Text

Offsite Dose Calculation Manual, Revision 36, Program Manual
	ML12123A040
Person / Time
Site:	Seabrook
Issue date:	04/26/2012
From:	Robinson D; NextEra Energy Seabrook
To:	; Office of Nuclear Reactor Regulation
References
	SBK-L-12097
	Download: ML12123A040 (421)
	v • d • e

RMD Controlled Copy SEABROOK STATION PROGRAM MANUAL Offsite Dose Calculation Manual ODCM Rev. 36 Manual Owner: 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 REMP 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.Revisions to the ODCM require an interdisciplinary review documented on the following page as well as a SORC review prior to implementing the change. An interdisciplinary review includes as a minimum the potential impact the change has on the respective departments' programs and procedures.

The interdisciplinary review shall involve the following departments:

Chemistry, Radiation Protection, I&C, and Operations.

The Originator shall ensure the interdisciplinary reviews are completed prior to the SORC meeting.Page I of 2 ODCM Rev. 30 1 Interdisciplinary Review ODCM Rev. No.Sheet of Review Organization Reviewed By/Date: Originator Date Date Chemistry Manager Page 2 of 2 ODCM Rev. 30 OFFSITE DOSE CALCULATION MANUAL (ODCM)TABLE OF CONTENTS CONTENT PAGE PART A: RADIOLOGICAL EFFLUENT CONTROL AND ENVIRONMENTAL MONITORING PROGRAMS

1.0 INTRODUCTION

A.1-1 2.0 RESPONSIBILITIES (PART A) A.2-1 3.0 DEFINITIONS 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.5-11 6.0 RADIOACTIVE LIQUID EFFLUENTS A.6-1 6.1 Concentration A.6-1 6.2 Dose A.6-9 6.3 Liquid Radwaste Treatment System A.6-11 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 Plant Operations Monitoring Program A.9-1 9.2 Land Use Census A.9-I1 Page I ODCM Rev. 36 CONTENT PAGE 9.3 Interlaboratory Comparison Program A.9-13 9.4 Dry Fuel Storage Facility Monitoring Program A.9-14 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

1.1 Responsibilities for Part B 1.2 Summary of Methods, Dose Factors, Limits, Constants, Variables and Definitions 2.0 METHOD TO CALCULATE OFF-SITE LIQUID CONCENTRATIONS 2.1 Method to Determine F 1 ENG and C1NG 2.2 Method to Determine Radionuclide Concentration for Each Liquid Effluent Source 2.2.1 Waste Test Tanks 2.2.2 Turbine Building Sump 2.2.3 Steam Generator Blowdown Flash Tank 2.2.4 Primary Component Cooling Water (PCCW) System 2.2.5 Water Treatment System (Condensate Polishing System)3.0 OFF-SITE DOSE CALCULATION METHODS 3.1 Introductory Concepts 3.2 Method to Calculate the Total Body Dose from Liquid Releases 3.2.1 Method I 3.2.2 Method 1I 3.3 Method to Calculate Maximum Organ Dose from Liquid Releases 3.3.1 Method I 3.3.2 Method II 3.4 Method to Calculate the Total Body Dose Rate from Noble Gases 3.4.1 Method I Page 2 B.1-1 B.l-1 B.1-2 B.2-1 B.2-1 B.2-2 B.2-2 B.2-3 B.2-3 B.2-3 B.2-3 B.3-1 B.3-2 B.3-4 B.3-4 B.3-5 B.3-6 B.3-6 B.3-7 B.3-8 B.3-8 ODCM Rev. 36 CONTENT PAGE 3.4.2 Method II 3.5 Method to Calculate the Skin Dose Rate from Noble Gases 3.5.1 Method I 3.5.2 Method II 3.6 Method to Calculate the Critical Organ Dose Rate from Iodines, Tritium and Particulates with T 1/2 Greater than 8 Days 3.6.1 Method I 3.6.2 Method II 3.7 Method to Calculate the Gamma Air Dose from Noble Gases 3.7.1 Method I 3.7.2 Method II 3.8 Method to Calculate the Beta Air Dose from Noble Gases 3.8.1 Method I 3.8.2 Method H 3.9 Method to Calculate the Critical Organ Dose from Jodines, Tritium and Particulates 3.9.1 Method I 3.9.2 Method II 3.10 Method to Calculate Direct Dose from Site Operations 3.10.1 Method 3.11 Dose Projections 3.11.1 Liquid Dose Projections 3.11.2 Gaseous Dose Projections 3.12 Method to Calculate Total Dose from Plant Operations 3.12.1 Method 4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 4.1 Plant Operations REMP 4.2 Dry Fuel Storage Facility Monitoring Program 5.0 SETPOINT DETERMINATIONS B.3-10 B.3-11 B.3-11 B.3-14 B.3-15 B.3-15 B.3-18 B.3-19 B.3-19 B.3-21 B.3-22 B.3-22 B.3-24 B.3-25 B1.3-25 B.3-27 B.3-28 B.3-28 B.3-30 B.3-30 B.3-31 B.3-35 B.3-35 B.4-1 B.4-1 B.4-1 B.5-1 ODCM Rev. 36 Page 3 CONTENT 5.1 Liquid Effluent Instrumentation Setpoints 5.1.1 Liquid Waste Test Tank Monitor (RM-6509)5.1.1.1 Method to Determine the Setpoint of the Liquid Waste Test Tank Monitor (RM-6509)5.1.1.2 Liquid Waste Test Tank Monitor Setpoint Example 5.1.2 5.1.3 5.1.4 5.1.5 5.1.6 Turbine Building Drains Liquid Effluent Monitor (RM-6521)Steam Generator Blowdown Liquid Sample Monitor (RM-6519)PCCW Head Tank Rate-of-Change Alarm Setpoint PCCW Radiation Monitor Water Treatment Liquid Effluent (CPS Rad Monitor RM-6473)5.2 Gaseous Effluent Instrumentation Setpoints 5.2.1 Plant Vent Wide-Range Gas Monitors (RM-6528-1, 2 and 3)5.2.1.1 Method to Determine the Setpoint of the Plant Vent Wide Range Gas Monitors (RM-6528-1, 2 and 3)5.2.1.2 Plant Vent Wide Range Gas Monitor Setpoint Example for Limiting Case 5.2.2 Waste Gas System Monitors (RM-6504 and RM-6503)5.2.3 Main Condenser Air Evacuation Monitor (RM-6505)6.0 LIQUID AND GASEOUS EFFLUENT STREAMS, RADIATION MONITORS AND RADWASTE TREATMENT SYSTEMS 7.0 BASES FOR DOSE CALCULATION METHODS 7.1 Liquid Release Dose Calculations 7.1.1 Dose to the Total Body 7.1.2 Dose to the Critical Organ 7.2 Gaseous Release Dose Calculations PAGE B.5-1 B.5-1 B.5-1 B.5-5 B.5-6 B.5-7 B.5-8 B.5-9 B.5-9 B.5-11 B.5-11 B.5-11 B.5-13 B.5-16 B.5-16 B.6-1 B.7-1 B.7-1 B.7-4 B.7-4 B.7-7 7.2.1 Total Body Dose Rate from Noble Gases 7.2.2 Skin Dose Rate from Noble Gases 7.2.3 Critical Organ Dose Rate from lodines, Tritium and Particulates with Half-Lives Greater Than Eight Days 7.2.4 Gamma Dose to Air from Noble Gases 7.2.5 Beta Dose to Air from Noble Gases 7.2.6 Dose to Critical Organ from lodines, Tritium and Particulates with Half-Lives Greater Than Eight Days 7.2.7 Special Receptor Gaseous Release Dose Calculations Page 4 B.7-7 B.7-8 B.7-11 B.7-13 B.7-14 B.7-16 B.7-18 ODCM Rev. 36 CONTENT PAGE 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 7.2.7.3 Critical Organ Dose Rate from lodines, 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 Range Gas Monitor Setpoints B.8-5 8.3 Basis for PCCW Head Tank Rate-of-Change Alarm Setpoint B.8-10 8.4 Basis for Waste Gas Processing System Monitors (RM-6504 and RM-6503) B.8-11 8.5 Basis for the Main Condenser Air Evacuation Monitor Setpoint (RM-6505)

B.8-13 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 (Hogging Mode) B.8-15 References R-1 Appendix A: Dose Conversion Factors A-1 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 ODCM Rev. 36 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.5-8 A.5.2-1 Radioactive Gaseous Effluent Monitoring Instrumentation A.5-17 A.5.2-2 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements A.5-19 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 A.9.4-1 Dry Fuel Storage Facility Radiological Environmental Monitoring Program A.9-15 PART B TABLES B.1-1 Summary of Radiological Effluent Part A Controls and Implementing Equations B. 1-3 B. 1-2 Summary of Method I Equations to Calculate Unrestricted Area Liquid Concentrations B. 1-6 B. 1-3 Summary of Method I Equations to Calculate Off-Site Doses from Liquid Releases B. 1-7 B. l-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-l1 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.l-8 Summary of Variables B.l-15 Page 6 ODCM Rev. 36 CONTENT PAGE B.1-9 Definition of Terms B.1-22 B.1-10 Dose Factors Specific for Seabrook Station for Noble Gas Releases B. 1-23 B.1-1 I 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 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.4-2 Dry Fuel Storage Radiological Environmental Monitoring Stations B.4-6 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.7-4 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.4-7 B.4-2 Radiological Environmental Monitoring Locations Between 4 Kilometers and 12 Kilometers from Seabrook Station B.4-8 B.4-3 Radiological Environmental Monitoring Locations Outside 12 Kilometers of Seabrook Station B.4-9 B.4-4 Direct Radiation Monitoring Locations within 4 Kilometers Page 7 ODCM Rev. 36 CONTENT PAGE of Seabrook Station B.4-10 B.4-5 Direct Radiation Monitoring Locations Between 4 Kilometers and 12 Kilometers from Seabrook Station B.4-1 1 B.4-6 Direct Radiation Monitoring Locations Outside 12 Kilometers of Seabrook Station B.4-12 B.4-7 Dry Fuel Storage Radiological Environmental Monitoring Stations B.4-13 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. 36 LIST OF EFFECTIWE PAGES PAGE REV.PAGE REV.PAGE REV.Cover Abstract 1 &2 TOC 1 -8 LOEP I A.1-1 A.2-1 A.3-1 A.4-1 thru A.4-5 A.5-1 thru A.5-20 A.6-1 thru A.6-12 A.7-1 thru A.7-12 A.8-1 and A.8-2 A.9-1 thru A.9-15 A.10-1 thru A.10-3 B. 1-0 thru B. 1-28 B.2-1 thru B.2-3 B.3-1 thru B.3-35 B.4-1 thru B.4-13 B.5-1 thru B.5-19 B.6-1 thru B.6-3 B.7-1 thru B.7-35 B.8-1 thru B.8-18 R-1 36 30 36 36 33 24 24 32 34 33 25 32 35 29 33 32 32 34 33 28 24 28 21 A-1 thur A-15 B-1 thru B-52 C-1 C-2 C-3 C-4 C-5 C-6 21 25 28 28 16 16 16 28 Page I ODCM Rev. 36 PART A RADIOLOGICAL EFFLUENT CONTROL AND ENVRONMENTAL 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.An ODCM Surveillance Requirement may be considered met if surveillance is performed within 1.25 times the stated surveillance interval.

