted for the total nuclide concentration of permit. The concentrations are then 'scaled' and then stored Internally. As a result, the concentrations displayed on the screen will remain unchanged. (See the Underlying Calculations section for Pre-Release'Perrnit'Processing for an explanation of the 6scaling" of concentrations.) NOTE: This method requires the VAXGSP (F12) file transfer has occurred bringing the representative nuclide concentration values to the screen prior to 'Save' of data. 6.7.1.3 For releases with low orzero gamma emitter concentrations that result In a pre-diluted MPCOratio less than 10%, a default concentration will be used for setpolntcalculations. This default'concentration will not be used for updating curie, dose rates, or dose totals. The default nuclide will be attained from the DEF NUC parameter. The default concentratldn~for this nuclide will be attained from the DEF._CONC parameter.-The default type for the default nuclide should be attained from the DEF._TYPE parameter. 6.7.1.4 The Monitor Response Screens for Release Points and Discharge Points (Screen 4.08) will no longer appear while processing a Pre-Release Permit when the Response Option is set to wY' on the Release Point transaction [EM-DM-RP (Form 1)]. 6.7.1.5 Prior to entering the Gaseous Permit Approval Screen (Screen 4.09): IfIt Is determined that the computed maximum waste flow Is less than the anticipated waste flow, the anticipated waste flow will be changed to have the value of the computed maximum waste flow. Ifthe anticipated waste flow is modified, setpoint, dose, and dose rate values will be recalculated based on the new value.

II Software Requirements Specification RS48448-04 6.7.2 Associated Reports 0 Gaseous Pre-Release Permit Reports shall be as described in section 4 (pages 4-49 through.4-58) of the EMS Operator's Manual (Reference 2.1.1), with the following revisions: 6.7.2.1 On the Pre-Release Permit Report (4.01), the Cumulative Month-to-Date Doses will appear on the pages with the report category of Cumulative Dose at Site Boundary and Cumulative Maximum Individual Dose for Controlling Age Group at Controlling Location. The Month-to-Date dose values will contain the summation of the doses for all "Open" and "Closed" permits Including the permit for which the report is being generated. These dose values will appear Immediately below the "This Release" row of doses. 6.7.2.2 On the Pre-Release Permit Report (4.01), the Oscaled" noble gas concentrations shall appear on the Isotopic Identification page of the report if the SCALNUC parameter Is set to "Yu for the release point where the release Is being made. 6.72.3 On the Pre-Release Permit Report (4.01), the Noble Gas Alert Alarm Setpolnt will appear below the Max Monitor Setpolnt values. The Alert Alarm Setpolnt will be calculated by multiplying the noble gas monitor setpoint value by a multiplier specified with the ALRTSET parameter mentioned above. 6.7.2.4 On the Gaseous.Special Report (4.02), the Noble Gas Alert Alarm Setpolnt will' appear below the Release Point and Discharge Point Setpoint values in the Radiation Monitor(s) portion of the report. It will be calculated as mentioned above. 6.725 -On the Pre-Release Permit Report (4.01), the Initial and Final Pressure parameters will be removed from the Pre-Release Data section of page one of the report. 6.7.3 Underlying Calculations The calculations performed by the EMS, software for Gaseous Pre-Release Permits shall produce the same results as those described in Chapter 3 (section 3.1-3.6) of the EMS Technical Reference Manual (Reference 2.1.2), with the following revisions and clarifications:

                                                                                                @1

Software Requirements Specification Ii RS41448044 6.7.3.1 Dose Calculations will appear in the site specific technical reference manual as follows: For Noble Gas Total Body Dose Rate (for vents or stacks <c8 meters): Dt = shf *XQg

• 8 76 0 -a Fo *Z (K1 *ORiv) where Dt = the total body dose rate due to gamma emissions by noble gas releases from vent v (mrernlyr) shf shielding factor (dimensionless)

QRIl release rate of noble gas radionuclides, i, In gaseous effluents from vent or stack v ( pCilsec). FO occupancy factor defined for the receptor at the given location (dimensionless)

                        *K1     =   total body dose factor due to gamma emissions for noble gas radionucfide I (mreri/yr per pCVm 3 )

XIag = highest value of the noble gas 1-hour X/Q for gamma radiation for vent or stack v at the site boundary, (sec/m 3) 8760-a= adjustment factor used to convert the 1-hour X/a value to an average 1 year X/Q value (dimensionless) where 8760 = number of hours Ina year

                               -a       =   "a' factor for gamma noble gas X/Q

• For Noble Gas Total Body'Dose (for vents or stacks < 80 meters):

                                =    shf Fo

• z (Ki
• QRlv)
• XfQg* ta I,

(5256 *105 l dur) where Dth = total body dose from gaseous effluents (mrem) 5.256

• 105 = number of minutes in a year dur = duration of the release (minutes)

                                                     -fqe_ kl

II Software Requirements Specification RS-8448-04 t-a = adjustment factor to convert the 1-hour X/Q value to the short term X/I value for the release (dimensionless) where t = duration of release (hours) a = "a factor for gamma noble gas X/Q

• For Noble Gas Skin Dose Rate (for vents or stacks < 80 meters):

Ds= shf

• Fo *Z QRiv * [(L - XQ -8760-b) + (1.1 Mj
• X/Qg . 8 7 60 a)]

where Ds = skin dose rate from gaseous effluents (mrernyr) X/Q = highest value of the noble gas 1-hour X/Q for vent or stack V at the site boundary (sec/r 3 ) Ml = air dose factor due to gamma emissions for noble gas radionucfide i (mrad/yr per pC0r m 3 ) 1.11 = conversion factor from mrad to mrem 4 = skin dose factor due to beta emissions for noble gas radionuclide I (mrerrmyr per pCVm3 ) b = afactor for noble gas X/Q

• For Noble Gas Skin Dose (for vents or stacks < 80 meters):

shf* FO Z Riv" [(4

• X/Q
• t-b) + (1.1 IM 1 . Xjg
• t-a)]

DSk = (5.256

• 105 I dur) where Dsk = total skin dose from gaseous effluents (mrem)

Software Requirements Specification RS W8044

• For Noble Gas Air Dose due to gamma radiation (for vents or stacks < 80 meters):

Dy = (3.17 10-8)uiXfQg t-a Fo *- MI *Qiv where DY = total gamma air dose from gaseous effluents (mrad) 3.17

• 10-8 inverse of number of seconds in a year Qiv- = release of noble gas radionuclides, i, in gaseous effluents from vent or stack v (pCi)

QIv = QRiv dur 60 where 60= number of seconds in a minute For Noble Gas Air Dose due to beta radiation (for vents or stacks < 80 meters): Dp (3 .17 -8).SQ fb -Fo.-NI-Qiv where Dp = total beta air dose from gaseous effluents (mrad) N1 = air dose factor due to beta emissions for noble gas radionuclide I (mradfyr per pCVm3 )

• For Critical Organ Dose Rate-Inhalation Pathway and all Pathways for H-3, C-14 (for vents or stacks < 80 meters):

DRTa = XOr' 8760Cc

• PrpTa
• QRiv where DRTa = dose rate for age group a and organ T from lodines and particulates with half lives greater than 8 days In gaseous effluents (mrem/yr)

Pi'pa = dose factor for each radionuclide i. pathway p, organ T, and age group a (mremlyr per pCI/m 3 ) la \

I I Software Requirements Specification RS-8448-04 X/Or = highest value of the radioiodine/particulate 1-hour XJQ for vent or stack v at the site boundary (sec/m 3) c = "aw factor for Radiolodine/Particulate X/Q Note: It is assumed Pipta will not contain long term X/Q or D/Q values. For Critical Organ Dose Rate-Ground and Food Pathways (for vents or stacks < 80 meters): DRTa = D/Q 8760-d RipTa.QFR where D/Q = highest value of the 1-hour deposition factor at the distance of the site boundary (lImZ) d = ag factor for DlQ RipTa = dose factor for each radionuclide i, pathway p, organ T, and age group a (m2

• mrem/yr per pCVsec)

Note: It is assumed RipTa will not contain long term X/Q or DIQ values.

• For Critical Organ Dose-inhalation Pathway and all Pathways for H-3, C-14 (for vents or stacks < 80 meters):

DTa = (3.17*10 8 )*XlQr*tc*FO*" Pipra4Qiv where DTa = dose for age group a and'organ T from lodines and particulates with half lives greater than 8 days in gaseous effluents (mrem) Note: It is assumed PipTa will not contain long term X/C or D0/ values. 0

Software Requirements Specification RS-8448-04 For Critical Organ Dose-Ground and Food Pathways (for vents or stacks < 80 meters): Dta (3.17

• 10-8)
• DQ*t-d *FoZ Ripa*iv Note: it Is assumed RipTa will not contain long term X/Q or D/Q values.

6.7.3.2 On the Nuclide Concentration Screen (Screen 4.06), nuclide concentrations will be scaledw If the SCALNUC parameter is set properly for a Release Point. This "scaling" is described as follows: Cinew = (to s) -Ci' - . , where

                                    =     concentration (after"scaring") of nuclide.

s = sum of all nuclide concentrations on the Nuclide Concentration Screen. t = total nucilde concentration entered by the user Co = concentration (before "scaiing") of nuclide1

                                        .     ,    ..   .    .. ^ ,   . .  .    ..
                        *    \
                                                          -2O

I I Software Requirements Specification RS-8448-04 6.8 Gaseous Post-Release Processing 6.8.1 User Interface and Functionality Gaseous Post-Release Processing functionality for the EMS software shall be as described in section 4 of the EMS Operator's Manual (Reference 2.1.1), with the following revisions: 6.8.1.1, On the Gaseous Permit Definition Screen (Screen 4.14): The Initial Pressure and Final Pressure parameters shall be deleted. 6.8.1.2 On the Gaseous Nuclide Concentration Screen (Screen 4.15): If the SCALNUC parameter Is set to "Y*, when exiting the Concentration Screen by hitting 'Process (Do), the user will be prompted for the total nuclide concentration of permit. The value entered for the total nuclide concentration while opening the permit shall be displayed as a default value which Ican be modified. Once the value is entered/accepted the concentrations are then 'scaled" and then stored Internally. As a result, the concentrations displayed on the screen will remain unchanged. (See the Underlying Calculations section for Post-Release Permit Processing for an explanation of the 'scaling" of concentrations.) NOTE: This method requires the VAX GSP (F1 2) file transfer has occurred bringing the representative nuclide concentration values to the screen prior to "Save" of data. 6.8.1.3 The Monitor Response Screens for Release Points and Discharge Points (Screen 4.08) will appear while processing a Post-Release Permit when the Response Option is set to "Y* on the Release Point transaction [EM-DM-RP (Form 1)]. These screens will appear following the Nuclide Concentration Screen (Screen 4.15). The monitor response values should Include the monitor background values. 6.8.1.4 (Item removed, since actual waste flow is known at time of post release processing.) 6.8.2 Associated Reports Gaseous Post-Release Permit Reports shall be as described in section 4 (pages 4-58 through 4-63) of the EMS Operator's Manual (Reference 2.1.1), with the following revisions: 0

Software Requirements Specification RS84804 6.8.2.1 On the Post-Release Permit Report (4.03), the Cumulative Month-to-Date Doses will appear on the pages with the report category of Cumulative Dose at Site Boundary and Cumulative Maximum Individual Dose for Controlling Age Group at Controlling Location. The Month-to-Date dose values will contain the summation of the doses for all "Open" and Closed' permits including the permit for which the report is being generated. These dose values will appear Immediately below the "This Release" row of doses. 6.8.2.2 On the Post-Release' Permit Report (4.03), the 'scaled' noble gas concentrations shall appear on the Isotopic Identification page of the report If the SCALNUC parameter Is set to 'Y' for the release point where the release Is being made. 6.8.2.3 On the Post-Release Permit Report (4.03), the Initial and Final Pressure parameters will be removed from the Pre-Release Data section of page one of the report.' 6.8.3 Underlying Calculations The calculations performed by the EMS software for Gaseous Post-Release Permits shall produce the same results as those described In Chapter 3 (section 3.7) of the EMS Technical Reference'Manual (Reference 2.1.2), with the following revisions and cLauficatlons: 6.8.3.1 Dose Calculations will appear in the site specific technical reference manual as follows: '

• For Noble Gas Total Body Dose Rate (for vents or stacks < 80 meters):

Dt = shf X/Qg 8760-a - Fo .z (rj *QRiv) where'-

                          *Dt      =  the total body dose rate due to gamma emissions by noble gas'releases from vent v (mremlyr) shf     =   shielding factor (dimensionless)

QRiv = release rate of noble gas radionuclides, i, in gaseous effluents from vent or stack v ( pCVsec). Fo z!.Aoccupancy factor defined for the receptor at the given location (dimensionless) 1$ = total body dose factor due to gamma emissions for noble gas radionuclide I (mremfyr per pCi/m3 )

                                               .22-

I I Software Requirements Specification RS-448-04 X(Qg = highest value of the noble gas 1-hour X/Q for gamma radiation for vent or stack v at the site boundary, (sec/m 3 ) 8760 a= adjustment factor used to convert the 1-hour X/Q value to an average 1 year X/O value (dimensionless) where 8760 = number of hours in a year a = 'a' factor for gamma noble gas X/O For Noble Gas Total Body Dose (for vents or stacks < 80 meters): shf

• Fo
• Z (KI - QRiv)
• XtQg . t a Do =

(5.256* 105 / dur) where Dt - total body dose from gaseous effluents (mrem) 5.256* 10 5 = number of minutes In ayear dur = duration of the release (minutes) S' t-a = adjustment factor to convert the 1-hour X/Q value to the short term X/Q value for the release (dimensionless) where t = duration of release (hours) a = ae factor for gamma noble gas X/Q

• For Noble Gas Skin Dose Rate (for vents or stacks <80 meters):

Ds = shf

• Fo - XQRiv* -[(4 XtQ 8760-b) + (1.1 1M; X/Qg . 8 7 60-a)]

where Ds = skin dose rate from gaseous effluents (mremlyr) XIQ = highest value of the noble gas 1-hour X/Q for vent or stack v at the site boundary (sec/m 3 ) Software Requirements Specification RS-4A48044 ml = air dose factor due to gamma emissions for noble gas radionuclide I (mrad/yr per pCVm 3) 1.11 = conversion factor from mrad to mrem 4j = skin dose factor due to beta emissions for noble gas radionuclide I (mrem/yr per pCi/m 3) b = -a factor for noble gas X/Q For Noble Gas Skin Dose (for vents or stacks < 80 meters): shf

• Fo
• Riv * [(4 X/Q
• t-b) OF(1.11I M;
• XjQg *t-a.)]

(5.256

• 105 dur) where DSk = total skin dose from gaseous effluents (mrem)
• For Noble Gas Air Dose due to gamma radiation (for vents or stacks < 80 meters):

Dy 'r=(3.17* 10-8) *X/Og *t-a - FO"

• M;
• Qiv where D = total gamma air dose from gaseous effluents (mrad) 3.17
• 10-8 Inverse of number of seconds in a year Qiv= release of noble gas radionuclides, i, in gaseous effluents

                                   -froim vent or stack v (pCi)

Qiv = QRiv dur 60 where 60= number of seconds in a minute I I Software Requirements Specification RS-8448-04

• For Noble Gas Air Dose due to beta radiation (for vents or stacks <

80 meters): S DP (3.17

• 10-8)XIQ tbFo Z Ni Oiv where DP - total beta air dose from gaseous effluents (mrad)

N1 = air dose factor due to beta emissions for noble gas radionuclide i (mrad/yr per pCVm 3)

• For Critical Organ Dose Rate-Inhalation Pathway and all Pathways for H-3, Ce14 (for vents or stacks < 80 meters):

DRya X/Or 8760c Z Pipra *QRi where DRTa = dose rate for age group a and organ T from iodines and particulates with half lives greater than 8 days in gaseous effluents (mremtyr) PipTa, dose factor for each radionuclide i, pathway p, organ T, and age group a (mrenlyr per pCVm 3) 0, X/Or =. highest value of the radioiodine/partlculate 1-hour X/Q for vent or stack v at the site boundary (sec/M 3 ) c = a factor for Radiolodine/Particulate X/Q Note: It is assumed Pipa will not contain long term X/Q or D/Q values.

• For Critical Organ Dose RateGround and Food Pathways (for vents or stacks < 80 meters):

DR~a = D/Q*8760-d*ZRi *'QRjv where DIQ = highest value of the 1-hour de position factor at the distance of the site boundary (1/mi) d = Wa factor for DIQ t 9'w

Software Requirements Specification RS-844804 RipTa = dose factor for each radionuclide i, pathway p, organ T, and age group a (m2 *mrern/yr per uCVsec) Note: It is assumed RipTa will not contain long terri X/Q or DIQ values. For Critical Organ Dose-Inhalation Pathway and all Pathways for H-3, C-1 4 (for vents or stacks < 80 meters): DTa = (3.17 108) - XQQr t-c Fo

• ZPjpTa
• ajv where DTa = dose for age group a and organ T from lodines and particulates with half lives greater than 8 days in gaseous effluents (mrem)

Note: It is assumed PjpTa will not contain long term X/Q or DIQ values.

• For Critical Organ Dose-Ground and Food Pathways (for vents or stacks < 80 meters):

Dra D~,

                                =

(3.17 108)- D/Q t-d Fo

                                       .- , .i .a.

RjpTaQiv

                                                                                       . .h Note:        It is assumed RipTa will not contain long term XIQ or DIQ values.

6.8.3.2 On the Nuclide Concentration Screen (Screen 4.15), nuclide concentrations will be scaled' If the SCALNUC parameter is set properly for a Release Point. This "scaling" is described as follows: Cinew = (t I s)

• Cl where Cijnw = concentration (after scaiung") of nuclidei s = sum of all nucilde concentrations on the Nuclide Concentration Screen.

t = total nuclide concentration entered by the user C1 = concentration (before scaling") of nuclidej II Software Requirements Specification RS-8448-04 6.9 Gaseous Permit Editing 6.9.1 User Interface and Functionality Functionality for editing gaseous permits through the EMS software shall be described in section 4 of the EMS Operators Manual (Reference 2.1.1), with the following revisions: The appearance and functionality of the gaseous permit definition screen, the monitor response screen, and nuclide concentration shall be modified as described for the Pre- and Post-Release stages in sections 6.7.1 and 6.8.1 above; 6.9.2 Associated Reports The permit report format and contents for edited open and closed gaseous permits shall be as specified above for driginal permit reports, In sections 6.7.2 and 6.8.2, respectively. 6.9.3 Underlying Calculations The calculation methods for editing open and closed gaseous permits shall be specified for original calculations, in sections 6.7.3 and 6.8.3, respectively. 6.10 Gaseous Permit Deletion Functionality for deleting gaseous permits through the EMS software shall be described section 4 or the EMS operator's Manual (Reference 2.1.1). Software Requirements Specification RS-8448-04 6.11 Semi-Annual Reporting 6.11.1 User Interface and Functionality Semi-Annual Reporting functionality for the EMS software shall be as described in section 5 of the EMS Operator's Manual (Reference 2.1.1), with the following revisions: 6.11.1.1 On Report 5.01 (Gaseous Summation of All Releases): Compute each value on line A.3 of the report by taking the greater of J 1°° Dag / QLag I lOO'inDabIQL1_a where Dag = the gamma air dose in the applicable quarter at the site boundary receptor due to noble gas emissions (mrem) Dab - the beta air dose In the applicable quarter at the site boundary due to noble gas emissions (mrem) QLag = the quarterly limit on Dag (mrem) [usually 5]

                          )Lab = the quarterly limit on Dab (mrem) [usually 10]

A note will be made at the bottom of the report stating whether the beta air dose and its associated limit or gamma air dose and Its associated limit were used for the Percent of Applicable Limit of Fission and Activation Products.. The values on lines B.3, C.3, and D.3 will be the equivalent. They will be calculated as follows: the greatest(overT) of { 100-(Z DjT)IQy 1II Software Requirements Specification RS-8448-04 where DiT the dose to organ T of the controlling receptor, in the applicable quarter, due to gaseous emissions of radionuclide I (mrem) The summation is over all non-noble gas radionuclides with half-lives greater than 8 days, Including radioiodines, particulates, and tritium. QLro = the quarterly limit on the controlling receptor organ dose due to gaseous effluents (mrem) [usually 7.5] 6.11.1.2 On Report 5.02 (Liquid Summation of All Releases):

• For each quaiter q In the report, calculate the reportable dilution volume (DVrql in liters) for the portion of the quarter that is within the report dates. It is the sum of the reportable monthly dilution volumes (DVrm) in user units for all the months In the quarter that'are within the report dates:

DV' = 28.31685 - sd Wolf - X DVrm

                    'The values D~rm are from the column tvol of the QDVOL table. The value DVr' Is Included In the report on line F. and Is used in the calculations below.' sd'_ivolr should be the user unit conversion factor to convert from user units toVft 3 . 28.31685 is a unit conversion factor from ft3 to liters.'-

• For each space on a line titled "AVERAGE DILUTED CONCENTRATION DURING PERIOD', the average concentration (Cq, in pCi/ml) for the respective quarter Is computed as follows (where i ranges over only the nuclides in the category):

Cq = Z CIq -Z [Actq /(1000

• DVrq)]

where Actiq = total activity of nuclide I released during the portion of the quarter q that is within the period (pCi) DVrq = reportable dilution flow for the portion of quarter q that is within the report period (liters), as calculated above.