This 25% extension facilitates surveillance scheduling and considers plant operating conditions that may not be suitable for conducting the surveillance test. This provision is not intended to be used repeatedly merely as an operational convenience to extend surveillance intervals beyond those specified.

A.1-1 ODCM Rev. 33 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 3.0 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 implementation 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 Alarm/Trip Setpoints set to ensure that the limits of Control C.6.1.1 are not exceeded.

The Alarm/Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM), Part B.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 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).C.6.1.1 -RADIOACTIVE LIQUID EFFLUENTS

-CONCENTRATION CONTROL -At All Times 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 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 microCurie/ml total activity.A.4-1 ODCM Rev. 32 1 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 31-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-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.A.4-2 ODCM Rev. 32 1 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, 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:

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* 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.A.4-3 ODCM Rev. 32 1 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 -1.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 km (5 miles) the location in each of the 16 meteorological sectors of the nearest milk animal, the nearest residence, and the nearest garden** of greater than 50 m 2 (500 ft 2) 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. 32 1 C.9.4.1 DRY FUEL STORAGE FACILITY MONITORING PROGRAM CONTROL -At All Times The Dry Fuel Storage Facility Radiological Environmental Monitoring Program shall be conducted as specified in Table A.9.4-1.A.4-5 ODCM Rev. 32 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 channels 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 a time period determined by an evaluation conducted in accordance with the requirements of the Corrective Action Program. An evaluation is not required if the noncompliance is a consequence of surveillance testing or planned maintenance.

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 alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.A.5-1 ODCM Rev. 34 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., 2.b. and 2.d.) is required "at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days when continuous, periodic, or batch releases are made." Additionally, ACTION 31 of Table A.5.1-1 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 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. 34 The following actions are required when the Service Water side 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 samples 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 system with the Service Water system drained and the PCCW system in operation shall continue per the requirements of Technical 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 (1-RM-6515 and 1-RM-6516) sense radiation in the 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 system, 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, by RDMS data base manipulation to verify alarm/trip functions, does not prevent the channel from continuously monitoring radiation levels (except for the WRGM DCOT due to low and high activity sample flow paths). 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.

A.5-3 ODCM Rev. 34 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.5.1-1 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.5.1-1 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.During power operations, SGBD demineralizer effluent is normally aligned to the Main Condenser Hotwell for water recovery.

However, following SGBD / SGBD demineralizer system maintenance, or for plant operational requirements, the system effluent can be aligned to the Turbine Building Sump to maintain secondary chemistry and plant related parameters.

When SGBD demineralizer effluent is aligned to the Turbine Building Sump, the effluent to Outfall 001 is monitored by RM-652 1. The Turbine Sump Radiation Monitor RM-6521, although in the flow path of the SGBD demineralizer effluent, cannot perform the alarm and trip function of the SGBD Flash Tank Monitor RM-6519 because it cannot achieve the same sensitivity to radiation.

However, RM-6521 shall have its set points established per plant procedures since the discharge flow path is through the SGBD demineralizers (where a potential to acquire radioactivity exist), but after passing through RM-6519. If either RM-6519 or RM-6521 is inoperable, then the periodic sampling requirements of Table A.5.1-1 Action 30 apply. It should be noted that when SGBD demineralizer effluent is aligned to the Turbine Building Sump, flow indicator 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 DF-FT-5957) so that Table A.5.1-1 ACTION 31 does NOT have to be met.Steam generators may be drained using the wet lay-up pumps directly to the circulating water system, if no secondary steam pressure is available and provided the steam generator(s) liquid radioactivity is less than both 10 CFR 20 (Appendix B, Table 2, Column 2 instantaneous release);

and 10 CFR 50 Appendix I (annual release).

Table A.5.1-1 Actions 30 and 31 apply in this case as both the radiation monitor (RM-RM-6519) and the flow rate monitor (SB-FE-1918) are bypassed.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. 34 TABLE A.5.1-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS OPERABLE ACTION INSTRUMENT

1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release a. Liquid Radwaste Test Tank Discharge b. Steam Generator Blowdown Flash Tank Drain c. Turbine Building Sump Effluent Line d. Water Treatment Liquid Effluent Discharge 1 1*1 1 29 30 30 29 31 31 2. Flow Rate Measurement Devices a. Liquid Radwaste Test Tank Discharge b. Steam Generator Blowdown Flash Tank Drain 1 1*c. Circulating Water Discharge N.A.d. Water Treatment Liquid Effluent Discharge 1 31*Only applicable when steam generator blowdown is directed to the discharge transition structure without intermediate collection.

The required radiation monitoring channel is RM-6519. The flow path must include a flow indicator which can be used to provide total flow discharged during period of interest.**Pump performance curves generated in place should be used to estimate flow.A.5-5 ODCM Rev. 34 TABLE A.5.1-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION (Continued)

MIMMUM ACTION CHANNELS INSTRUMENT OPERABLE 3. Radioactivity Monitors Providing Alarm but Not Termination of Release a. Primary Component Cooling Water System (in lieu of 32 service water monitors)4. Rate of Change Monitor a. Primary Component Cooling Water System Head Tank 33 (in lieu of service water monitors)A.5-6 ODCM Rev. 34 TABLE A.5.1-1 (Continued)

ACTION STATEMENTS ACTION 29 -ACTION 30 -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.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 10-7 microCurie/ml

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the specific activity of the secondary coolant is greater than 0.01 microCurie/gram DOSE EQUIVALENT 1- 13 1, or b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of the secondary coolant is less than or equal to 0.01 microCurie/gram DOSE EQUIVALENT 1-131.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 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases.

Pump performance curves generated in place may be used to estimate flow.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 analyzed for radioactivity until the inoperable channel(s) is restored to OPERABLE status.With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the radioactivity level is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during actual releases.ACTION 31 -ACTION 32 -ACTION 33 -A.5-7 ODCM Rev. 34 TABLE A.5.1-2 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHECK SOURCE CHECK CHANNEL CALIBRATION CHANNEL OPERATIONAL TEST INSTRUMENT

1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release a. Liquid Radwaste Test Tank Discharge b. Steam Generator Blowdown Flash Tank Drain c. Turbine Building Sumps Effluent Line d. Water Treatment Liquid Effluent Discharge 2. Flow Rate Measurement Devices a. Liquid Radwaste Test Tank Discharge*

b. Steam Generator Blowdown Flash Tank Drain***c. Circulating Water Discharge

- d. Water Treatment Liquid Effluent Discharge D D D D D(3)D(3)D(3)D(3)P M M P N.A.N.A.N.A.N.A.R(2)R(2)R(2)R(2)R R N.A.R P(1)Q(1)Q(1)P(1)N.A.N.A.N.A.N.A.*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.5-8 ODCM Rev. 34 TABLE A.5.1-2 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REOU1REMENTS (Continued)

CHANNEL INSTRUMENT CHECK SOURCE CHECK CHANNEL CALIBRATION CHANNEL OPERATIONAL TEST 3. Radioactivity Monitor Providing Alarm but Not Termination of Release a. Primary Component Cooling Water System (in lieu of service water monitors)4. Rate of Change Monitor a. Primary Component Cooling Water System (in lieu of service water monitors)D M R(2)Q(1)D(4)N.A.R N.A A.5-9 ODCM Rev. 34 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 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic, or batch releases are made.(4) CHANNEL CHECK shall consist of verifying indication of tank level during periods of release.CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .A.5-1 0 ODCM Rev. 34 5.2 Radioactive Gaseous Effluent Monitoring Instrumentation 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 a time period determined by an evaluation conducted in accordance with the requirements of the Corrective Action Program. An evaluation is not required if the noncompliance is a consequence of surveillance testing or planned maintenance.

SURVEILLANCE REQUIREMENTS S.5.2 Each radioactive gaseous 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.5.2-2.BASES The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents.

The Alarm/Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM (Part B) to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. The sensitivity of any noble gas activity monitors used to show compliance with the gaseous effluent release requirements of Control C.7.2.1 shall be such that concentrations as low as 1 X 10-6 liCi/cc are measurable.

A.5-11 ODCM Rev. 34 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.

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 10- 1 tCi/cc Mid Range 10- -10 3 High Range 10 10 5 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 and MMOD 02-0531)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&I) 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.5-12 ODCM Rev. 34 At all times,-Heat tracing (HT) on the sample lines from the plant vent to the WRGM.Note: Dewpoint measurements may be 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 Comoonent Action Low range NG detector High range NG detector Mid range NG detector RM-P-391 Enter Action 33. Perform grab sampling as required.Enter Action 33. 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 unlikely event that elevated effluent concentrations above the capability of the low range detector are present, then grab sampling or backup monitoring will/may be required.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-240-1]

starts on increasing activity.)

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 or 3 is in service then no further action is required.

If the low range P&I filters are out of service then comply with Action 35 within one hour.Enter Actions 32, 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.

Actions 32 and 35 are satisfied ifP-240-2, and filters F-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 Actions 32 and 35 within one hour of identifying the out of service condition.

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 high range detectors.

P-240-1 (High range pump)A.5-13 ODCM Rev. 34 P-240-2 (Low range pump)HT circuit: CP-434 Ckt 28.(Sample line temperature less than 20'F above ambient as indicated by computer points B5946 / B5948.)Flow rate monitor and/or sampler flow rate monitor.Enter Actions 32, 33 and 35. Action 33 is satisfied by performing grab samples. Actions 32 and 35 are satisfied by ensuring one of the following sample configurations are in service within one hour of identifying the out-of-service condition; the operation of P-391 with filters F-156-7 & 8,or installing a portable sample pump across valves V2, and V5 to facilitate P&I grab sampling.Enter Action 36. Action 36 is satisfied and the WRGM may remain OPERABLE with CP-434 Ckt 28 out of service provided that CP-426 Ckt 46 is energized within 1 hr of the out-of-service condition.

If the sample line temperature can not be maintained greater than or equal to 207F above ambient, the particulate and iodine samples remain valid provided moisture is not present.Action 36 provides moisture monitoring capability.

Comply with Action 32.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.A.5-14 ODCM Rev. 34 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.

A procedural method of collecting the grab sample from the plant vent release pathway may require 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 allows effluent releases 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). Action 32 supports Action 35 by providing periodic sample flow rate monitoring for use in the iodine and particulate activity determinations.

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).It must be noted that Actions 32 and 35 pertain to the iodine and particulate samplers.

For noble gas collection, Action 33 is applicable which requires grab samples be taken once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and analyzed for radioactivity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . 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 equipment, 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 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.

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 due to low and high activity sample flow paths). 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.A.5-15 ODCM Rev. 34 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 or as an alternate to RM-6504 and ACTION Statement

33 not entered. If RM-6503 were considered an 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.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-1602.

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.Table A.5.2-1, Radioactive Gaseous Effluent Monitoring Instrumentation, specifically lists RM-6504 as the instrument required to satisfy the Limiting Condition for operation.

This table also states that the monitor provide the functions of alarm and automatic termination of release.A.5-16 ODCM Rev. 34 TABLE A.5.2-1 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS OPERABLE INSTRUMENT

1. (Not Used)2. PLANT VENT-WIDE RANGE GAS MONITOR a. Noble Gas Activity Monitor b. Iodine Sampler c. Particulate Sampler d. Flow Rate Monitor e. Sampler Flow Rate Monitor f. Sample Line Temperature

3. GASEOUS WASTE PROCESSING SYSTEM (Providing Alarm and Automatic Termination of Release -RM-6504)a. Noble Gas Activity Monitor (Process)4. TURBINE GLAND SEAL CONDENSER EXHAUST a. Iodine Sampler b. Particulate Sampler c. Sampler Flow Rate Indicator d. Noble Gas Activity Monitor (RM 6505)APPLICABILITY ACTION 1 1 1 1******33 35 35 32 32, 35 36 1 1 1*33 35 35 32, 35 34***At all times.(Not Used.)When the gland seal exhauster is in operation.

A.5-17 ODCM Rev. 34 TABLE A.5.2-1 (Continued)

ACTION STATEMENTS ACTION 32 -ACTION 33 -ACTION 34 -ACTION 35 -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 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .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 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . For RM-6504, RM-6503 may be used instead of taking grab samples.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 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and analyzed for radioactivity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .With the number of channels OPERABLE less than the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue 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.If, for any reason, the sample line temperature cannot be maintained greater than or equal to 200 F above outside ambient air temperature, the WRGM may remain OPERABLE provided dewpoint measurements are obtained every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months verifying that conditions do not exist for condensation in the sample line with the inservice operating sample pump. (CX0901.38)

ACTION 36 -A.5-18 ODCM Rev. 34 TABLE A.5.2-2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 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. 240W 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.5-19 ODCM Rev. 34 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 Alarm/Trip 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.5-20 ODCM Rev. 34 6.0 RADIOACTIVE LIQUID EFFLUENTS 6.1 Concentration CONTROLS C.6. 1.1 The concentration of radioactive material released in liquid effluents at the point of discharge from the multiport diffuser (see Technical Specifications Figure 5.1-3) shall be limited to 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 1 0-4 1Ci/ml total activity.APPLICABILITY:

At all times.ACTION: With the concentration of radioactive material released in liquid effluents at the point of discharge from the multiport diffuser exceeding the above limits, restore the concentration to within the above limits within 15 minutes.SURVEILLANCE REQUIREMENTS 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. l.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 H.A design objectives of Appendix I, 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 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months 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 often 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.A.6-1 ODCM Rev. 33 I 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 .Ci/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 2X 10-4 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 CFR 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 5xl 0-7 gCi/cc, the compensatory samples for inoperable SGBD Flash Tank and Turbine Building Sump Monitors are required to be counted to an LLD of I xl0-7 pCi/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 lx1 0-7 tCi/cc is therefore an acceptable method of complying with these requirements; it is not necessary to meet the LLD of lxlO-8[tCi/cc specified as the equivalent sensitivity of the PCCW Head Tank Rate of Change Monitor.A.6-2 ODCM Rev. 33 I TABLE A.6.1-1 RADIOACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit of Minimum Analysis Detection Liquid Release Type Sampling Frequency Frequency Type of Activity Analysis (LLD) (')(PCi/ml)A. 1. Liquid Radwaste Test Tanks P P Principal Gamma Emitters(3) 5x10" 7 (WL-TK-63A+B)

Each Batch Each Batch (BRS-TK58A+B)

SGBD Waste Holdup Sump 2. Neutralization Tank (TK-32) 1-131 1xl0" 6 3. Low Conductivity Tank P M Dissolved and Entrained x10-5 (TK-274) One Batch/M Gases (Gamma Emitters)4. Steam Generator Drains (i.e. P M(4) H-3 lx10-5 from wet lay-up)(6)

Each Batch Composite 5. Temporary Tanks via CPS Gross Alpha 1xl0-7 Discharge Line (14)(Batch Release)(2)

P Q(4)(9) Sr-89, Sr-90 5x10-8 Each Batch Composite Fe-55 1x l0 " 6 B. 1. Turbine Building Sump W W Principal Gamma Emitters(3) 5x10-7 Effluent(8) Grab Sample 2. Steam Generator Blowdown (6)(8)3. Condensate Polishing 1-131 1xl0" 6 Steam Generator Blowdown Demineralizer Megarinse (8)(12)(Continuous Release) (5) W M Dissolved and Entrained 1xl0-5 Grab Sample Gases (Gamma Emitters)A.6-3 ODCM Rev. 33 TABLE A.6.1-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM (Continued)

Lower Limit of Detection Minimum Analysis (LLD) (1)Liquid Release Type Sampling Frequency Frequency Type of Activity Analysis (PCi/ml)B. (Continued)

W M H-3 1xl0 5 Grab Sample Gross Alpha 1xl0-7 W Q(9) Sr-89, Sr-90 5x1O-8 Grab Sample Fe-55 1X10-C. Service Water(7)(1 0) W W Principal Gamma Emitters(3) 5x10-7 Grab Sample 1-13 1 lx10 "6 W M Dissolved and Entrained lxl0 5 Grab Sample Gases (Gamma Emitters)W M H-3 1xl0" 5 Grab Sample Gross Alpha lxl0" 7 W Q(9) Sr-89, Sr-90 5x10 8 Grab Sample Fe-55 lx10-6 A.6-4 ODCM Rev. 33 TABLE A.6.1-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM (Continued)

Lower Limit Minimum Analysis of Detection_j Frequency Type of Activity Analysis (LLD) (')Liquid Release Type Sampling Frequency (PCi/ml)D. Subsurface Dewatering (11) M M Principal Gamma Emitters (3) 5xl0.7 Grab Sample H-3 2x] 0-O M M Gross Alpha lxl0-7 Grab Sample M Q(9) Sr-89, Sr-90 5x1O"8 Grab Sample Fe-55 I1 x 0-6 E. Storm Drains (13) W W Principal Gamma Emitters (3) 5x] 0-7 Composite Sample (4)H-3 2x1O-6 W M Gross Alpha lx10 7 Composite Sample W Q(9) Sr-89, Sr-90 5x1O8 Composite Sample Fe-55 lx O0-6 P -Prior to Discharge W -Weekly M -Monthly Q -Quarterly A.6-5 ODCM Rev. 33 TABLE A.6.1 -1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM (Continued)

Notations (1) 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 E x V x 2.22 x 106 x Y x exp (-AA t)Where: LLD = the "a priori" lower limit of detection (microcurie per unit mass or volume), 4.66 = a constant derived from the Kalpha and Kbeta 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, A = the radioactive decay constant for the particular radionuclide (s-1), and A t = 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 prori (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.

(2) A batch release is the discharge of liquid wastes of a discrete volume. A batch discharge that is interrupted and reinitiated at a later time with no additional input (verified input isolated) is considered to be one release. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed to assure representative sampling.A.6-6 ODCM Rev. 33 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.

During plant operation in Modes 1 through 4, Steam Generator Blowdown Flash Tank releases are considered continuous releases (sampled prior to or during the release) due to the potential input from primary to secondary system leakage during discharges.-

During plant outages (Modes 5 and 6), Steam Generator drains are considered as batch type releases (input isolated, recirculated and sampled prior to release).(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 for other than naturally occurring radioactivity.

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 1 x1 0-3 [LC/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 lx10-4 ptC/cc, sample SWS every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for principal gamma emitters.A.6-7 ODCM Rev. 33 I TABLE A.6.1-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM (Continued)

Notations (Continued)

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 lxl0--tgC/cc.

If this sensitivity cannot be achieved, the SWS will be sampled once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .(8) If the Turbine Building Sump, Steam Generator Blowdown Flash Tank, Steam Generator Blowdown Demineralizer Megarinse, or Condensate Polishing Demineralizer Megarinse discharges isolate due to high concentration of radioactivity, that liquid stream will be sampled and analyzed for Iodine-131 and principal gamma emitters prior to release.(9) Quarterly composite analysis requirements shall only be required when analysis for principal gamma emitters indicate positive radioactivity other than naturally occurring.

(10) 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.() Principal gamma emitters and tritium shall be analyzed monthly in subsurface dewatering samples. Sample and 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.for other than naturally occurring radioactivity.

(12) Condensate Polishing Demineralizer or Steam Generator Blowdown Megarinse water shall be sampled and analyzed for principal gamma emitters and 1-131 prior to initiation of the megarinse discharge to the circulating water system.(13) The Storm Drain System is not designed as a pathway for plant related liquid effluent waste releases.

Storm drains are routinely used as the plant subsurface de-watering, secondary steam trap condensation and auxiliary boiler liquid release pathway to the circulating water system.Incidental releases from other sources that contain insignificant quantities of radioactivity may enter and use the storm drain system as a pathway to the circulating water system. Principal gamma emitters and tritium shall be analyzed for weekly. Sample and 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 for other than naturally occurring radioactivity, from an unidentified source.(14) The Condensate Polishing System (CPS) has the provision to discharge temporary tanks as batch releases through the same discharge path as the Neutralization Tank (TK-32). The requirements for discharges from a temporary tank through this CPS pathway are equivalent to those of the Neutralization Tank.A.6-8 ODCM Rev. 33 1 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 REQUIREMENTS 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.6-9 ODCM Rev. 33 I BASES This Control is provided to implement the requirements of Sections II.A, III.A, and [V.A of Appendix I to 10 CFR Part 50. The Control implements the guides set forth in Section H.A 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 material in liquid effluents to UNRESTRICTED AREAS will be kept as low as is reasonably achievable.

The dose calculation methodology and parameters in the ODCM implement the requirements in Section III.A of Appendix I that 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.A.6- 10 ODCM Rev. 33 1 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 31-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.6-11 ODCM Rev. 33 1 BASES The OPERABILITY of the Liquid Radwaste Treatment System ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment.

The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept as low as is reasonably achievable.

This specification implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the Liquid Radwaste Treatment System were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix A to 10 CFR Part 50 for liquid effluents.

A.6-12 ODCM Rev. 33 1 7.0 RADIOACTIVE GASEOUS EFFLUENTS 7.1 Dose Rate CONTROLS 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 (see Technical 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 and less than or equal to 3000 mrems/yr to the skin, and b. For Iodine-131, for Iodine-133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrems/yr to any organ.APPLICABILITY:

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.7-1 ODCM Rev. 25 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 10 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 Detection0 1)(LLD) (gtCi/cc)1. Plant Vent M(3)(4) M Principal Gamma lxiO-4 Grab Sample Emitters(2)

H-3 lxl0" 6 Continuous(5)

W(6) 1-131 IxI0-1 2 Charcoal Sample Continuous(5)

W(6) Principal Gamma 1xl10 1 1 Particulate Sample Emitters(2)

Continuous(5)

M Gross Alpha 1x 10 1 1 Composite Particulate Sample Continuous(5)

Q Sr-89, Sr-90 1xl 011 Composite Particulate Sample 2. Condenser Air M(7) M(7) Principal Gamma lxl0-4 Removal Exhaust Grab Sample Noble Gases Emitters(2)

H-3 1x10-6 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(') (LLD)(PCi/cc)3. Gland Steam Packing Continuous W Principal Gamma 1xl0-11 Exhauster Particulate Sample Emitters(2)

Continuous W 1-131 IxIO112 Charcoal Sample Continuous M Gross Alpha 1xl 011 Composite Particulate Sample Continuous Q Sr-89, Sr-90 1x10a1" Composite Particulate Sample(8)4. Containment Purge p(3) P Principal Gamma 1xlO0 Each Purge Grab Each Purge Emitters(2)

Sample H-3 (oxide) 1x10-6 A.7-4 ODCM Rev. 25 TABLE A.7.1-1 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM (Continued)

Notations (1) 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 E x V x 2.22 x 106 x Y x exp (-2A t)Where: LLD = the "a priori" lower limit of detection (microcurie per unit mass or volume), 4.66 = a constant derived from the Kalpha and Kbeta 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, 2 = the radioactive decay constant for the particular radionuclide (s-1), and A t = 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.