• Compute each value on Nne A.3 and B.3 of the report by taking the greater of { 100. D1t I QLbt 3,

Software Requirements Specification RS-8448-04 where

                           'Da      -the liquid total body dose in the applicable quarter at the site boundary receptor (mrem)

Djo = the liquid maximum'organ dose in the applicable quarter at the site boundary (mrem) QLft = the quarterly limit on Dit (mrem) [usually 1.5] 0- 10 = the quarterly limit on D1O(mrem) [usually 5] A note will be made at the bottom of the report stating whether the liquid total body dose and its-associated limit or maximum organ dose and Its associated limit were used for the Percent of Applicable Umit. Compute each value on fine C.3 of the report as follows: PqUIOOCq/Ldg where Cq = sum of noble gas concentrations Pq = Percentage applicable to a given quiarter for dissolved and entrained gases Ldg = Uquid dissolved gas limit (pCI/ml) [usually 2.OE-04] 6.11.2 EMS Trend Plots Trend Plotting functionality for the EMS software shall be described in section 5 of the EMS Operator's Manual (Reference 2.1.1) with no revisions. 6.12 End-of-the-Year Data Archiving 6.12.1 User Interface and Functionality _End-of-the-Year Data Archiving functionality for the EMS software shall be described in sectiorn 6 'of the EMS Operator's Manual (Reference 2.1.1) with no revisions..

                                               .30-

II Documentation Review Report 0 Document Reviewed Sa'S AKRONe 10i o )UAS ma- d4'8-a4 Does the document meet the requirements? or No Is the document approved?) or No If not please state the exceptions: Signature: 6 1M 4/,U014;k Date:9 //5'<-Soz

                                               ,Aa-3 4                         00-5792 0

Documentation Review Report Document Reviewed 5, sE CA,-Io rt,5 t,-usc Does the document meet the requirements?9-or No Is the document approved?(Yi or No ifnot please state the exceptions: Signature: n I A-ela41 Date: ___ /___ _4 _ T4P- 3xS 00-6722

APPENDIX C: EMS SOFTWARE DOCUMENTATION ATTACHBMENT 4: TECHMCAL REFERENCE MANUAL, EFFLUENT MANAGEMENT SYSTEM NAESCO SEABROOK STATION, JULY 1994, FP 75486 C-6 ODCM Rev. 22

Inc. Canberra Industries, Parkway 800 Research 06450 9-, Meriden, CT July 1994 Station NAESCO Seabrook Reference Manual EMS Technical O.j 07-0625 i I i o i Inc. I 1994, Canberra Industries, I Copyright ile  ! Printed in U.S.A. 4 I i i ODCM Q6V2a II Covfc i

II Copyright 1994, Canberra Industries, Inc. All rights reserved. This manual contains proprietary information; no part of it may be reproduced or used in any form or by any means - graphic, electronic, or mechanical, including photocopying, recording, or information storage and retrieval systems - without the written permission of Canberra Industries. The information in this manual describes the product as accurately as possible, but is subject to change without notice. Printed in the United States of America.

                                                                                                   @60 09 fn 37 A                    OM-reJ Z2-a.

0 -

0. CoG fr TABLE OF CONTENTS IA..

_l l Page CHAPTER 1 INTRODUCTION ............................. 1-1 1.1 SETPOINT CALCULATIONS ..................................................................................... 1-2 1.2 RELEASE PROCESSING ........................................................................................... 1-3 1.3 COMPOSITE NUCLIDES ........................................................................................... 1-3 2-1 CHAPTER 2 LIQUID RELEASE CALCULATIONS .......................................................... 2-1 2.1 LIQUID PRE-RELEASE PERMIT .............................. ; 2-1 2.2 10CFR20 COMPLIANCE ..... ................................................................................. 2-1 Dissolved and Entrained Gases ......................  ; .; 2-4 2.3 MAXIMUM WASTE FLOW .;..;.................................................................................... 2-4 2.4 MINIMUM DiLUTION FLOW RATE ............................................................................ 2-5 2.5 SETPOINT CALCULATIONS ..................................................................................... 2-5

• '0 Recommended Setpoint.................................................................................... 2-8 Setpoint In pCiml Clm.... ........ 2-8 Recommended Setpoint In User Units (e.g. cpm) ..................................... 2-10 Setpoint for Discharge Point ..................... 2-11 Setpoit InpC Vsec......

Sepints inC c.............................................................................. 2-13 2.6 DOSE CALCULATIONS FOR LIQUID RELEASES ..................... 2-13

     .2.7 31 DAY PROJECTED DOSE CALCULATIONS ........................................................ 2-17 2.8 POST-RELEASE PROCESSING ............................................................................... 2-18 CHAPTER 3 GASEOUS RELEASE CALCULATIONS                                  .....................................................                              3-1 3.1 GAS PRE-RELEASE PERMIT .............................................. 3-1 3.2 RADIONUCLIDE ACTIVITIES AND COMPOSITE VALUES ......................... ........... 3-1 Activity Released ..............................................                                                                                  3-2 3.3 IOCFR20 COMPLIANCE                               ..............................................                                                          3-2 3.4 SETPOINT DETERMINATION                                   ............................                                                                    3-5 Noble Gases .............................                                                                                                         3-5 Radioiodines and Particulates ............................                                                                                        36 3.4a SETPOINTS .............................                                                                                                                 3-6 3.4b REPORTED SETPOINTS ............                                          ..                                                                          3-10 Setpoints InpC~lsec .3.........................................................................................                                 3-11 3.5 MAXIMUM WASTE FLOW ............................                                                                                                        3-12 3.6 DOSE RATE AND CUMULATIVE DOSE CALCULATIONS ...................................... 3-12 Noble Gas Dose and Dose Rate Calculations .............................................. 3-12 i

P3Q V6- - GoDC 1(2of2.2

II Organ Dose Calculations............................................. 3-15 3.7 RESOLVING DOUBLE-COUNTING OF DOSE AND ACTIVITY .............................. 3-16 A.10 3.8 31 DAY PROJECTED DOSE CALCULATIONS ........................................................ 3-17 3.9 GAS POST-RELEASE PROCESSING ....................................................................... 3-18 CHAPTER 4 LIQUID DOSE FACTOR EQUAT .~4.. ........................................................ 4-1 4.1 POTABLE W ATER ...................................................................................................... 4-2 4.2 AQUATIC FOODS PATHWAYS ................................................................................. 4-2 4.3 SHORELINE RECREATION PATHWAY.................................................................... 4-3 4.4 IRRIGATED VEGETABLE PATHWAY ....................................................................... 4-4 4.5 REDUCTION TO NUREG-0133 EQUATIONS ........................................................... 4-5 . CHAPTER 5 GAS DOSE FACTOR CALCULATIONS ...................................................... 5-1 -5.1 INHALATION PATHWAY............................................................................................ 5-1 5.2 GROUND PLANE PATHWAY .......................  ; .. 5-2 5.3 MILK PATHWAY A.. ........................ . 5-2 Carbon-14 In Milk ..................... 5-4 TMitium in Milk ...................  ; 5-4 5.4 MEAT PATHWAY. ...................... . . 5-5 Carbon-14 In'Meat . ........... 5-5 Tritiurh in M eat ............................................................................................. 5.5 VEGETABLE PATHWAY ..................... Carbon-14 in Vegetables . 5-6 5-6 5-7 0* Tritiurn in Vegetables .... 5-7 5.6 REDUCTION TO NUREG-0133 EQUATIONS ........................................................... 5-7 APPENDIX A REFERENCES ............................................................................................ A-1 Ii P5 39 ebcu-, BU-ZZ

0

   .0, CHAPTER I INTRODUCTION The Effluent Management System (EMS) Software implements the requirements for determining limits and doses for the routine liquid and gaseous releases from nuclear power plants.            The calculations and methodology are based ,on.those described in U. S. Nuclear Regulatory Commission Regulatory Guide 1.109 and references described therein.      These equationsreduce to those described in NUREG-0133 by properselection of parameters.

This manual describes the..calculations used in the LRW/GRW,.program for handling liquid and gaseous releases and preparing the semi-annual report, and the equations used in the DFP option for calculating the relevant dose factors. This manual describes the new 10CFR20 (1992) as well as old 10CFR2Q requirements. For' a nuclear power plant, the Off-Site Dose Calculation Manual (ODCM) describes the methods -used at that plant for complying with 0 . the effluent release portions of the technical specificat ons and the requirements of 10CFR20 and Appendix I of 10CFR50. The concentration and dose limits that are required to be met are: o For radioactive liquid effluents, the concentrations released to areas beyond the site boundary are limited to: MPC values given in old 10CFR20, Appendix B, Table II. OR - ECL values given 'in new 10CFR20, Appendix B, Table 2. where ECL values are effluent concentration limit values. o For radioactive liquid'effluents, the maximum dose-to any member of,.the public will be less than the limits given in 10CFR50, Appendix I; o For gaseous effluents,'the- old 10CFR20 requires that the dose -rate at any location beyond the site boundary will be limited to the annual dose limits given in the Technical Specifications and' corresponding to the concentrations in Appendix B of the old 3OCFR20. The old 10CFR20 approach for gaseous effluents has been accepted by the NRC for use le. under the new 10CFR20. 1-1 age r~vu

II o For gaseous effluents, the maximum dose to any member of the public will be less than the limits given in 10CFR50, Appendix I. o The maximum dose to any member of the public will not exceed the limits given in 40CFR190. The equations employed for calculating the dose and dose factors are taken from NUREG-0133 1 and Regulatory Guide 1.109.2 For a particular nuclear plant,. the ODCM describes the physical configuration of release sources and release points for routine and non-routine liquid and gaseous effluents, the monitor setpoint calculations, dose, and dose rate calculations. .1.1 SETPOINT CALCULATIONS Calculations are made for the radiation monitors to determine the alarm/trip setpoint so that 10CFR20 compliance is met. For the old 10CFR20 compliance, liquid calculations use the maximum permissible concentrations from 10CFR20 App. B, Table 2, column 2, and the more conservative value (smaller) of the soluble and insoluble values while gas calculations use dose rate equations and limits from NUREG-0133. To comply with the new 10CFR20 requirements, the effluent concentration limits are used for liquid setpoint calculations. For gaseous setpoint calculations under the new 10CFR20, the NRC is still allowing the use of dose rate equations and limits from NUREG-0133.. In the terminology of EMS, individual sources of radiation, such as storage tanks, the. containment building, etc., are defined as "release points." Several release points may lead to the same "discharge point." Setpoint calculations produce monitor limiting values in activity units (pCi/ml or pCi/cc). These are then converted to user units, e.g; counts per minute (cpm). For gaseous releases, setpoint can be reported as release rates (pCi/sec}. The reporting units for each monitor can be defined separately. EMS allows setpoints to be set for both the release points and the discharge points. In the case that the release point and the discharge point are the same, or use the same physical monitor, the same discharge setpoint value is reported for both. This use of the same discharge setpoint value can be disabled.

                                                                         'S 1-2 PD 41

0 '_ EMS has a "nuclide specific" option. in this option only the nuclides listed in the monitor slope table are used in the setpoint calculations. 1.2 RELEASE PROCESSING For batch releases, the processing of releases consists of sampling the tank or volume of air, analyzing the radionuclide content, then using the radionuclide' concentrations and estimated release flows, volumes, etc. and 6alculating the doses and setpoints, comparing to the IOCFR20 limits, and -comparing to the 10CFRSO limits.- If the limits are not exceeded, the pre-release permit is signed off 'and the release can occur. After the-release, post-release processing performs the same calculations (except the setpoints are not needed) and the database is updated with the actual values for the release. For continuous releases, -many installations prefer not, to generate an actual pre-release permit, but for the sake of analogous operation, pre-release"'calculations must still be made 'in EMS. After review, the post-release calculations are made to update the database. EMS does not allow more that'one.'open release'at a time for a single release point. However, multiple releases may be open for'one discharge point. Also, for discharge points, the setpoint is calculated by summing over all open releases for the time period involved. An alternative approach if a new permit must be opened before the actual information for the previous permit are available is to go ahead and close the release using the pre-release values and then edit this closed release later when the actual information becomes available.