A.7-5 ODCM Rev. 25 TABLE A.7.1-1 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, 1-131, 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 1-131 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 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing, or after removal from sampler. Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least seven (7) days following each shutdown, startup, or THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER within a one-hour period and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing.

When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLDs may be increased by a factor of 10.This requirement does not apply if 1) analysis shows that the DOSE EQUIVALENT 1-131 concentration in the reactor coolant has not increased more than a factor of 3; and 2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.(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 C.7.2.1 The air dose due to noble gases released in gaseous effluents to areas at and beyond the SITE BOUNDARY (see Technical Specification Figure 5.1-1) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radiation, and b. During any calendar year: Less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.

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 BASES This Control is provided to implement the requirements of Sections II.B, HI.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 III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated.

The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109,"Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977, and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977. The ODCM 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 form with 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 REQUIREMENTS S.7.3.1 Cumulative dose contributions for the current calendar quarter and current calendar year for Iodine- 131, Iodine- 133, tritium, and radionuclides in particulate form with half-lives greater than.8 days shall be determined in accordance with the methodology and parameters in Part B of the ODCM at least once per 31 days.A.7-9 ODCM Rev. 25 BASES This Control is provided to implement the requirements of Sections II.C, IH.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 [V.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 lII.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 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical 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 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.7-11 ODCM Rev. 25 BASES The OPERABILITY of the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment.

The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept as low as is reasonably achievable.

This Control implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I to 10 CFR Part 50, for gaseous effluents.

A.7-12 ODCM Rev. 25 8.0 TOTAL DOSE CONTROL C.8.1.1 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 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.6.2.1 .a, C.6.2.I .b, C.7.2.1 .a, C.7.2.1 .b, C.7.3.1 .a, or C.7.3.1 .b, calculations shall be made including direct radiation contributions from the units and from outside storage tanks to determine whether the above limits of Control C.8. 1.1 have been exceeded.

If such is the case, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.8.2, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR 20.2203(a)(4), 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 REQUIREMENTS 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 plant facilities, including radwaste storage tanks and Dry Fuel Storage Facility, 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. 32 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 I, and if direct radiation doses from the units (including outside storage tanks, solid radwaste storage and dry fuel storage, 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 [equivalent to 10 CFR 72.104(a) for Dry Fuel Storage considerations], 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. 32 1 9.0 RADIOLOGICAL ENVIRONMENTAL MONITORING 9.1 Plant Operations Monitoring Program CONTROL C.9.1.1 The Radiological Environmental Monitoring Program (REMP) shall be conducted as specified in Table A.9.1-1.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 limits of Control C.6.2.1, C.7.2.1, or C.7.3.1. When more than one of the radionuclides in the REMP are detected in the sampling medium, this report shall be submitted if concentration (1) concentration (2)reporting level (1) + reporting level (2) + ...> 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 Control C.6.2.1, C.7.2.1, or C.7.3.1. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report required by Technical Specification 6.8.1.3 and Part A, Section 10.1 of 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. 35 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.9.1-1 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 implementsSection 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. 35 TABLE A.9.1-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway and/or Number of Representative Samples and Sampling and Collection Type and Frequency of Sample Sample Locationsa Frequency Analysis 1. 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-km range from the site;The balance of the stations to be placed in special interest areas such as population centers, nearby residences, schools, and control locations.

2. AIRBORNE Radioiodine and Samples from five locations

Continuous sampler Radioiodine Canister: Particulates operation with sample Three samples from close to the three SITE collection biweekly 1-131 analysis biweekly BOUNDARY locations, in different sectors, (approx. 14 days), or (approx. 14 days).of high calculated long-term average more frequently if ground-level D/Q. required by dust loading. Particulate Sampler: One sample from the vicinity of a Gross beta radioactivity community having the highest calculated analysis following filter long-term average ground-level D/Q. changec;Gamma isotopic analysise of composite (by location)quarterly.

A.9-3 ODCM Rev. 35 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 Locationsa 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 isotopice and 1-131 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 yr.f 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. 35 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 a Frequency Analysis 4. (Continued)

One sample of each of three commercially Sample in season, or Gamma isotopic analysis' on and recreationally important species in semiannually if they are edible portions.b. Fish and vicinity of plant discharge area. not seasonal.Invertebrates One sample of similar species in areas not influenced by plant discharge.

c. Food Products Samples of three (if practical) different Monthly, when available.

Gamma isotopice and .1-131 kinds of broad leaf vegetation' grown analysis.nearest each of two different off-site locations of highest predicted long-term average ground-level D/Q if milk sampling is not performed.

h One sample of each of the similar broad leaf Monthly, when available.

Gamma isotopice and 1-131 vegetationg grown at a control location, as analysis.for example 15-30 km distant in the least prevalent wind direction, if milk sampling is not performed.

h A.9-5 ODCM Rev. 35 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 1 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 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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.g. If broad leaf vegetation is unavailable, other vegetation will be sampled.h. Vegetation samples may be taken in the general vicinity of the designated sample location, due to availability.

A.9-6 ODCM Rev. 35 TABLE A.9.1-2 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSISa'rg Lower Limit of Detection (LLD)b Fish and Water Airborne Particulate or Invertebrates Milk Food Products Sediment Analysis (pCi/kg) Gas (pCi/mi 3) (pCi/kg, wet) (pCi/kg) (pCi/kg, wet) (pCi/kg, dry)Gross Beta H-3 Mn-54 Fe-59 Co-58, 60 Zn-65 Zr-Nb-95 1-131 Cs-134 Cs-137 Ba-La-140 4 3,000 15 30 15 30 15C 0.01 130 260 130 260 15 15 18 15c' d 0.07 0.05 0.06 130 150 1 15 18 15cd 60e 60 80 150 180 A.9-7 ODCM Rev. 35 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(-AAt)

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 Kalpha and Kbeta 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; 2 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. 35 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 LLDs 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. 35 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/m 3) 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*1-131 100 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

/ site groundwater monitoring well reporting level = 20,000 pCi/kg (2E-05 gCi/ml)A.9-10 ODCM Rev. 35 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 of the 16 meteorological sectors of the nearest milk animal, the nearest residence, and the nearest garden** of greater than 50 m 2 (500 ft 2) producing broad leaf vegetation.

APPLICABILITY:

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.9-11I ODCM Rev. 35 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 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 m 2 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. 35 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 REQUIREMENTS 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. 35 9.4 Dry Fuel Storage Facility Monitoring Program CONTROL C.9.4.1 The Dry Fuel Storage Facility radiological environmental monitoring program shall be conducted as specified in Table A.9.4-1 APPLICABILITY:

At all times.ACTION: With the Dry Fuel Storage Facility radiological environmental measurements not being conducted as specified in Table A.9.4-1, 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.

SURVEILLANCE REQUIREMENTS S.9.4.1 The Dry Fuel Storage Facility radiological environmental measurements shall be performed pursuant to Table A.9.4-1 from the specific locations given in the table(s) and figure(s) in Part B of the ODCM, and shall be analyzed pursuant to the requirements of Table A.9.4-1.BASES The Dry Fuel Storage radiological environmental monitoring program required by this Control provides representative measurements of direct (including scattered) radiation exposure at those locations that have the highest potential for dose to MEMBERS OF THE PUBLIC resulting from dry fuel storage operations.

The design of the storage facility is such that there are no liquid or gaseous effluents released to the environment from DFS and, therefore, no associated exposure pathways for liquids and gases requiring the collection and analysis of such sample media. As a result, only direct (including scattered) radiation from the DFS modules need to be monitored for integrated exposures in areas where doses to MEMBERS OF THE PUBLIC need to be limited.This measurement program provides information to support the determination of compliance with the dose criteria of 10 CFR 72.104(a), 40 CFR 190, and 10 CFR 20.1301(a)(1) for that portion of the total annual dose contributed by DFS.A.9-14 ODCM Rev. 35 TABLE A.9.4-1 DRY FUEL STORAGE FACILITY RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Locations Collection Frequency Type and Frequency of Analysis DIRECT RADIATION b 14 routine monitoring stations with two or Quarterly.

Gamma dose quarterly.

more dosimeters placed at each station as follows: 2 nearby locations on-site where MEMBERS OF THE PUBLIC can congregate.

5 locations along the site boundary in areas of high potential for public exposure (shared TLD Locations with plant REMP, Table B.4-1).2 locations near the closest site boundary approach to the DFS in different sectors.5 locations beyond 15 km from the site (shared TLD locations with plant REMP, Table B.4-1).a. Specific parameters of distance and direction sector from the centerline of the DFS Pad, and additional description where pertinent, shall be provided for each sample location in Table B.4-2 in the ODCM, Part B. Deviations are permitted from the required measurement schedule if dosimeters are unobtainable due to circumstances beyond the control of the licensee, such as lost of dosimeters resulting from unauthorized removal from the field. All deviations from the measurement schedule shall be documented in the Annual Radiological Environmental Operating Report as specified in Part A, Section 10.1.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.

A.9-15 ODCM Rev. 35 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, and an analysis of trends of the results of the radiological environmental Surveillance activities for the report period, including a comparison with preoperational studies, with operational Controls, as appropriate, and with previous environmental Surveillance reports, and an assessment of the observed impacts of the plant operation on the environment.

The reports shall also include the results of the Land Use Census required by Control C.9.2.1.The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in Part 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, and discussion of all deviations from the sampling schedule; discussion of environmental sample measurements that exceed the reporting levels but are not the result of plant effluents, pursuant to ACTION b. of Control C.9.1.1; and discussion of all analyses in which the LLD required was not achievable.

- - - 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. 29 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. 29 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 a 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. 29 SEABROOK STATION ODCM PART B RADIOLOGICAL CALCULATIONAL METHODS AND PARAMETERS B.1-0 ODCM Rev. 33

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 1OCFR 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 1 January 1993. The decision to continue the use of the "old" version of 10 CFR Part 20 is based on an NRC letter dated June 30, 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 3.11, can be used as an indicator of when additional assessments of on-site exposure to members of the public is advisable (see Section 3.11.2). Off-site critical organ doses from station effluents should not, however, be the only indicator of potential on-site doses.B.1-1 ODCM Rev. 33 1 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 H 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.I-10 through B. 1-14 and the Regulatory Limits are summarized in Table B. 1-l.The variables and special definitions used in this ODCM, Part B, are in Tables B. 1-8 and B. 1-9.B. 1-2 ODCM Rev. 33 I TABLE B. 1 -1

SUMMARY

OF RADIOLOGICAL EFFLUENT PART A CONTROLS ANTD IMPLEMENTING EOUJATIONS ND IMPLEMENTING E TIONS Part A Control Category Method 10)Eq. 2-1 Limit< 1.0 C.6.1.1 Liquid Effluent Concentration Total Fraction of ECL Excluding Noble Gases Total Noble Gas Concentration Eq. 2-2 Eq. 3-1 C.6.2.1 Liquid Effluent Dose Total Body Dose Organ Dose Eq. 3-2< 2 x 10" 4 iCi/ml<1.5 mrem in a qtr.< 3.0 mrem in a yr.< 5 mrem in a qtr.< 10 mrem in a yr.< 0.06 mrem in a mo.< 0.2 mrem in a mo.< 500 mrem/yr.C.6.3.1 Liquid Radwaste Treatment Operability C.7.1.1 Gaseous Effluents Dose Rate Total Body Dose Organ Dose Total Body Dose Rate from Noble Gases Skin Dose Rate from Noble Gases Organ Dose Rate from 1-131, 1-133, Tritium and Particulates with T 1/2> 8 Days Eq. 3-1 Eq. 3-2 Eq. 3-3 Eq. 3-4 Eq. 3-5< 3000 mrem/yr.< 1500 mrem/yr.B.1-3 ODCM Rev. 33 I TABLE B. 1-1

SUMMARY

OF RADIOLOGICAL EFFLUENT PART A CONTROLS AND IMPLEMENTING EQUATIONS (Continued)

Part A Control C.7.2.1 Gaseous Effluents Dose from Noble Gases Category Gamma Air Dose from Noble Gases BetaAir Dose from Noble Gases Organ Dose from Iodines, Tritium and Particulates with T 1/2 > 8 Days Method 10)Eq. 3-6 Limit< 5 mrad in a qtr.Eq. 3-7 C.7.3.1 C.7.4.1 C.8.1.1 Gaseous Effluents Dose from 1-131, 1-133, Tritium, and Particulates Ventilation Exhaust Treatment Total Dose (from All Sources)Eq. 3-8< 10 mrad in a yr.< 10 mrad in a qtr.< 20 mrad in a yr.< 7.5 mrem in a qtr.< 15 mrem in a yr.< 0.3 mrem in a mo.< 25 mrem in a yr.< 25 mrem in a yr.< 75 mrem in a yr.Organ Dose Eq. 3-8 Total Body Dose Organ Dose Thyroid Dose Footnote (2).C.5.1 Liquid Effluent Monitor Setpoint Liquid Waste Test Tank Monitor Alarm Setpoint Eq. 5-1 Control C.6.1.1 B. 1-4 ODCM Rev. 33 1 TABLE B.1-1

SUMMARY

OF RADIOLOGICAL EFFLUENT PART A CONTROLS AND IMPLEMENTING EQUATIONS (Continued)

Part A Controls Category Method 1(1)Limit C.5.2 Gaseous Effluent Monitor Setpoint Plant Vent Wide Range Gas Monitors Alarm/Trip Setpoint For Total Body Dose Rate Alarm/Trip Setpoint for Skin Dose Rate Eq. 5-5 Eq. 5-6 Control C.7.1.1 a (Total Body)Control C.7.1.1 a (Skin)(1) More accurate methods may be available (see subsequent chapters).

(2) Part A Control C.8.1.1a 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 II calculation, 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. 33 1 TABLE B. 1-2

SUMMARY

OF METHOD I EQUATIONS TO CALCULATE UNRESTRICTED AREA LIQUID CONCENTRATIONS Equation Number Category Equation 2-1 2-2 Total Fraction of ECL in Liquids, Except Noble Gases Total Activity of Dissolved and Entrained Noble Gases from all Station Sources FENG~ Z___ , 10 P ECLi (Cpi C1 .ml ) <- 2E-04 B.1-6 ODCM Rev. 33 1 TABLE B. 1-3

SUMMARY

OF METHOD I EQUATIONS TO CALCULATE OFF-SITE DOSES FROM LIOUID RELEASES Equation Number Category Equation 3-1 3-2 Total Body Dose Maximum Organ Dose D tb (mrem) = k YZ Q i DFL itb it Dmo (mrem) = k Q iDFL.o 1 imo B.1-7 ODCM Rev. 33 I TABLE B.1-4

SUMMARY

OF METHOD I EQUATIONS TO CALCULATE DOSE RATES Category Equation Number Receptor Locationa Release Heightb Equation Total Body Dose Rate From Noble Gases 3-3a OS E= 0.85

z(i

DFBi)3-3b 3-3c 3-3d 3-3e 3-3f OS EC EC G E G E G D)tb(g) 3.4

10i

DFBi)i DtbE(e) = 0.0015

Doi

DFBi)i DtbE~g) 0.0074

Y(Oi

DFBi)R R btbR(e)fDtbR(g)= 0.038

10i

DFBi)i= 0.2

D(Qi* DFBi)i aOS = Off-Site, EC = Science & Nature Center, formerly the Education Center, R = The "Rocks" bE = Elevated, G = Ground B. 1-8 ODCM Rev. 33 1 TABLE B. 1-4

SUMMARY

OF METHOD I EQUATIONS TO CALCULATE DOSE RATES (Continued)

Category Equation Number Receptor Locationa Release Heightb Equation Skin Dose Rate From Noble Gases 3-4a 3-4b 3-4c 3-4d 3-4e 3-4f OS OS EC EC E G E G E G Dskin(e) =

DF i(e))Dskin(g) = (,

DF i(g))DskinE(e)

=0.0014

Doi

DF iE(e))i DskiiE(g)

=0.0014

Doi

DF iE(g))DskinR(e)

= 0.0076

Doi

DF iR(e))i= 0.0076 *

DF' iR(g))i R R aOS =Off-Site, EC = Science & Nature Center, formerly the Education Center, R = The "Rocks" bE = Elevated, G = Ground B. 1-9 ODCM Rev. 33 TABLE B.1-4 OF METHOD I EOUATIONS TO CALCULATE DOSE RATES S UMMARY (Continued)

Category Equation Number Receptor Locationa Release Heightb Equation Critical Organ Dose Rate From 1-13 1, 1-133, H-3, and Particulate With T 1/2>8 Days 3-5a OS E 1co(e) = 1 (0,i* DFG' ico(e))3-5b 3-5c 3-5d 3-5e 3-5f OS EC EC G E G E G 1c.o(g) = (0

DFG' ico(g))i0.0014 *

DFG' icoE(e))OcoE(g) = 0.0014

10i* DFG' icoE(g))i DcoR(e) = 0.0076 *

DFG' icoR(e))1DoR(g) = 0.0076

i

DFG' icoR(g))R R aOS = Off-Site, EC = Science & Nature Center, formerly the Education Center, R = The "Rocks" bE = Elevated, G = Ground B.1-10 ODCM Rev. 33 TABLE B.1-5

SUMMARY

OF METHOD I EQUATIONS TO CALCULATE DOSES TO AIR FROM NOBLE GASES Category Equation Number Receptor Locationa Release Heightb Equation Gamma Dose to Air From Noble Gases 3-6a 3-6b 3-6c 3-6d 3-6e 3-6f OS OS EC EC E G E G E G Dae) = 3.2E-07

t-°275 * -(Qi* DF)i Dyi 1 (g) = 1.6E-06

t"°2 9 3 * *(Q* DFI')i D)irE(e) = 4.9 E- 10

t°-0252

E(Qi *DFir)DairE(g) = 4.4E-09 *t".321

E(Qi

DFI)i R R D.rR(e) = 5.1E-09

t-°-155

Z(Qi

DF')DarirR(g)

= 4.1E-08 *t'0.204 * (Qi

DFD')i aOs = Off-Site, EC = Science & Nature Center, formerly the Education Center, R = The "Rocks" bE = Elevated, G = Ground B.1-1lI ODCM Rev. 33 Cat Beta Dose From Nob TABLE B. 1-5

SUMMARY

OF METHOD I EQUATIONS TO CALCULATE DOSES TO AIR FROM NOBLE GASES (Continued) egory Equation Receptor Release Number Locationa e b Equation to Air 3-7a OS E D e = 4lE-07 *

I(Q* DF le Gasesaue 3-7b OS G Dar(g) = 6.0E-06

t 0 3 1 9 * -(Qi

D i 3-7c EC E DairE(e) = 1.8E-09

t-'319

_(Q *i 3-7d EC G D'E = 24E-08 * -0347 * *3-7e R E DairR(e) = 3.9E-08

t"°.24

Y(Qi *i 3-7f R G Dair(g) = 4.6E-07

t-0.67 * -(Qi *i P3)'F P)F P)x)fi):)R')aOs = Off-Site, EC = Science & Nature Center, formerly the Education Center, R = The "Rocks" bE = Elevated, G = Ground B.1- 12 ODCM Rev. 33 TABLE B.1-6

SUMMARY

OF METHOD I EQUATIONS TO CALCULATE DOSE TO AN INDIVIDUAL FROM TRITIUM, IODINE AND PARTICULATES Category Equation Number Receptor Locationa Release Heightb Equation Dose to Critical Organ From Iodines, Tritium, and Particulates 3-8a OS E Dco(e) = 14.8

t" 0" 2 9 7*

DFGico(e))

3-8b 3-8c 3-8d 3-8e 3-8f OS EC EC G E G E G Dco(g) = 17.7

t 3 1 6 * (Qi

DFGico(g))

DcoE(e) = 3.3 E- 02 * -0.3 4 9

Z(Qi

DFGicoE(e))

DcoE(g) = 3.3 E- 02 * -".3 4 7 * (Qi DFGicoE(g))

DcoR(e) = 7.3 E- 02

t".248 * (Qi DFGico R(e))DcoR(g) = 8.6 E- 02

t-0 2 6 7* -(Qi DFGicoR(g))

R R aOS = Off-Site, EC = Science & Nature Center, formerly the Education Center, R = The "Rocks" bE = Elevated, G = Ground B.1-13 ODCM Rev. 33 TABLE B. 1-7

SUMMARY

OF METHODS FOR SETPOINT DETERMINATIONS Equation Number Category Equation 5-1 Liquid Effluents:

Liquid Waste Test Rsetpoint ( "CiMI fl Fd I cyi SFm X DFminy Tank Monitor (RM-6509)

(1)5-23 PCCW Rate-of-Change Alarm Gaseous Effluents:

Plant Vent Wide Range Gas Monitors (RM-6528-1, 2, 3)RCset(gph)IxlO-8.SWF. 1 PCC 5-5 5-6 Total Body 1 R (PCi/sec) 588 1 f, tb DFBc R i 3000 1 fv skin (,uCi/sec)

DE'Skin (1) This equation maybe used for other effluent radiation monitors, such as the CPS Rad Monitor (RM-6473) where the fraction of total FCL (f,) is administratively adjusted for the particular pathway.B.1-14 ODCM Rev. 33 Variable NG Chi cNG Cdi Cpi Cy i Da~r(e)D-P(g)Da'IE(e)Dair E (g)DaIR(e)DarR(g)MDir(e)D air (g)DayixE(e)D~alrE(g)MDirR(e)TABLE B. 1-8

SUMMARY

OF VARIABLES Definition Concentration at point of discharge and entrained noble gas "i" in liquid pathways from all station sources Total activity of all dissolved and entrained noble gases in liquid pathways from all station sources Concentration of radionuclide "i" at the point of liquid discharge Concentration of radionuclide "i"= Concentration, exclusive of noble gases, of radionuclide "i" from tank "p" at point of discharge Concentration of radionuclide "i" in mixture at the monitor Off-site beta dose to air due to noble gases in elevated release= Off-site beta dose to air due to noble gas in ground level release= Beta dose to air at Science & Nature Center due to noble gases in elevated release= Beta dose to air at Science & Nature Center due to noble gases in ground level release Beta dose to air at "Rocks" due to noble gases in elevated release Beta dose to air at "Rocks" due to noble gases in ground level release Off-site gamma dose to air due to noble gases in elevated release Off-site gamma dose to air due to noble gases in ground level release Gamma dose to air at Science & Nature Center due to noble gases in elevated release Gamma dose to air at Science & Nature Center due to noble gases in ground level release Gamma dose to air at "Rocks" due to noble gases in elevated release Units RUCi/ml jtCi/ml jtCi/ml[LCi/ml[ICi/ml mrad mrad mrad mrad mrad mrad mrad mrad mrad mrad mrad B.1-15 ODCM Rev. 33 TABLE B. 1-8

SUMMARY

OF VARIABLES (Continued)

Variable DairR(g)Dco(e)Dco(g)DcoE(e)DcoE(g)DcoR(e)DcoR(g)Dd Dyfnte Dmo Ds Dtb DFmin DFtiny DF'i DF'iE DF'iR DFBi Definition Gamma dose to air at "Rocks" due to noble gases in ground level release Critical organ dose from an elevated release to an off-site receptor Critical organ dose from a ground level release to an off-site receptor Critical organ dose from an elevated release to a receptor at the Science & Nature Center Critical organ dose from a ground level release to a receptor at the Science & Nature Center Critical organ dose from an elevated release to a receptor at the "Rocks" Critical organ dose from a ground level release to a receptor at the "Rocks"= Direct dose= Gamma dose to air, corrected for finite cloud= Dose to the maximum organ= Dose to skin from beta and gamma Dose to the total body= Minimum required dilution factor based on all (beta -emitting and gamma -emitting) radionuclides 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 1 (dimensionless).