• -10 1.3 COMPOSITE NUCLIDES The standard radionuclide analysis, with high-resolution germanium detectors, quantifies the gamma-emitting radionuclides. Pure beta emitters, nuclides that decay by X-capture, and alpha emitters are handled with other detection mechanisms. These are usually not tracked individually by sample, but as a composite of many samples over a month or quarter period. The concentrations of the composite nuclides are combined with the concentrations of the individual nuclides determined from gamma analysis for each sample.

                                      - i-3                           P5'4z 0tLn4%. 2

For liquid releases, the composite nuclides are generally H-3, Fe-55, Sr-89, Sr-90, and gross alpha. For gaseous releases, Fe-55 is generally not included. S,' In EMS, these are contained in an editable file designated by the composite ID number. Each release point definition specifies which composite ID is used with the release point. These can be the composite nuclides, or any other nuclides desired. Composite samples produced by taking portions of the samples from individual releases are analyzed after the releases are over. Since these generally do not vary much from one period to the neXt, it is common to use the most recent values. However, EMS provides the option of updating the composite values for the proper time period and recalculating the activity and dose.values in the database. The EMS composite update process processes only those nuclides listed in the .Composite ID for.the release point. For each nuclide, the curies and doses based on the previous value are subtracted from the cumulative totals and the curies and doses based on the correct value are added into the cumulative totals. For the setting of flags to control options in the EMS code, see the NAESCO Seabrook Station EMS Operator's Manual 07-0589. 460' P3L?3

                                                                    *0-1-4 343eQ..z

0 0 CHAPTER 2 LIQUID-RELEASE CALCULATIONS 2.1 LIQUID PRE-RELEASE PERMIT A liquid pre-release permit is generated with a program that uses the nuclide activities to determine the radiation monitor setpoint (for 10CFR20 compliance) and the potential doses for 10CFI.50 compliance. Continuous releases are treated similarly.

• -0 2.2 10CFR20 COMPLIANCE 10CFR20 compliance calculations are broken down into two paths. The first path calculates com~pliance with the old 10CFR20 in which the calculations are based on Maximum Permissible Concentrations. The second path complies with the new IOCFR20 and is Effluent Concentration Limits based.

10CFR20 requires that the sum of concentrations divided by MPC (old 10CFR20) or ECL (new 10CFR20) values must not exceed unity for MPCs or 10 for ECLs: OLD 10CFR20 NEW 0CFR20 S = X. C./MPCi S I I 1 1. OR , S - Z. C./ECL. ' 10

                                                          .1  1 for concentrations Ci released from the site.           MPCi is the maximum permissible concentration from the old IOCFR20, Appendix Br Table I}, Column 2, for nuclide i-and ECLi is the effluent concentration I.I   limit from the new 10CFR20, Appendix B, Table 2, Column 2, for nuclide i.

2-1 f5t At oCID -/

II If the summation is greater than the limit, then dilution is required. The required dilution factor is: to If the 10CFR20 option is OLD: C. I N*PC. I Dreq=. _fo R max where Dreq = Total required dilution factor Ci Concentration of nuclide.i in pCi/mL MPCi = Maximum permissible concentration of nuclide i in pCi/mIL f Release point setpoint safety factor (usually equal to. 0.5) from the release point definition. Rmax =The maximuim MPC ratio from the release point setpoint definition. If the IOCFR20 option is NEW: 0@ C. X -1 ECL.

        -D       X   I=C          I req,g          f

• R max C.

ECL. D = i-ng i req,ng f

• R max D D req req,g I Dreq,ng where Dreq,g = Required dilution factor for gamma-emitters Dreq,ng = Required dilution factor for non-gamma-emitters ECLi Effluent concentration limit of nuclide i in pCi/mL a! '

2-2 A-2 L4 5 1_3 (:ZDC4vV '&%17_z_

and the sums extend over gamma-emitters (g) and non-gamma-emitters 0 (ng), respectively. Any nuclides with MPCi 00 are excluded from the sum. Any nuclides with ECLj 0Oare excluded from the sum. The available dilution flow is the minimum dilution stream flow that can be ensured for the period of the release, corrected for other releases in process and any -activity in the dilution stream, and reduced by a safety factor.- avail Fant (ff/100) (I- z C./XXX.) where C; Concentration (pCi/ml) for nuclide i for the dilution stream sample XXXi MPCi or ECLi - ff - Flow safety factor, in percent Fant = Anticipated dilution flow rate for the 'release-O 0 ' The anticipated dilution factor is then D = ( wa+s f F )/w ant waste alloc avail waste where F t waste flow anticipated for this release F avail ' available dilution flow f ' fraction of availablp dilution stream flow allocated to alloc this releases 0.0 2-3 eLP9 OOLP-.". nt'v Z'Z'

II Dissolved and Entrained Gases To implement 10CFR20, it is also required that the total concentration of dissolved and entrained gases in liquid effluents be less than a specified value (normally, 2 E-04 pCi/mL under OLD 10CFR20, or 1 E-04 pCi/mL under NEW 10CFR20). EMS stores this limit in the Activity Limits transaction, checks this limit for each liquid permit, and indicates on the permit approval screen whether or not it is exceeded. To include dissolved noble gases in the Dreg calculation, the database must also contain the same limiting value, as the liquid MPC or ECL for each noble gas nuclide. 2.3 MAXIMUM WASTE FLOW The maximum waste flow calculation is based on the setting of the SETOPT option in the WFLOW M class of options in the Release Point Setpoint definition. This option can take on four values: NONE, NWAS (no waste), CALC or DOSE. For liquid releases, NONE, NWAS, and CALC are allowed. For liquid releases, Wmax the minimum of Rwmax and Rcwma_ where Wman = Maximum permissible waste flow rate for this release Rwmax = Release point maximum waste flow rate, as-set in the release point definition If the SETOPT option-- NONE: R = waste flow rate for the sample, Fwaste If the required dilution factor, Dreq (section 2.2) for the sample is greater than 1, Rcwmax becomes: Favail *falloc Rc~wmax =~ Dreq 1.0 If the SETOPT option = CALC Favail ' falloc + Fwaste Rcwmax, Dreq 2-4 P o

0 0 If the SETOPT option = NWAS Favail falloc Rcwmax Dreq 2.4 MINI1MUM DILUTION FLOW RATE

             -If Dre    > 1f the minimum dilution flow rate is determined as follows:                      ;

If the SET OPT option is NWAS: . Fwaste Dreq min dflow = C. falloc (ff /100) * [ 1 - x i. II] 1xxx. i where XXXi is MPCi under OLD 10CFR20, and is ECLi under NEW 10.0 10CFR20. If the SET OPT option is other than NWAS: rwaste * (Dreq - 1.O) ain dflow = C. falloc (ff / 100) 3[ 1 - Z - I xxx. 1 Otherwi .se:- - - .- - min dflow 0.0 2.5 SETPOINT CALCULATIONS Setpoints are calculated for individual release points, and for the discharge point that may combine'several release points. A setpoint adjustment factor, Sadj is determined from the value of Dreq' l ,* ' 2-5 P L48

If Dreq > 1 or the dilution factor option is N, and the setpoint equation is set to STD: Sadj = Dant / Dreq If the dilution factor option is Y. and 0 < Dreq < 1.0 or if the dilution factor option is Y, and the setpoint equation is set to NO DIL, then no credit is taken for dilution, and the setpoint adjustment factor is: Sadj = 1/Dreq If neither of these conditions is true, Sadj = 0. After the above tests, further tests are made based on the setting of the setpoint equation option, SETPEQN. These may change Sadj as follows: If the SETP EON is set to DILUTf and Fwaste > 0, then: rant + Ewaste Sadj Fwaste If the SETPEQN is set to STD, and the SETOPT option is set to NWAS, and Fwaste > 0, then: s falloc Favali 5 adj - Fwaste. Dreq Otherwise, if the SETP EQN option is set to STD, and the SETOPT option is set to other than NWAS, and Fwaste > 0, then: (falloc Fanvail.) + Fwaste Sadj Fwaste Dreg Otherwise, if the SETP EQN option is set to LOW ACT, and the SET OPT option is set to NWAS, and Fwaste > 0, then: (falloc 0 avail) Dreqng Fwaste Sadj = Dreq, g

• 2-6 9 obce r2v2.

Otherwise, if the SETP_EQN _option is set to LOW ACT, and the SETOPT option is set to other than NWAS, and Fwaste > 0, then: (falloc Favail) +'Fwaste reqng

            '                 ~waste     -

Sadj D Dreqg Otherwise, Sadj is unchanged. The setpoint adjustment factor is further tested against a limiting value (Sadjlim which is set using the.Release Point transaction in Database Ma.ntenance). If Sadj > Sadjlim then Sadj inSadjlim. All of this leads to the maximum setpoint value, Sna_, based on the gamma-emitting radionuclide mix: S (Pci/m1)~= S I C. @ max adj . where the sum extends over -all gamma-emitting nuclides (nudlides of type other than 0) in which -their concentrations are.greater than 0. In user units (cpm or other as set in the Flow Monitor' Parameters transaction in Database Maintenance), the maximum setpoint is: S (cpm.) S (R -B) +B max ad) mon where - B c monitor background (cpm) Rmon -monitor response (cpm) offset,+ slope

• X Ci + quad - (z Ci) + B where offset,' slopef nd quad are the coefficients-in a quadratic fit to the monitor response to nuclide activity.

2-7 .C 's o .

II EMS provides an option to calculate nuclide specific responses so that Rmon is the sum of responses for each nuclide, rather than the sum of the nuclide concentrations, as shown above. In the nuclide-specific case, 2 R = Z [offset.+ slope.o C. + quad. (C + B mon 1 1 1 1 where the sum extends over all nuclides which have response factors stored in the database for the monitor of interest. Recommended Setpoint The setpoint recommended for actual use is based on a comparison of the maximum setpoint calculated as above, to setpoints based on expected response time a tolerance factor (to allow for variations in monitor response during release) and to default values determined by the user. The user can restrict which setpoint value is usually reported by what values are used when setting these tolerance factors and default setpoint values. The default setpoint in user units (e.g. cpm). can be defined with or without background included. If the cunitnopt parameter (defined in the release point and discharge point tables) equals 0, the default value does not include background; and the current background is added to the default value to get the reported default setpoint. Otherwise, the current background is not added to the default value. Setpoint in VCi/ml Note: In this version of the software, the reported setpoint is the user units setpoint. The setpoint calculations using the original concentrations (pCifml) is. still being -done by the software and stored in the sampledata table. To get reported setpoints in pCi/ml, the monitor slope should be set to 1.0 in the Release Point transaction of Database Maintenance and the UNITS parameter for the monitor should be set to pCi/ml in the Activity Monitors transaction. If the Isotopic specific response option is turned on for the release point,.then this individual nuclide slopes in the Monitor Slopes can be used to map the response from the nuclide to that of a monitor calibration source (e.g. Cs137 equivalent response) 2-8 P5i

0 , A candidate setpoint is calculated-based on the expected response: S = f x C. exp tol i1 where ft

                  .     = setpoint tolerance factor (can be set for the release point using QBF) 2 if not specified by the user Now compare the Sexp value to the default table value Sdef:

If Seip < Smax and if S < S and S 1 S

      -            exp     def.          def -  Mae then use Sdef                         Case 1 i

• Otherwise use Sexp Case 2,5 If Sexp Smax use Smax Case 3 If Smax °, use Sdef Case 4 Case 4 occurs if no activity is detectable in the-sample (Sadj = 0).