= Composite skin dose factor for off-site receptor Composite skin dose factor for Science & Nature Center= Composite skin dose factor for the "Rocks"= Total body gamma dose factor for nuclide 'T' (Table B.1-10)Units mrad mrem mrem mrem mrem mrem mrem mrem mrad mrem mrem mrem ratio mrem-sec/gCi-yr mrem-sec/gCi-yr mrem-sec/gCi-yr 3 mrem pCi- yr ODCM Rev. 33 B.1-16 TABLE B. 1-8

SUMMARY

OF VARIABLES (Continued)

Definition Variable Units DFBc DFLitb DFLimo DFBico(e)DFGico(g)DFGicoE(e)

DFGicoE(g)

DFGicoR(e)

DFGicoR(g)

DFG'ico(e)

DFG'icoE(e)

DFG'icoE(g)

DFG' icoR(e)Composite total body dose factor Site-specific, total body dose factor for a liquid release of nuclide "i" (Table B.1-1 1)Site-specific, maximum organ dose factor for a liquid release of nuclide "i" (Table B.I-11)Site-specific, critical organ dose factor for an elevated gaseous release of nuclide "i" (Table B. 1-12)Site-specific critical organ dose factor for a ground level release of nuclide "i" (Table B.1-12)Science & Nature Center-specific critical organ dose factor for an elevated release of nuclide "i" (Table B. 1-14)Science & Nature Center-specific critical organ dose factor for a ground level release of nuclide "i'(Table B.1-14)The "Rocks"-specific critical organ dose factor for an elevated release of nuclide "i" (Table B. 1-15)The "Rocks"-specific critical dose factor for a ground level release of nuclide "i" (Table B.1-15)Site-specific critical organ dose rate factor for an elevated gaseous release of nuclide "i" (Table B. 1-12)Site-specific critical organ dose rate factor for a ground level release of nuclide "i" (Table B.1-12)Science & Nature Center-specific critical organ dose rate factor for an elevated release of nuclide "i" (Table B. 1-14)Science & Nature Center-specific critical organ dose rate factor for a ground level release of nuclide "i" (Table B. 1-14)The "Rocks"-specific critical organ dose rate factor for an elevated release of nuclide "i" (Table B. 1-15)3 mrem pCi- yr mrem/gCi 1 mrem/[tCi mrem/gCi mrem/gCi mrem/ntCi mrem/gCi mrem/[lCi mrem/gCi mrem-sec/4Ci-yr mrem-sec/ýtCi-yr mrem-sec/[tCi-yr mrem-sec/[tCi-yr mrem-sec/gCi-yr B.1-17 ODCM Rev. 33 TABLE B. 1-8

SUMMARY

OF VARIABLES (Continued)

Variable Definition Units DFG' icoR(g) The "Rocks"-specific critical organ dose rate factor for a mrem-sec/jiCi-yr ground level release of nuclide "i" (Table B.1-15)DFSi Beta skin dose factor for nuclide "i" (Table B.I-IO) mrem-m 3 pCi- yr DF'i Combined skin dose factor for nuclide "i" mrem- sec/gCi-yr (Table B.I-10)DFI" Gamma air dose factor for nuclide "i" (Table B. 1-10) mrad-m 3 p Ci- yr DFip Beta air dose factor for nuclide "i" (Table B.1-10) mrad- m 3 pCi- yr Ico(e) Critical organ dose rate to an off-site receptor due to mrem elevated release of iodines, tritium, and particulates yr IDco(g) Critical organ dose rate to an off-site receptor due to mrem ground level release of iodines, tritium, and particulates yr DcoE(e) Critical organ dose rate to a receptor at the Science & mrem Nature Center due to an elevated release of iodines, yr tritium, and particulates DcoE(g) 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 IcoR(e) Critical organ dose rate to a receptor at the "Rocks" due mrem to an elevated release of iodines, tritium, and yr particulates DcoR(g) Critical organ dose rate to a receptor at the "Rocks" due mrem to a ground level release of iodines, tritium, and yr particulates Dskin(e) Skin dose rate to an off-site receptor due to noble gases mrem in an elevated release yr Dskin(g) Skin dose rate to an off-site receptor due to noble gases mrem in a ground level release yr.IskinE(e)

Skin dose rate to a receptor at the Science & Nature mrem Center due to noble gases in an elevated release yr B.I1-18 ODCM Rev. 33 TABLE B.1-8

SUMMARY

OF VARIABLES (Continued)

Variable Definition Units DskinE(g)

Skin dose rate to a receptor at the Science & Nature mrem Center due to noble gases in a ground level release yr DskinR(e)

Skin dose rate to a receptor at the "Rocks" due to noble mrem gases in an elevated release yr DskinR(g)

Skin dose rate to a receptor at the "Rocks" due to noble mrem gases in a ground level release yr Dtb(e) Total body dose rate to an off-site receptor due to noble mrem gases in an elevated release yr Dtb(g) Total body dose rate to an off-site receptor due to noble mrem gases in a ground level release yr DtbE(e) Total body dose rate to a receptor at the Science & mrem Nature Center due to noble gases in an elevated release yr DtbE(g) Total body dose rate to a receptor at the Science & mrem Nature Center due to noble gases in a ground level yr release DtbR(e) Total body dose rate to a receptor at the "Rocks" due to mrem noble gases in an elevated release yr DtbR(g) 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 m-ECLi Effluent concentration limit (ECL) for radionuclide "i" (excluding dissolved and entrained noble gas) as gtCi/ml specified in 10 CFR 20, Appendix B, Table 2.fv 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 ft 3/sec F= Flow rate past liquid waste test tank monitor gpm Fmax = Maximum allowable discharge flow rate from liquid gpm test tanks based on all (beta -emitting and gamma -emitting) radionuclides B.1-19 ODCM Rev. 33 TABLE B.1-8

SUMMARY

OF VARIABLES (Continued)

Variable Units Fmaxy F fgi ad fI ;f 2; f 3; f 4; f 5 ENG F 1 MPCi Qi Qi Rsetpoint 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 The fraction of the offsite limiting total body dose rate administratively assigned to monitored ground level release Flow rate past plant vent monitor Release reduction factor to be administratively assigned to account for potential unmonitored contributions from the Turbine Gland Seal Exhaust Fraction of total ECL associated with Paths 1, 2, 3, 4, and 5, the sum of which is < 1.Total fraction of ECL in liquid pathways (excluding noble gases)Maximum permissible concentration for radionuclide"i" ("old" 10 CFR 20, Appendix B, Table II, Column 2)= Release to the environment for radionuclide "i"= Release rate to the environment for radionuclide "i"= Liquid monitor response for the limiting concentration at the point of discharge Response of the noble gas monitor to limiting total body dose rate Response of the noble gas monitor to limiting total body dose rate= Shielding factor= Detector counting efficiency from the gas monitor calibration Detector counting efficiency for noble gas "i" Detector counting efficiency from the liquid monitor calibration Rskin gpm Dimensionless cc sec Dimensionless Dimensionless Dimensionless pCi/cc curies, or gt curies[tCi/sec lICi/ml cpm, or gCi/sec cpm, or gCi/sec Dimensionless cpm or mR/ hr P Ci- cc U Ci/cc cpm or mR/hr pU Ci- cc P Ci/cc cps/gCi/ml ODCM Rev. 33 Rtb SF Sg Sgi SI B. 1-20 TABLE B.1-8

SUMMARY

OF VARIABLES (Continued)