.00. 2-9 p3 52.

0. Case 1 . Case 2 Case 3 Case 4 Case 5 0 Sexp --- Sdef --- Smax---- Smaxa=--- Smax-__ Smax S cxp ---- Sexp Sdef Sdef --- Sdef S.cIp ---- 0 ---- 0 -- e-- 0---- 0 ---- 0 ---- Use Sdef Use Sexp Use S... Use Sdef Us~e Se....p Schematic of Liquid Setpoint Cases Recommended Setpoint in User Units (e.g. cpm) The candidate setpoint based on expected monitor response is calculated as follows: S (cpm)m-f *(R - B) + f *B ex(p P tol mon Btol where fBtol ' background tolerance factor (set using QBF on the

                        .releasept      table)

If the default setpoint value includes background: rp 40a.. 2-10 Pj> r5 2a)

0* . If the default setpoint value does not include background: eBrp = B nd cn where B is the monitor background count rate and B~p is used below. If Sexp (cpm) < Sma. (cpm) and if Sexp (cpm) < Sdef (cpm) + Brp and Sdef (cpm) + BrP then use Sdef (cpm) + Bp Case 1 Otherwise, use Sexp (cpm) Case 2, 5 If Sep (cpm) > S (cpm) is' use S (cpm) ' - Case 3 If Smax (cpm) 0, use Sdef + Brp Case 4*. NOTE: Smax is due to concentration only (i.e.', excludes background) for Case 4 Setpoint for Discharcre Point For the discharge point, the total MPC/ECL fraction is: x C./ZCN ) F + z C. /MPC.)

• F 3.~ ~ 3 ,,00 F +F

                                        *0 OR

( Z C /ECL.) . F. + ( -C./ECL.) .F i . 0 3. 0 i. F 0

                                             + F 2-11 P~5S D~o",-rev ?,Z

II where ( Ci/MPCi )0 =total MPC fraction for existing concurrent releases for this discharge point excluding this additional release. z Ci/WPCi total MPC fraction for the new release (z Ci/ECLi ) total ECL fraction for existing concurrent releases for this discharge point excluding this additional release. z Ci/ECLi total ECL fraction for the new release discharge point waste flow excluding new the release point waste flow to be added. F projected waste flow for the new release point to be added The radiation monitor for the discharge point has setpoint equations identical to those presented above for the release points with the following exceptions:

1. The LOW ACT setpoint equation option is not. supported.

2. For the nuclide-specific response, the concentrations are modified as in:

Cdp C. F/(F + F )]

2. 2 0 R

4pmon (E (offset + slope.. Cdp + quadi

                                            +  ua.

(C4P 2 cd)

                                                                 + R
                                                               ]+dpm where CoP          the discharge point isotope concentration from this release point Rdpmon = the discharge monitor response in user units R      =     the discharge monitor response before the current release is added including the background 2-12                              P9 55 OBXX\Aylrows2-

For non-isotope specific response: Rdpmon =[offset + slope *CdP + quad *(Cd)] + Rdpmon, where CdP l Z Cj] [F/ (F+F;)] Setpoints in PCi/sec-Setpoints in units of yCi/sec can be obtained by setting the UNITS parameter for the monitor to "pCi/s" or "/lCi/sec" (Case sensitive. Ist 5 characters must match) in the Activity Monitors transaction and setting the monitor slope to 1.0 as in'the pCi/ml setpoint calculation. The user units setpoint, as calculated above for the setpoint in pCi/ml units, will be multiplied by the corresponding 'effluent flow rate (release point or discharge point) for the monitor to get a reported setpoint in pCi/sec. 2.6 DOSE CALCULATIONS FOR LIQUID RELEASES The EMS software calculates and stores the dose for each receptor, for-each nuclide, and for each organ. The dose is the total.over all pathways which apply to that receptor. A receptor is defined by receptor ID, age group (infant, child, teen, or adult), sector, and distance from the plant. The equation used in the liquid permit processing to calculate the dose received by receptor r from a released nuclide i is: m Ai. Z Ats Ci F where: The sum extends over all time periods. DiTr the cumulative dose or dose commitment to the total body or an organ T by nuclide i for receptor r from the liquid effluents for the total time period of the release, in mrem. AiTr site-related'ingestion dose or dose commitment factor for receptor r to the total body or organ T for radionuclide i, in mrem/hr per pCi/ml. AiTr is available as an editable table, but can be recalculated with different parameters and pathways with the Dose Factor Processing (DFP) option. The 2-13 Pn5t OL0DC4&Ae g2Q

equations used are presented in Chapter 4 of this manual. N Ats = length of time period s, over which the concentration and F value are averaged, for all liquid releases, in hours. cis the average concentration of radionuclide i in undiluted liquid effluent during time period At 8 from any liquid release, in yCi/ml. Fsr the near. field average dilution factor for receptor r during any liquid effluent release. F F = sr Denom The value of Denom depends upon several variables and nested.if statements. The derivation of the Denom value is shown in the logic and equations shown below. 2-14 mf S I

0 - 0 If the STREAMFLO option in the OPTIONS Table is set to Y, then Denom Fstrm (Uf /60) (U/Rmir) else (river stream'flow is not used) If the denom typ option from the Options Table is 1, (dose from a dilution stream) then Denom = (Uf/60) * (1/RMIX) Else if denomtyp is 2, (dilution flow includes waste flow) then Denom 'a Fdil * (Uf/60) * (l/Ri=) else (denom typ is not 1 or 2) if the QVOPT option in the OPTIONS table is set to ON, (dilution flow is from the QDVOL table) then Denom' (Fw + 'Fvol (Uf/S0) (l/Rmi=) 09 else (the normal standard calculation) Denom - (Fw + Fdil). O(Uf/60) (/Rmix) end of if on QV OPT option end of if on denom typ option end of if on stream flo -option If Denom is greater than 0.0 then If Denom > 1000. and option to limit the denominator is Y, then Denom 1000. end of if denom is too large end ofif"denom is greater than 0.0 P . Denom. Denom / (Uf/60) 2-15 P3I 56

Where: Fst m= River stream flow past the site in user liquid flow rate units. The value used during permit processing is the value obtained from the STATIONDATA table. The'.value is entered into the STATIONDATA table using the QBF-utility. If the value is to be changed often, it would be possible to write a command procedure which get the value from the user and write it into the table. Fw flow rate of undiluted waste effluent in user liquid flow rate units. Uf/60 - Flow rate units conversion factor for liquid releases/60. Uf converts from user units to CFM so this factor converts to CES. Fdil = flow rate of-the dilution flow in user liquid flow rate units. Rmix mixing ratio -fraction of the release that reaches the receptor. Separate mixing ratios are stored for each pathway for each receptor. A mixing, ratio of zero for a pathway receptor indicates that the pathway is not present for the receptor. The first non-zero value is used in the dose calculation. The different mixing ratios for the pathways are incorporated into the composite Ai factors calculated by the dose factor processing (DFP) program. Fqvol= Flow rate from user entered quarterly dilution flow rate. These values are fromi the AFLOW column of the QDVOL table' for the' release. If stream flow option is being used and the average river stream flow is known at the time the liquid release is processed, then the command procedure which runs the liquid permit processing could be' modified to ask for the stream flow value and put it into the stationdata table before the permit is processed. If the average river stream flow is not known at the time the liquid release is processed, then some other provision,must be made for correcting the cumulative dose totals in the CUMDOSE table so that it is based on the correct stream flow value. If the average stream'flow for 2-16

0 the month is used, then each liquid'release point-entry in the CUMDOSE table for the month could be multiplied by the ratio of the actual stream flow for the month divided by the default value contained in the STATIONDATA table. Caution: Since there is no record stored in the database of what stream flow value was used to calculate the dose values, the user must verify that no correction is applied more than -once to each dose value. 2.7 31 DAY PROJECTED DOSE CALCULATIONS The 31 Day Projected Dose values appear on the Standard and Special Permit Reports. The Projected Dose values are calculated as follows: DPT _(DT

• aT Pi where:

DPT =the 31 Day Projected Dose by organ T, by reactor unit DT =the total dose in mrem by organ T, by reactor unit for the quarter containing the release start date from all 0 closed and open releases when an answer of "YV is specified for the "Update Totals" field on the release point definition screen.- p =the Projection Factor which is the result of 31 divided by the number of days from start of the quarter containing the release start date to the end of the release. The quarterly and annual projection values on the standard pre-release report use a projection factor with 92 days or 365 days instead of 31 days in the numerator and do not include the additional anticipated dose term. DaT 'Additional Anticipated Dose for liquid releases by organ T and quarter of release by reactor unit. NOTE: The 31 day dose projections on the Approval/Results screen is the site total for all units.' 2.8 POST-RELEASE PROCESSING After the release is made, actual concentrations are used to check

0. 10CFR20 limits, and the actual dilution flow and waste flow are used instead of the anticipated dilution flow and waste flow.

2-17 P3L"ob

I' For batch releases, the duration is determined from the start and end dates and times, and is-used with the volume input to calculate the release rate. I., Dose calculations are-the same as for the pre-release, but with actual release flow rates and release duration. Setpoint calculations are not performed at the post-release stage. 4. 4* w_, 2-:18 p2b toI

9-0

                                    .CHAPTER 3 GASEOUS RELEASE CALCULATIONS The "annual average X/Q" method is used, in which fixed X/Q and D/Q values are used for each, receptor for all-dose calculations, regardless of actual wind direction and speed prevailing during a given release. Doses are calculated for each receptor location and 0*    age group specified in the Gas Receptors transaction.            The controlling individual is the age group and location which receives the maximum organ dose.