Definition Variable Units Sli X/Q[X/Q] Y"= Detector counting efficiency for radionuclide "i"= Average long-term undepleted atmospheric dispersion factor (Tables B.7-4, B.7-5, and B.7-6)Effective long-term average gamma atmospheric dispersion factor (Tables B.7-4, B.7-5, and B.7-6)= Service Water System flow rate= Primary component cooling water measured (decay corrected) gross radioactivity concentration Unitless factor which adjusts the value of atmospheric dispersion factors for elevated or ground-level releases with a total release duration oft hours sec 3 m sec 3 m cps/gCi/ml SWF PCC gph jtCi/ml t-a Dimensionless B.l1-21 ODCM Rev. 33 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 any 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.B. 1-22 ODCM Rev. 33 TABLE B.1-10 DOSE FACTORS SPECIFIC FOR SEABROOK STATION FOR NOBLE GAS RELEASES Radio-nuclide Gamma Total Body Dose Factor 3 DFBi (mrem- m3 pCi- yr Beta Skin Dose Factor 3 mrem- m pCi- yr Combined Skin Dose Factor for Elevated Release Points I mrem- sec DFi(e) ( -)p. Ci- yr Combined Skin Dose Factor for Ground Level Release Points , mrem- sec Dr i(g) ( ).t Ci- yr Beta Air Dose Gamma Air Dose Factor Factor DFif (mrad -m3 pCi -yr 7 mrad- m3 DF.( ..)1 pCi- yr Ar-41 Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Kr-90 Xe-131m Xe-133m Xe-133 Xe-135m Xe-135 Xe-137 Xe-138 8.84E-03 7.56E-08 1.17E-03 1.61E-05 5.92E-03 1.47E-02 1.66E-02 1.56E-02 9.15E-05 2.51E-04 2.94E-04 3.12E-03 1.81E-03 1.42E-03 8.83E-03 2.69E-03 1.46E-03 1.34E-03 9.73E-03 2.37E-03 1.01E-02 7.29E-03 4.76E-04 9.94E-04 3.06E-04 7.11E-04 1.86E-03 1.22E-02 4.13E-03 1.09E-02 1.81E-05 2.35E-03 1.11E-03 1.38E-02 1.62E-02 2.45E-02 2.13E-02 5.37E-04 1.12E-03 5.83E-04 3.74E-03 3.33E-03 1.14E-02 1.20E-02 6.20E-02 7.28E-05 1.92E-02 1.35E-02 1.21E-01 8.1OE-02 1.66E-01 1.34E-01 5.35E-03 1.12E-02 4.39E-03 1.98E-02 2.58E-02 1.28E-01 7.60E-02 3.28E-03 2.88E-04 1.97E-03 1.95E-03 1.03E-02 2.93E-03 1.06E-02 7.83E-03 1.11E-03 1.48E-03 1.05E-03 7.39E-04 2.46E-03 1.27E-02 4.75E-03 9.30E-03 1.93E-05 1.23E-03 1.72E-05 6.17E-03 1.52E-02 1.73E-02 1.63E-02 1.56E-04 3.27E-04 3.53E-04 3.36E-03 1.92E-03 1.51E-03 9.21E-03 8.84E-03 = 8.84 x 10-3 B. 1-23 ODCM Rev. 33 1 TABLE B.1-11 DOSE FACTORS SPECIFIC FOR SEABROOK STATION FOR LIQUID RELEASES Total Body Maximum Organ Dose Factor Dose Factor (mrem) (mrem)Radionuclide DFLitb ( ei DFLi mo Ci PCi DF pmo(H-3 3.02E-13 3.02E-13 Na-24 1.38E-10 1.42E-10 Cr-51 1.83E-11 1.48E-09 Mn-54 5.15E-09 2.68E-08 Fe-55 1.26E-08 7.67E-08 Fe-59 8.74E-08 6.66E-07 Co-58 2.46E-09 1.40E-08 Co-60 6.15E-08 9.22E-08 Zn-65 2.73E-07 5.49E-07 Br-83 1.30E-14 1.89E-14 Rb-86 4.18E-10 6.96E- 10 Sr-89 2.17E-10 7.59E-09 Sr-90 3.22E-08 1.31E-07 Nb-95 5.25E-10 1.58E-06 Mo-99 3.72E-11 2.67E-10 Tc-99m 5.22E-13 1.95E-12 Ag-110n 1.OIE-08 6.40E-07 Sb-124 1.71E-09 9.89E-09 Sb-125 6.28E-09 8.3 1E-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-131rm 3.16E-08 2.94E-06 Te-132 9.06E-08 3.80E-06 1-130 2.75E-11 3.17E-09 1-131 2.30E- 10 1.OOE-07 1-132 6.28E- 11 6.36E-11 1-133 3.85E-11 1.15E-08 1-134 1.19E-12 1.41E-12 1-135 5.33E-11 4.69E-10 Cs-134 3.24E-08 3.56E-08 Cs-136 2.47E-09 3.27E-09 Cs-137 3.58E-08 4.03E-08 Ba-140 1.70E-10 3.49E-09 La-140 1.07E-10 4.14E-08 Ce-141 3.85E-11 9.31E-09 Ce-144 1.96E-10 6.46E-08 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. 1-24 ODCM Rev. 33 TABLE B.l-12 DOSE AND DOSE RATE FACTORS SPECIFIC FOR SEABROOK STATION FOR IODINES, TRITIUM AND PARTICULATE RELEASES Radio-nuclide H-3 Cr-51 Mn-54 Fe-59 Co-58 Co-60 Zn-65 Sr-89 Sr-90 Zr-95 Nb-95 Mo-99 Ru-103 Ag-110m Sb-124 1-131 1-133 Cs-134 Cs-137 Ba-140 Ce-141 Ce-144 Other*Critical Organ Dose Factor for Elevated Release Point (mrem)DFGic.(e) (rei 3.08E-10 8.28E-09 1.11E-06 1.06E-06 5.56E-07 1.21E-05 2.33E-06 1.98E-05 7.21E-04 1.1OE-06 2.01E-06 1.63E-08 3.03E-06 5.02E-06 1.83E-06 1.47E-04 1.45E-06 5.62E-05 5.47E-05 1.55E-07 2.65E-07 6.09E-06 4.09E-06 Critical Organ Dose Factor for Ground Level Release Point (mrem)DFGi.co(g)

KuCir 3.76E-09 2.89E-08 3.79E-06 3.65E-06 1.91E-06 4.12E-05 7.93E-06 6.73E-05 2.47E-03 3.77E-06 6.86E-06 1.1OE-07 1.04E-05 1.72E-05 6.28E-06 5.04E-04 5.72E-06 1.91E-04 1.86E-04 6.39E-07 9.28E-07 2.09E-05 1.39E-05 Critical Organ Dose Rate Factor for Elevated Release Point mrem- sec DFGic~e ~yr-p Ci)9.7 1E-03 2.91E-01 4.38E+O 1 3.53E+01 2.OOE+01 5.42E+02 7.82E+01 6.24E+02 2.27E+04 3.63E+O 1 6.40E+01 5.39E-01 9.62E+01 1.80E+02 6.15E+01 4.64E+03 4.57E+01 1.81E+03 1.79E+03 5.01E+00 8.45E+00 1.93E+02 1.29E+02 Critical Organ Dose Rate Factor for Ground Level Release Point (mrem- sec)D FG 'ico(g) ( r -e"l C 1.19E-01 1.01E+00 1.50E+02 1.21E+02 6.88E+01 1.85E+03 2.66E+02 2.12E+03 7.79E+04 1.24E+02 2.20E+02 3.56E+00 3.3 1E+02 6.15E+02 2.11E+02 1.59E+04 1.80E+02 6.18E+03 6.09E+03 2.06E+01 2.96E+01 6.62E+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.1-25 ODCM Rev. 33 1 TABLE B. 1-13 COMBINED SKIN DOSE RATE FACTORS SPECIFIC FOR SEABROOK STATION SPECIAL RECEPTORS(')

FOR NOBLE GAS RELEASE Science & Nature Center Combined Skin Dose Rate Factor for Elevated Release Point DF' iE(e) (mrem-sec p Ci- yr Science & Nature Center Combined Skin Dose Rate Factor for Ground Level Release Point DF' ((mrem- sec)iE(g) ,p Ci- yr The "Rocks" Combined Skin Dose Rate Factor for Elevated Release Point DF' (mrem-sec) iR(e) ,Ci- yr The "Rocks" Combined Skin Dose Rate Factor for Ground Level Release Point DF' (mrem-sec) iRgu "Ci- yr Radio-nuclide Ar-41 Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Kr-90 Xe-131m Xe-133m Xe-133 Xe-135m Xe-135 Xe-137 Xe-138 1.57E-02 2.35E-05 3.84E-03 2.16E-03 2.3 1E-02 2.23E-02 3.73E-02 3.15E-02 9.52E-04 1.99E-03 9.20E-04 5.24E-03 5.32E-03 2.14E-02 1.78E-02 1.17E-01 1.13E-04 4.08E-02 3.09E-02 2.60E-01 1.44E-01 3.34E-0 1 2.64E-01 1.19E-02 2.48E-02 9.11 E-03 3.61E-02 5.4 1E-02 2.89E-01 1.49E-01 9.73E-02 1.07E-04 3.16E-02 2.29E-02 2.OOE-01 1.25E-01 2.68E-01 2.14E-01 8.96E-03 1.87E-02 7.16E-03 3.07E-02 4.23E-02 2.16E-01 1.21E-01 6.99E-0 I 5.57E-04 2.69E-01 2.15E-01 1.73E+00 8.18E-01 2.12E+00 1.64E+00 8.07E-02 1.68E-01 5.91E-02 2.11E-01 3.53E-01 2.OOE+00 9.27E-01 (1) See Seabrook Station Technical Specification Figure 5.1-1.B. 1-26 ODCM Rev. 33 I TABLE B.l-14 DOSE AND DOSE RATE FACTORS SPECIFIC FOR THE SCIENCE & NATURE CENTER FOR IODINE, TRITIUM, AND PARTICULATE RELEASES Radio-nuclide H-3 Cr-51 Mn-54 Fe-59 Co-58 Co-60 Zn-65 Sr-89 Sr-90 Zr-95 Nb-95 Mo-99 Ru-103 Ag-il0m Sb- 124 1-131 1-133 Cs-134 Cs-137 Ba-140 Ce-141 Ce-144 Other*Critical Organ Dose Factor for Elevated Release Point (mrem)DFGicoE(e) ( ,uCi 6.45E- I1 4.98E-09 1.39E-06 3.09E-07 3.89E-07 2.17E-05 7.34E-07 1.15E-07 5.14E-06 3.38E-07 1.53E-07 1.62E-08 1.30E-07 3.43E-06 6.96E-07 7.79E-07 1.84E-07 6.83E-06 1.03E-05 1.14E-07 4.09E-08 6.95E-07 2.26E-06 Critical Organ Dose Factor for Ground Level Release Point (mrem)D FG icoE(g) ] .uCi 9.27E-10 2.88E-08 5.7 1E-06 1.89E-06 2.1OE-06 8.03E-05 3.19E-06 1.6 1E-06 7.19E-05 2.57E-06 9.35E-07 1.92E-07 8.64E-07 1.54E-05 4.46E-06 1.08E-05 2.56E-06 2.53E-05 3.8 1E-05 1.42E-06 4.5 1E-07 9.11 E-06 9.24E-06 Critical Organ Dose Rate Factor for Elevated Release Point DFG' icoE(e) (mrem- sec) 2.03E-03 2.12E-0 I 6.24E+01 1.29E+0 I 1.72E+01 9.78E+02 3.31 E+0 1 3.63E+00 1.62E+02 1.35E+01 6.43E+00 5.58E-01 5.33E+00 1.55E+02 2.89E+0 1 2.47E+01 5.83E+00 3.08E+02 4.64E+02 3.85E+00 1.45E+00 2.27E+0 1 1.02E+02 Critical Organ Dose Rate Factor for Ground Level Release Point (mrem- sec)DFG' coE(g) ,p Ci- yr 2.92E-02 1.11 E+00 2.39E+02 7.16E+O1 8.26E+0 I 3.63E+03 1.33E+02 5.08E+01 2.27E+03 9.15E+01 3.53E+01 6.21E+00 3.19E+01 6.34E+02 1.67E+02 3.41E+02 8.11E+01 1. 14E+03 1.72E+03 4.54E+01 1.48E+01 2.90E+02 3.9 1E+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. 33 1 TABLE B.1-15 DOSE AND DOSE RATE FACTORS SPECIFIC FOR THE "ROCKS" FOR IODINE, TRITIUM, AND PARTICULATE RELEASES Critical Organ Dose Factor for Elevated Release Point (mrem)DFGioR(e) ( ,uCi Critical Organ Dose Factor for Ground Level Release Point mrem)DFGicoR(g) )Ci Critical Organ Dose Rate Factor for Elevated Release Point DFG ( mrem- sec)D o Ci-yr Critical Organ Dose Rate Factor for Ground Level Release Point mrem- sec)DFG' icoRg) (,uCi- yr Radio-nuclide H-3 Cr-51 Mn-54 Fe-59 Co-58 Co-60 Zn-65 Sr-89 Sr-90 Zr-95 Nb-95 Mo-99 Ru-103 Ag-I 10m Sb-124 1-131 1-133 Cs-134 Cs-137 Ba-140 Ce-141 Ce-144 Other*6.85E-10 2.68E-08 5.84E-06 1.74E-06 2.01 E-06 8.83E-05 3.23E-06 1.23E-06 5.48E-05 2.22E-06 8.59E-07 1.50E-07 7.74E-07 1.54E-05 4.04E-06 8.27E-06 1.95E-06 2.78E-05 4.19E-05 1.1OE-06 3.59E-07 7.02E-06 9.56E-06 6.45E-09 I .75E-07 3.18E-05 1.17E-05 1.25E-05 4.09E-04 1.80E-05 1.15E-05 5.14E-04 1.68E-05 5.79E-06 1.34E-06 5.47E-06 8.77E-05 2.80E-05 7.73E-05 1.83E-05 1.29E-04 1.94E-04 9.99E-06 3.14E-06 6.46E-05 5.09E-05 2.16E-02 1.07E+00 2.55E+02 6.78E+01 8.11E+01 3.97E+03 1.37E+02 3.88E+01 1.73E+03 8.14E+01 3.37E+01 4.92E+00 2.95E+01 6.47E+02 1.56E+02 2.61E+02 6.18E+01 1.25E+03 1.89E+03 3.56E+01 1.20E+01 2.25E+02 4.16E+02 2.03E-0 I 6.53E+00 1.3 1E+03 4.29E+02 4.79E+02 1.85E+04 7.29E+02 3.63E+02 1.62E+04 5.83E+02 2.13E+02 4.32E+01 1.96E+02 3.53E+03 1.01 E+03 2.44E+03 5.77E+02 5.80E+03 8.77E+03 3.19E+02 1.02E+02 2.05E+03 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. 33 1 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. 1.1, which limits the total fraction of ECL in liquid pathways, other than noble gases FENG (denoted here as F ) at the point of discharge from the station to the environment (see Figure B.6-1).EllNG F is limited to less than or equal to ten, i.e., F'q < 10.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 CNG , is limited to 2E-04 gCi/ml, i.e., C 1 NG < 2E-04 piCi/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 F 1 ENG AND C 1 NG 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 IEN'; and then separately determine the total concentration at the point of discharge of all dissolved and entrained noble gases from all NG station sources, denoted CG , as follows: F ENG E Y _ Cpi 10 (2-1)1 P i ECL.Ci/ml)/.zcil/ml