3.1 GAS PRE-RELEASE PERMIT The pre-release permit is produced by a program that uses user-entered estimates of flow rates and release times to calculate doses and activities. The dose rate from the potential release is added to the maximum dose rate occurring for all other releases during the duration of this release for 10CFR20 compliance. The noble gas or air doses and the organ doses are checked against the corresponding limits for 10CFR50 compliance. 3.2 R.ADIONUCLIDE ACTIVITIES AND COMPOSITE VALUES The radionuclide results are read from one set of composite activity database records,-and from three spectrum analysis result files, and saved in an activity array; If a nuclide'appears in more than one spectrum, only the last value read for that nuclide is used. In S.9 case of duplication, the one not desired should be edited out of the nuclide list. The samples are read in the following order: 3-1 c f i_v Z, z -

1. Composite Records

2. Particulate File

3. Radioiodine File

4. Noble Gas File The activity (Qi) and the activity release rate (;i) are calculated for each nuclide i.

Activity Released For the plant stack and turbine building vent:

• Vv
• 28316.85.- UF
• je-6 (pUCi) -(uCi/ml) (cubic feet) (ml/cubic feet) where:

Vv vent release volume in user units (usually FT3 ) Cr.= concentration in uCi/mi UF = the flow-rate units conversion factor which converts from user units to CFM Note: The Ci value also includes the scaled noble gas nuclides for a release (if any exists). The activity release rate in yCi/sec is Ci

• Vf
• 28316.85
• U./60 For containment purge:

Qi = Ci

• pump release rate (CFM)
• 28316..85
• UF/60 Qi Qi
• duration of release (min) 60 3.3 IOCFR20 COMPLIANCE 2The maximum dose rate during the release is determined by summing together the dose rates for this release, with all concurrent releases in the database forzthe time of the release.

3 0 3-2 diL4A. V-

0-0 The database contains all releases for which both pre- and post-release reports have been made (the post-release program enters the data into the cumulative totals). Pre-releases that have not been completed, and which occur during the release under consideration, are also added into the maximum dose rate to account for releases not yet added to the cumulative totals. The three dose rates (whole body, skin, organ) are compared to the old 10CFR20 limits (old and new 10CFR20 are described below) as defined in the Dose Limits transaction in Database Maintenance. The dose rate at or beyond the site boundary due to gaseous effluents from the site is -limited to: (a) Release rate limit for noble gases: Zi.Kshf Zv I (X/Q)V. Q6iv < 500 mRem/yr

• f 1 s OR vshfI [ V.r Q.v] < O mRem/yr
• f 1 f v I alloc s Ei vae 10-0 Elevated Stack a 80m X.. shf (L. + 1.1M.3 z V(X/Q)v. Q.v] < 3000 mRem/yr fa1 cf vy OR I shf Z.(L. (X/Q) + 1.1.) Q.J <.3000 ..

v i i. ir 3.r iv mRem/yrf. .f Elevated Stack 2 80m where the terms are defined below. (b) Release rate limit for all radionuclides and radioactive materials in particulate form, with half lives greater than 8 days: i p v p ip my ] < 1500 mRem/yr falloc f 00 where: 3-3 P3 ivLf~

II i = index over all radionuclides v = index over all vents or stacks for the unit p = index over all pathways r 5 index for receptor locations Ki= the total body dose factor due to gamma emissions for noble gas radionuclide i, in mrem/yr per jICi/m 3 . Li = the skin dose factor due to beta emissions for noble

      -gas radionuclide i, in mrem/yr per yCi/m 3 .

Vir - the elevated plume gamma total body dose factor for nuclide i at receptor location r, in mrem/yr per PCi/sec. Mi= the air dose factor due to gamma emissions for noble gas radionuclide i, in mrad/yr per pCi/m 3. Bir= the elevated plume gamma skin dose factor for nuclide i at receptor location r, in mrad/yr per pCi/sec. - 1.1 = mrad to mrem conversion factor in mrem/mrad Pip the dose factor for the critical organ for nuclides other than noble gases for the inhalation pathway (in units of mrem/yr per pCi/m 3 ) and for ground plane and food pathways (in units of m2 (mrem/yr per pCi/sec)). The most restrictive age group is used. fp = factor to select which pathways are included in the calculation. Factor - 1 to include a pathway, 0 to exclude. Wmv (X/Q) v for tritium and the inhalation pathway and = (D/Q)iv for other nuclides and pathways. (X/Q)vr the highest value of the annual average atmospheric dispersion factor at the site boundary, fori all sectors, in sec/m 3. (4/Q)mv = the highest value of the annual average atmospheric dispersion factor at the distance of the site boundary, for all sectors, in sec/m 3. _ C>Dcv-- fc,."( -&7,

(D/Q)mv = the highest value of the annual average deposition factor at the distance of the site boundary, for all sectors, in m72 ' iv = the average release rate of nuclide i in gaseous effluent from release point v, in pCi/sec. Noble gases may be 'averaged over a period of 1 hour, and any other nuclides may be averaged over a period of 1 week. 500 site dose rate limit for whole body in mrem/year. 3000 site dose rate limit for.skin in mrem/year 1500 site dose rate limit for any organ in mrem/year shf noble gas dose shielding factor fallocw fraction of the dose limit allocated to this release point

          *fs =    safety factor for the release pointf 3.4  SETPOINT DETERMINATION Setpoints are determined from Dose Rate Limits set forth in           the Technical Specifications and stored in the Dose Limits Table.

The ratio of dose rate limit to dose rate for a single release point is given below for these three cases:

 'Noble Gases nratio      rg   lesser of the ratios

} (total body dose'rate limit/total body dose rate) and (skin dose rate limit/skin ddse rate) '

         =  for a vent release, lesser of 500 mrem/yr s*f !   K   Q.   (X/Q) 3-5                              f  Lop v  2.

1L and 3000 mrem/yr shf Z (L. + 1.lM.)

• Q. (X/Q) 1 1 IV my a for an Elevated Stack 2 80m, lesser of 500 mrem/yr U.

shf Z V. Q. and 3000 mrem/yr shf (L * (X/Q) + 1.lBi=J i Radiojodines and Particulates In these cases, the ratio is obtained by summing over the appropriate nuclide indices: 1500 mrem/yr zpratio , maximum organ dose rate z Pi

• Qiv
• Wmv When the sum is over nuclides and the inhalation, ground plane and cow's milk pathways are all turned on.

3.4a SETPOINTS Setpoints are determined for radiation monitors on individual release points, and also for radiation monitors at the discharge points that may combine the effluent from several release points. Calculations for the monitor response are made for noble gases, radioiodines, and particulates. For a release point, the expected monitor response to a given nuclide concentration is: RMon = monitor response (cpm) i B

            - offset + [slope

• vCi] + [quad' (vCi) 2 ] + B 3-6 Ace rev zoo

0 - where offset, slope, and quad are the coefficients in a quadratic fit to the monitor response-to nuclide activity, and B is the monitor background. EMS provides an option to calculate nuclide specific responses so that Rmon is determined from the response for each nuclide, rather than the sum of the nuclide concentrations, as shown above. In that case, - Rmon ( offseti + [slopei

• Cj] + [quadi ( Ci)2]) + B

        *The' expected response for discharge points is based on the sum of the expected response for releases already in progress plus the expected response due to release point being considered.

I = R + X [offset + slope.

• C + quad. * (C]P) I dpmon dpmon 0 i .i 0 . where cip C. * (F /F rp dp

                                              -)

Ci = concentration for the release point

                                        -   I I

e. 3-7 fcej lo-z

I, Frp = flow rate for the release point 4 rdp flow rate for the discharge point Rdpmon discharge point monitor response for the release in progress RdpmonO the discharge monitor response before the current release is added including the background and offseti, slopei and quadi are the quadratic response coefficients of the discharge point monitor. Non-isotope specific response: po = offset + slope* (ZCdP ) + quad CP ) 2 + Rdpmon pon 0 All other equations are the same as for the individual release point, but use the discharge point monitor response and the discharge point allocation factor and.safety factors. EMS allows for setpoint calculations based on the standard or response method. Thus, each release-point will have associated with 40 it, a setpoint equation: STD or RESP. This-can be set in the Release Point (Setpoint) transaction of Database Maintenance. If the release point setpoint equation STD The limiting setpoint for the monitor (in pCi/ml) is given by: Smax ' fs a falloc ratio

• SUM The limiting setpoint for the monitor (in user units, e.g., cpm) is given by:

Ua 'salloc ratio (Rmon B) + B ai-3-.8

                                                                        -    An

0 j where offset = 1. noble gas offset factor

2. radioiodine offset factor

3. particulate offset factor slope = 1. noble gas slope factor

2. radioiodine slope factor

3. particulate slope factor quad = 1. noble gas quadratic factor

2. radioiodine quadratic factor

3. particulate quadratic factor fs =safety factor for the release point falloc dose rate allocation factor for the release point ratio - 1. nratio for-noble gases

2. rpratio for radioiodines

3. rpratio for particulates SUM 1. Z noble gas concentrations, for noble gases

• I 2. Z radioiodine concentrations, for radioiodines

3. X particulate concentrations, for particulates Rmon

• 1. noble gas monitor response

2. radioiodine monitor response

3. particulate monitor response B 1. observed background response for the noble gas monitor

2. observed background response for the radioiodine monitor

3. observed background response for the particulate monitor NOTE :Separate calculations are made for noble gases, radioiodine, and particulates The limiting setpoint for gaseous releases is determined separately for noble gases,- radioiodines, -and particulates for each release point and discharge-point.

• 4)-

3-9 f., I10 r-

II If the release point setpoint equation = RESP The reported setpoint for the monitor (in pCi/ml) now becomes: Smax ' [mrtol * (SUM - B)] + (mrtolb

• B)

The limiting setpoint for the monitor (in user units, e.g., cpm) now becomes: vSUmax>. [mrtol * (Rmon -B)] + (mrtolb

• B) where.

mrtol 1. monitor response tolerance factor (noble gas)

2. monitor response tolerance factor (radioiodine)

3. monitor response tolerance factor (particulate)

SUM as defined above B = as defined above

            .mrtolb         1. monitor tolerance background factor (noble gas)

2. monitor tolerance background factor (radioiodine)

3. monitor tolerance background factor (particulate)

Rmon as defined above 3.4b REPORTED SETPOINTS The setpoint 'reported on the pre-release reports are in user defined units. If the release point setpoint equation is STD, then the maximum setpoint is compared with the response and default setpoints. NOTE :The response setpoint as defined in this section is not necessarily the same as the maximum setpoint based on the RESP setpoint equation, as defined in the previous section. Sresponse is defined below. 3-10 PC,'7{ oj. Ce. ,sL

The reported setpoint is as follows:

1. Reported Sresponse if Sresponse:<Smax. < -Sdefault OR if Sdefault < Sresponse < Smax

2. Reported Smax if Sresponse > Smax

3. Reported Sdefault if Sresponse < Sdefault < Smax where S as defined in the previous section r mrtol

• SUM [pCi/ml]

Sresponse L [mrtol * (Rmon - B)] + (mrtolb

• B) [User Units]

Sdefault normal setpoint defined for the release point in units of [pCi/ml] and [User Units]. NOTE Separate checks are made for each setpoint in [pCi/ml and [User Units] for the noble gas, radioiodine, and particulate monitors. - Setpoints in VCi/sec Setpoints in units of pCi/sec.can be obtained by setting the UNITS. parameter for the monitor to "pci/s" or i'pCi/sec" (Case sensitive. 1st 5 characters must match) in the Activity Monitors transaction and setting the monitor slope to 1.0 -as in the pCi/ml setpoint calculation. The user units setpoint, as calculated above for the setpoint in pCi/ml units, will beimultiplied-by the corresponding effluent flow rate (release point or discharge point) for the monitor to get a reported setpoint in pCi/sec. 3-11 _72-At~C r-,7z

II 3.5 MAXIMUM WASTE FLOW I1 The maximum waste flow calculation is based on what the WFLOWM option (release point setpoint calculation option) is set to. This option can take on one of three values: NONE, DOSE, and CALC. Gaseous release point setpoint WFLOW M can be set to either NONE or DOSE. For gaseous releases, Wmax the minimum of Rwm and Rrwmax where Rwmax -Release point maximum waste flow rate as stored in the release point definition If WFLOW M option - NONE Rcwmax 8 waste flow rate for the sample, Vf If WFLOW M option = DOSE fs-* nratio

• V
• wwmas=

Fwsfac 461., where fs = Safety factor for the release point nratio nratio as described in section 3.4 Vf = Waste flow rate for the release (sample) Fwsfac = Waste flow rate DOSE setpoint safety factor 3.6 DOSE RATE AND CUMULATIVE DOSE CALCULATIONS Noble Gas Dose and Dose-Rate Calculations The dose rate and dose contribution due to noble gases in gaseous effluents are calculated using the following expressions: 6I" 0_, 3-12

• f5'I3 ZqVw r'aizz_

00 For Noble Gas Air Dose due to gamma radiation (for vents or stacks < 80 meters): (3.17

• 10-8) *X/Q*

Vi~ e-

              -D y                                       taa _ f+/-

• For Noble Gas Air Dose due to beta radiation- (for vents or stacks < 80 meters):

Dp = (3.17 p 10-8) X/Q t-b

• f Z N
• Qiv For Nob:Le Gas Total Body Dose Rate (for vents or stacks < 80 meters):

Dt shf X/QgO 87 6 0 a* (xi*.v) For Nob]Le Gas Total Body Dose '(for vents or stacks < 80 meters): shf fo

• 1 (s'- QRiv)
• X/Q
• t-a Dtb (.256 *iO5 /.dur)

For Nobl.e Gas Skin Dose Rate '(for vents or stacks < 80 meters): Ds = shf

• f
• Z QR+/-. * ((Li
• X/Q * -b) + (l.llMi
• X/Qg I*-I 8760 )] -

87 60 For NobI.e Gas Skin Dose (for vents or stacks < 80 meters): shf*;f

• Z.QRi L*
• X/Q - (.b11
• X/Qg t-a)] v Dsk

                               ~~(5.256                  - 105./  dur) where Dp-       total beta air dose from gaseous effluents (mrad)

DY = total gamma air dose from gaseous effluents -(mrad) Dt = the total body dose rate due to gamma emissions by noble gas releases from vent v (mrem/yr) Dtb = total body dose from gaseous effluents (mrem) Ds = skin dose rate' from gaseous effluents (mrem/yr) Dsk ' skin dose from gaseous effluents (mrem) 00 1.11 conversion factor from mrad to mrem 3-13 cOxpA. r-C-V ZI

3.17

• 10-8 = inverse of number of seconds in a year 5.256
• 10=3 number of minutes in a year 87 60 -a= adjustment factor used to convert the 1-hour X/Q value to an average 1 year X/Q value (dimensionless) 8760 - number of hours in a year a "a" factor for gamma noble gas X/Q b " aa" factor for noble gas X/Q t a= adjustment factor to convert the 1-hour X/Q value to the short term X/Q value for the release (dimensionless) t = duration of release (hours) dur- duration of the release. (minutes) fo =occupancy factor defined for the receptor at the given location (dimensionless)

Xi = total body dose factor due to gamma emissions for noble gas radionuclide i (mremfyr per pCi/m 3) Li = skin dose factor due to beta emissions for noble gas radionuclide i (mrem/yr per pCi/m 3) Mi= air dose factor due to gamma emissions for noble gas radionuclide i *(mrad/yr per pCi/m 3 ) Ni= air dose factor due to beta emissions for noble gas radionuclide i. (mrad/yr per pCi/m 3 ) 3.17

• 10-8 D inverse of number of seconds in a year Qiv= release of noble gas radionuclides, i, in gaseous effluents from vent or stack v (pCi)

QRiv = release rate of noble gas radionuclides, i, in gaseous effluents from vent or stack v ( pCi/sec). shf = shielding factor (dimensionless) X/Q = highest value of the noble gas 1-hour X/Q for vent or stack v at the site boundary (sec/m3 ) 3-14 P§n15 O&'*Y4V r-w,z-Z--

X/Qg = highest value of the noble gas 1-hour X/Q for gamma radiation for vent or stack v at the site boundary, (sec/m 3 ) Organ Dose Calculations For Critical Organ Dose Rate--Inhalation Pathway and all Pathways for 9-3, C-14 (for vents or stacks < 80 meters): DRTa X/Qr

• 8760 ' Z PipTa
• QRi For Critical Organ Dose Rate--Ground and Food Pathways (for vents or stacks < 80 meters):

DRTa D/Q 8 7 6 0 -d -

• RipTa* QRv For Critical Organ Dose-Inhalation Pathway and all Pathways for H-3, C-14 (for vents or stacks < 80 meters):

Dta (3.17 10-8) X/Qr tc* fo C) PipTa iv For Critical Organ Dose-Ground and Food Pathways (for vents or stacks < 80 meters): DTa (3.17

• 10-8) .D/Q
• t ' f0 '.. Ripr* Qiv where DRTa ' dose rate for age group a and organ T from iodines and particulates with half lives greater than 8 days in gaseous effluents (mrem/yr)

Dta dose for age group a and organ T from iodines and particulates with half lives greater than 8 days in gaseous effluents (mrem) c = "a" factor for Radioiodine/Particulate X/Q d - "a" factor for D/Q D/QO highest value of the 1-hour deposition factor at the distance of the site boundary (1/m 2 ) PipTa = dose factor for each radionuclide i, pathway p, organ T, and age group a (mrem/yr per pCi/m 3) RipTa = dose factor for each radionuclide i, pathway p, organ T, and age group a (m2

• mrem/yr per pCi/sec)

W ~3-15

        ;Zc                                                                      r97z,

X/Qr = highest value of the radioiodine/pirticulate 1-hour X/Q for vent or stack v at the site boundary (sec/m3) Note: It is assumed PipTa will not contain long term X/Q or D/Q values. The maximum exposed individual is determined by the maximum dose received by any organ. The summation extends over all applicable nuclides and pathways. 3.7 RESOLVING DOUBLE-COUNTING'OF DOSE AND ACTIVITY Gaseous release points fall into three categories for double-counting of dose and activity. One, a release point will not have activity sampled twice. Two, a release point can have activity that is sampled again downstream and would be double-counted if no corrections were applied. Three, a release point can have samples containing activity already sampled once upstream which would be double-counted if no corrections were applied. The last two categories can be called the "CAUSE" release point and the "EFFECT" release point, respectively. To avoid double-counting dose and' activity, only the "EFFECT" release point will have its activity and concentrations corrected as follows. Corrected activity is calculated as follows: Acei =Ae. -Aci where: Acei =the corrected ;'EFFECT" release point activity for nuclide i which defaults to zero if its value is less than zero. Aei =the initial nEFFECT" release point activity for nuclide ei Aci -the "CAUSE" release point activity for nuclide i Corrected concentrations are' calculated as follows: Ccei = (Acei / Ve)

• 35.315 3-16 O9,DCio nsjzz,

where: Ccei -the corrected "EFFECT" release point concentrations for nuclide i Ve =the waste volume for the "EFFECT" release point 35.315 -conversion factor from Ci/ft3 to pCi/mi (Ci/ft 3 ft3 /1728 .in3

• in3 /16.387 cm3 )

3.8 31 DAY PROJECTED DOSE CAtCULATIONS The 31 Day Projected Dose values appear on the Standard and Special Permit Reports. The Projected-Dose values are calculated as follows: DpT -(DT

• P) + DaT where:

DpT -the 31 Day Projected Dose .by organ T, by reactor unit DT -the total dose in mrem 'by organ T, by reactor, unit for the quarter containing -the release start date from all closed and open releases when, an answer of "Y" is specified for the"Update Totals" field on the release point definition screen. p -the Projection Factor which is the result of 31 divided by the number of days from the start of the quarter to the end of the release. The quarterly and annual projection values on the standard pre-release report use a projection factor with 92 days or 365 days instead of 31 days in the numerator and do not include the additional anticipated dose term. DaT Additional Anticipated Dose for gaseous releases by organ T and quarter of release, by reactor unit. NOTE: The 31 day dose projections on the Approval/Results screen is the site total for all units. Se" 3-17

                                                                               '7

3.9 GAS POST-RELEASE PROCESSING After a pre-release permit has been approved, the post-release program is run to: o Enter actual release start and stop times, flow rates, etc. o Check 10CFR20 limits o Check 10CFR50 limits o Add the dose and activity data into the cumulative totals. Compliance with 10CFR20 limits is checked in the same way as described for the pre-release program. Dose rates are calculated and compared to lOCPR20 limits. Monitor setpoints are not calculated at the post release stage.

                                                                            .I*

3-18 p, 79 0DrCK_ &v1 2-

CHAPTER 4 LIQUID DOSE FACTOR EQUATIONS The DFP optionjis used to calculate the liquid dose factors described previously. Dose factors are calculated separately for each nuclide, organ, and age -group. The'age group, applied to a specific receptor's dose calculations,-. is part of the receptor specification. 0 0- For a particular receptor, the total dose factor (AiT*r) over each pathway p with its specific mixing ratio: is a sum 1I __ 1 m

                                     ,      A. .

A. mixrp 17 ,r,p R ixr - where Ai-r- p =the dose factor for nuclide i, organ T, receptor age group r, and pathway p - I RmiX r p - mixing ratio for the pathway Rmxr mixing.ratio for the receptor, which is the first non-zero value of vix r p encountered during the calculation . The user specifies which pathways are included by setting the mixing ratios for the pathways desired to the correct non-ze6ro -value. If the receptor mixing ratio for a given pathway is zero, that term is not included in the sum. I. 4-1 OCA' revz2

II The DFP option of EMS uses a more expanded form for iicuid dose factors than is given in NUREG-0133. These equations are taken from R.G. 1.109, and account for nuclide decay as well as shoreline 6* doses. If desired, parameters may be selected to reduce the calculations to match NUREG-0133 exactly. Four different forms of equations are used for the dose factors. 4.1 POTABLE WATER The dose factor for potable water is: AiTjorip ' ko * (UP /dW)

• Ni *.DFiT,r
• e(-Aitp) where AiT dose parameter for organ T. for the receptor age

                  -rp group r, for nuclide i, due to exposure pathway p, in mrem/hr per pCi/ml ko =        units conversion factor, = 1.142E5 =E6(pCi/pCi)

• 1000 (ml/Kg)/ 8760 hr/yr Urrp usage factor for pathway p and age group r dw - additional dilution factor for potable water Ni = fraction of the radionuclide activity released to the water discharge path that reaches a specific receptor.

DFiTr =1 ingestion.dose conversion factor for nuclide i for receptor age group r in organ T, in mrem/pCi (Tables E-7 to E-11 of R.G. 1.109) Ai so decay constant for nuclide i tp = average transit time in seconds 4.2 AQUATIC FOODS PATHWAYS The liquid dose factor is AiT.rP= ko

• Urp
• BF.'p
• N.
• DFiT. r
• e2p(-AltP) 4-2 I5 at

                                                                         &'V &8l

where BF4 rp bioaccumulation factor for pathway p and nuclide i (from Reg. Guide 1.109, Table A-1). Other variables are as defined on the previous page. 4.3 SHORELINE RECREATION PATHWAY - The pathway-specific dose factors for shoreline deposition are given by:-

                                                    -At 1I   e   i b      -. A  t A.            k    W   Ni. U-                        e     i   sd DFG.

iT,.r,p s s z. f ~r ,pI where Ws shoreline width factor ks conversion factor ko

• kc
• mtv/3600 kc water to sediment transfer coefficient in L/kg hr 2

mtv - Mass density of sediment in kg/m , 40 2 kg/m 3600. A- Seoconds per hour units conversion factor tb = length of.time sediment is exposed to contaminated water, 4.716E8 sec tsd ' transit time to, deposit activity on shoreline DFGiT - the dose conversion factor for standing on ground contaminated with nuclide i, in mrem/hr per pCi/m2 4-3

4.4 IRRIGATED VEGETABLE PATHWAY 00".' 5 A = 1.14

• 10
• U CF. DF .

iT,rp f,rp iv LT rr where: 1.14

• 105 > a units conversion factor CFiv = the concentration factor for radionuclide i in irrigated vegetables, as applicable to the vicinity of the plant site (pCi/kg) / (pCi/L).

Calculation of the Concentration Factor The calculation of the concentration .factor for radionuclide i in irrigated vegetables, CFiv as used in the equation for A1T. is calculated as follows for all radionuclides other than Tritium:

                                     -A ibt CFiv c N. M

• I ( - Ei e) fI iv( 2.

[ l.-1YA - P A. e t L v Ei 2. For Tritium, the equation is as follows: CFiv = Ni

• M
• Lv where M the additional dilution factor from the near field of the discharge structure to the point of irrigation water usage.

I the average irrigation rate during the growing

• season (L/m2 h).

r the fraction o~f irrigation-deposited activity retained on the edible portions of leafy vegetables. There are separate values available for radioiodines and particulates. Yv = the agricultural productivity of irrigated leafy vegetables (kg/m 2 ). 00. 4-4 Obcm rev 2h2i

fI = the fraction of the year that vegetables are irrigated. Biv = the crop to soil concentration factor applicable to radionuclide i. (pCi/kg vegetables)/(pCi/kg soil). P the effective surface density of soil (kg/m2 ). Ai the decay constant for radionuclide i (h- 1 ). AEi the effective removal rate for activity. deposited on crop leaves (h- 1 ), calculated as A 'i = iA; F ' 'w Aw = the rate constant for removal of activity from plant leaves by weathering (h 1). te 'the period of leafy vegetable exposure during the growing season (h). tb the period of long-term buildup of activity in soil (h). , th the time between harvest of'vegetable and human consumption (h). Lv the water-content of leafy vegetable edible parts (L/kg). 4.5 REDUCTION TO NUREG-0133 EQUATIONS NUREG-0133 does not have shoreline deposit equations, which can be eliminated by setting the Water Recreation Mixing Ratio to zero in the Liquid Receptor Transaction definition under EMS. For the other equations, reduction to NUREG-0133 is obtained by setting: Ni c 1 (this can be set in the definition of Fracti6n of Activity Reaching Receptor in DFP) average transit -time:t = 0 (this can be set in the definition of Dose Calculation'Parameters in:DFP) 4-5 p.8i 0_f. '2

                                                                       . 2.

CHAPTER 5 GAS DOSE FACTOR CALCULATIONS The DFP option is used to calculate the gas dose factors described previously. Dose factors are calculated-separately for each nuclide, organ, and age group. The age group, applied to a specific receptor's dose calculations, is part of the receptor specification. The same gas dose factors are used for both the site boundary dose rate calculations and for the maximum individual controlling location dose calculation. The dose factor for each particulate or iodine nuclide i (or tritium) is given-below. It is a function of pathway, organ, and age group. The pathways considered are:

1. Inhalation

2. Ground

3. Milk (Cow or Goat)

4. Meat

5. Vegetable 5.1 INHALATION PATHWAY PiTa K' (BR)a (DFAiT)a (r=em/yr per pCi/m 3 )

K' 1E6 pCi/PCi 5-1

                                                             .pa85 C)ViS77

(BR)a = breathing rate for age group a, in cubic m/yr (DFAiT)a = inhalation dose factor for organ T, for age group a, for nuclide i, in mrem/pCi 5.2 GROUND PLANE PATHWAY RiT.a K'K" (SF) DFGiTr (1 - e Ait) Ai ] (m2 -mrem/yr per pCi/sec). where Kr' = E6 pCi/PCi K" =8760 hr/yr A = decay constant for nuclide i, in sec' t exposure time (sec) = 4.73E8 (15 years) DFGit= ground plane conversion factor for nuclide i, organ T (The s'ame DFGIT factors apply to all age groups. The factors labelled total body in the database are applied to all other organs) SF shielding factor 5.3 MILK PATHWAY Rila X' (DFLiT)a e I Fmi UuQF

                           -a     :: .ap        ,-
                       -(A.+A )t                               -

r (1-e i we) (1 -e i b) (f-ei A + Biv p i W - 5-2

r (1e- (Ai+ A ) te) 11 (1 -eitb) + (1-f f )e i h + B. P S. i.v Ys (Ai + AW) p Ai 2 (m. - mrem/yr per pCi/sec) where K' - 1E6 pci/yci

      'OF            feed consumption rate by the milk animal (cow or goat)

(Kg/day) Ua= age group a milk consumption (cow or goat) Yp agricultural productivity by unit area of pasture feed grass, in Kg/sq. m, Ys = agricultural productivity by unit area of stored feed, in Kg/sq. m Fmi =stable element transfer coefficient for nuclide i, from feed to milk, in days/liter Biv factor for uptake of radionuclides from soil by crops r - fraction of deposited activity retained on animal feed grass (cow or milk). Separate values are used for radioiodines than all other particulates. (DFLiT)a ingestion dose 'factor for organ T. for nuclide i, for receptor in age group a, in mrem/pCi decay constant for nuclide i 5= decay constant for removal of activity on leaf and plant surfaces by weathering in sec- 1 tf = transport time from pasture to cow or goat to milk to receptor,-in sec. th = transport time from pasture to harvest to cow or goat to milk to receptor, in sec. te = seasonal crop exposure time, in sec. fp = fraction of year that p3 animal is on pasture 5-3 - - r__ a3 b7 Z COtCKVA Irk 72-2

f5 fraction of animal feed that is pasture grass while animal is on pasture Is Carbon-14 in Milk

                                        .-.                    A RiTa       K'K"' FJmijF  Uap (DFLiT)a PC (0.11/0.16) e' itf (m2-mrem/yr per pCi/sec) where K"'        1E3 gm/Kg PC  =

fractional equilibrium ratio 0.11 fraction of total plant mass that is natural carbon 0.16 c concentration of natural carbon in the atmosphere (g/m3) and all other parameters as defined above 6-Only OF and Vap depend on cow or goat. Tritium in Milk RiTa K'X"' Fni QFOUap (DFLiT)a (0.75) (0 ) t (m2-mrem/yr per pCi/sec) where K"' =. 1E3 gm/Kg E absolute humidity, gm/cubic meter 0.75 fraction of total feed that is water 0.5 = ratio of specific activity of feed grass water to the atmospheric water and all other parameters as defined above 64 5-4 D-rx:At few 2 Only QF and Uap depend on cow or goat. 5.4 MEAT PATHWAY RiTa = ' (DFLiT)a eAitf OF Ffi Uap r (l-e

                  -(A+A )t i w e)             l-e
                                             -At i b L     Y (A+)                                biv]

p i W pA r lt e l-(Ai+A )te 1 -Aitb + (1-f f )e i h + B.- P S Y (A + A) v p A. where Ffi = stable element transfer coefficient for nuclide i, from feed to meat, in days/Kg Uap = receptor's meat consumption (Kg/yr) th transport time from crop field to receptor, in sec tf = transport time from pasture to receptor, in sec and all other-factors are as described for the cow-milk pathway Carbon-14 in Meat RiTa - K'"' Ff F Uap (DFLiT)a Pc (0.11/0.16) e-Aitf (m2 -mrem/yr per pCi/sec) where all terms are as defined above. 5-5 py 8 Pi sfL

Tritium in Meat RiTa KK Ffi QF Uap (DFLiT)a (0.75) (0.5/H) e"itf 0 (m2_ m-nrem/yr per pCi/sec) where all terms are as-defined.above. 5.5 VEGETABLE PATHWAY iTa (DFLiT)a Ta fL eL itL rr (A.+A t -A t

                                                               ]

(l-e I ~w te) Biv (1!-e i b)1

                .Y   (A+    A)        +      -    PA
           +U        f e Jits a ,g

[r (I-e-(Ai+Aw)t B iv

                                                     -A t (i-e i b)

I* I II YST (Ai+ AW) p . .

                                             ~  '"          -

(m2 mrem/yr per yCi/sec) where a consumption rate of fresh leafy vegetation for age group a, in Kg/yr a consumption rate of stored vegetation for age' group, a, in Kg/yr us= fL= fraction of annual-intake of leafy vegetation grown locally fg fraction of annual intake of stored vegetation grown locally 5-6 tfC-,DUA reV 2 tL - average time between harvest of leafy vegetation and consumption, in sec. ts = average time between harvest of stored vegetation and consumption, in. sec. tb = long term sediment exposure time, in sec. te = seasonal crop exposure time, in sec. Yv = vegetation areal density, in Kg/m2 Ysv stored vegetation areal density, in KG/r2 p effective soil surface density B.V soil to vegetation transfer factor for nuclide i All other factors are as defined above. Carbon-14 in Vegetables RiTa KK L + US) (DFLir)a Pc (0.11/0.16) e-Aitf (la (m2 -mrem/yr per pCi/sec) where all variables are as defined earlier. Tritium in Vegetables Riea K'K"' (U US) (DFLiy)a (0.75) (0.5/H) e-Aitf (m2-mrem/yr per juCi/sec) where all variables are as defined earlier. 5.6 REDUCTION TO NUREG-0133 EQUATIONS Inhalation and ground plane pathways are the same in R.G. 1.109 and NUREG-0133. For the other pathways (milk, meat, and vegetable), these equations reduce to the NUREG-0133 values by setting: tb = 0 5-7 ckcJ4L rt zy

e1 te = 9.999E19 tf = 0 (in tritium equations only) NUREG-0133, which can be obtained by There are no C-14 equations in setting pc i t. 0- - - I 5-8 na92-PZCAA, 'rcv2 ) APPENDIX A REFERENCES

1. Boegli, J.S. R.R. Bellamy, W.L. Britz, and R.L.

Waterfield, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants, "NUREG-0133" (October 1978).

2. Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I, U.S. NRC Regulatory Guide 1.109, Rev. 1 (October 1977).

0* ; A-1 Cp93 as Vev2 2L}}