)and:-< 2E-04 (2-2)(gCi/ml) ([ICi/ml) (fiCi/ml)where: FI N Total fraction of ECL in liquids, excluding noble gases, at the point of discharge from the multiport difuser.B.2-1 ODCM Rev. 32 1 Cpi 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 (lCi/ml).ECLi = Effluent concentration limit (ECL) for radionuclide "i" (except for dissolved and entrained noble gases) in gtCi/ml as specified in 10 CFR 20, Appendix B, Table 2. See Appendix B for a list of ECL values.C 1 NG Total concentration at point of discharge of alldissolved and entrained noble gases in liquids from all station sources (g.Ci/ml)CING Concentration at point of discharge of dissolved and entrained noble gas "i" in liquids from all station sources (piCi/ml)2.2 Method to Determine Radionuclide Concentration for Each Liquid Effluent Source 2.2.1 Waste Test Tanks Cpi 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. 32 I 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 onceper week for determining the radioactivity released to the environment (see Table A.6.1-1).2.2.3 Steam Generator Blowdown Flash Tank The primary method to process radioactive secondary liquid from the steam generators is to direct steam blowdown flash tank bottoms cooler discharge to the floor drain tanks. If no secondary pressure is available, the steam blowdown and wet lay-ups pumps can be used. From the floor drain tanks, processing through the installed vendor resin skid (WL-SKD-135) to the waste test tanks is the preferred method. Other methods may be 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 normal recycling process (see Table A.6.1-1).2.2.4 Primary 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.2.2.5 Water Treatment System (Condensate Polishing System)The addition to the plant design of a Condensate Polishing System (CPS) for secondary side water treatment creates the potential for radiological contaminated effluents to be discharge from the Mixed Bed or Cation Demineralizers (during Megarinse) and from the Neutralization Tank (1-WT-TK-32) and the Low Conductivity Tank (1-CPS-TK-274) to the Circulating Water System via the Water Treatment System Piping following regeneration of the resin beds.Prior to discharge, each tank or vessel to be processed is sampled and analyzed for principal gamma emitters in accordance with the Liquid Sample and Analysis Program outlined in Part A to the ODCM.B.2-3 ODCM Rev. 32 I 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 II) for use when more refined results are needed. This chapter provides the methods, data, and reference material with which the operator can calculate the needed doses, dose rates and 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. 32 1 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 commitment, measured in mrem, which results from the exposure to radioactive materials that, because of uptake and deposition in the body, will continue to expose the body to radiation for some period of time after the source of radioactivity is stopped. The time frame over which the dose commitment is evaluated is 50 years. The phrases "annual dose" or "dose in one year" then 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 10 CFR Part 20. See Appendix B for a listing of these concentration limits.The quantities AD and f) 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. 32 1 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 CFR 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 can 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 H 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 H 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. 32 I 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 1I.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: Dtb = k Qi DFLitb (3-1)(mrem) = () (Ci)mrem where DFLitb Site-specific total body dose factor (mrem/ýCi) for a liquid release. It is the highest of the four age groups. See Table B.1-11.Qi Total activity (giCi) 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 ft 3/sec). For normal operations with a cooling water flow of 918 ft 3/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 ft 3/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.

13.3 -4 ODCM Rev. 32 1 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 multiport diffuser to unrestricted areas (at the edge of the initial mixing or prompt dilution zone that corresponds to a factor of 10 dilution), and 2. Any continuous or batch release over any time period 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 H 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 H calculations.

As noted in Section B.7.1, the mixing ratio associated with the edge of the 1IF surface isotherm above the multiport diffuser may be used in Method H 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. 32 1 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: Dmo = k Qi DFLimo (3-2)(mrem) = () (Ci)mrem where DFLimo Site-specific maximum organ dose factor (mrem/gCi) for a liquid release. It is the highest of the four age groups. See Table B.1-11.Q Total activity (gtCi) 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 ft 3/sec). For normal operations with a cooling water flow of 918 ft 3/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 ft 3/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. 32 1 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 II): 1. Liquid releases via the multiport diffuser to unrestricted areas (at the edge of the initial mixing or prompt dilution zone that corresponds to a factor of 10 dilution), and 2. Any continuous or batch release over any time period 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 1 approach as described in the Bases section, are also applied to Method II assessments, except that doses calculated to critical organs from radioactive effluents are evaluated for each of the four age groups to determine the maximum critical organ of an age-dependent individual via all existing exposure pathways.Table B.7-1 lists the usage factors for Method II calculations.

As noted in Section B.7.1, the mixing ratio associated with the edge of the 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.3-7 ODCM Rev. 32 1 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 IDb 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 H if a more refined calculation of 15,b is desired by the station (i.e., use of actual release point parameters with annual or actual meteorology to obtain release-specific X/Qs) or if Method I or Method IA predicts a dose rate greater than the Part A Control limit to determine if it had actually been exceeded during a short time interval.

See Part B, Section 7.2.1 for basis.Compliance with the dose rate limits for noble gases are continuously 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: 6D = 0.85

i Q DFBi (3-3a)mrem (pCi-sec) ( ,uCi) rmrem -m3 yr -UCi-m3 sec pCi-yr )where Db(e) = The off-site total body dose rate (mrem/yr) due to noble gases in elevated effluent releases, Q = the release rate at the station vents ([tCi/sec), for each noble gas radionuclide, "i", shown in Table B.I-10, and DFBi = total body gamma dose factor (see Table B.I-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. 32 1 3.4 Method to Calculate the Total Body Dose Rate from Noble Gases 3.4.1 Method I (Continued)

Itb(g) 3.4 *X(0i*DFBi) mre _pCi-sec yr PCI -rnec (3-3b)r mrem -m3ý pCi -yr )where Dtb(g)= The total off-site body dose rate (mrem/yr) due to noble gases in ground level equivalent effluent releases, and 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: DtbE(e)= 0.00 15

I (i

DFBi)(3-3c)For the Science & Nature Center, ground level effluent release: DtbE(g)= 0.0074

Y (i

DFBi)(3-3d)For the "Rocks," elevated effluent release: DtbR(e) = 0.038

Oj (i

DFBi)For the "Rocks," ground level effluent release: DtbR(g) =0.2 * (Qi

DFRl)(3-3e)(3-3f where DtbE(e), DtbE(g), DtbR(e), and DtbR(g)= 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, andand DFBi are as defined previously.

13.3 -9 ODCM Rev. 32 1 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. 1 (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. 32 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 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 Dskin 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 Oskin is 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 I)skin 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 7.22 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: Oskin(e) Qi

DFi(e)) (3-4a)mremn _ p/Ci) (mrem-sec) yr s u(i.iCi m -cyr where Dskin(e) = the off-site skin dose rate (mrem/yr) due to noble gases in an effluent discharge from an elevated release point, Q. = as defined previously, and DFl'e) the combined skin dose factor for elevated discharges (see Table B.1-l0).B.3-11 ODCM Rev. 32 1 For an off-site receptor and ground level release, the skin dose rate due to noble gases is: Dskin(g) X(.i

DFi(g)) (3-4b)where D skin(g) The off-site skin dose rate (mrem/yr) due to noble gases in an effluent discharge from a ground level release point, Q= as defined previously, and DFi'(j) = 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: DskinE(e)

=0.0014* Z (Qi

DFiE(e)) (3-4c)where DbskinE(e)

= The skin dose rate (mrem/yr) at the Science & Nature Center due to noble gases in an elevated release, Qi = as defined previously, and DFi'(e) = 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: DskinE(g)

= 0.0014* 3 (Qi

DFiE(g)) (3-4d)where IDskinE(g)

= the skin dose rate (mrem/yr) at the Science & Nature Center due to noble gases in a ground level release, Qi =as defined previously, and DFi'E(g) The combined skin dose factor for ground level discharges(see Table B. 1-13).B.3-12 ODCM Rev. 32 I For an on-site receptor at the "Rocks" and elevated release conditions, the skin dose rate due to noble gases is: OskinR(e)

= 0.0076 "'* (Qi

DFiR(e)) (3-4e)where IskinR(e)

= the skin dose rate at the "Rocks" due to noble gases in an elevated release, Q. = as defined previously, and DFi'(e) = The combined skin dose factor for elevated discharges (see Table B.l-13).For an on-site receptor at the "Rocks" and ground level release conditions, the skin dose rate due to noble gases is: OskinR(g)

= 0.0076

O (0i

DFi'R(g))

(3-4f)where lDski,(g)

= the skin dose rate (mrem/yr) at the "Rocks" due to noble gases in a ground level release, Qi = as defined previously, and DF'g = the combined skin dose factor for ground level discharges (see Table B.1-13).Equations 3-4a through 3-4f can be applied under the following conditions (otherwise, justify Method I or consider Method II).1. Normal operations (nonemergency event), and 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.3-13 ODCM Rev. 32 1 3.5.2 Method II Method II consists of the model and input data (skin dose factors) in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific data or assumptions have been identified in the ODCM. The general equation (B-9) taken from 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. 32 1 3.6 Method to Calculate the Critical Organ Dose Rate from lodines, Tritium and Particulates with TI2 Greater Than 8 Days Part A Control C.7.1.1 limits the dose rate at any time to any organ from 1311, 1331, 3 H and radionuclides in particulate form with half lives greater than 8 days to 1500 mrem/year to any organ. The Part A Control indirectly limits peak release rates by limiting the dose rate that is predicted from continued release at the peak rate. By limiting Dco to a rate equivalent to no more than 1500 mrem/year, we assure that the critical organ dose accrued in any one year by any member of the general public is less than 1500 mrem.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 Dc, 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'ico(e))

(3-5a)mrem). = P cAi)* (mrem_-sec) yr s e pCi-yr where 13 e) The off-site critical organ dose rate (mrem/yr) due to iodine, tritium, and particulates in an elevated release, Q. the activity release rate at the station vents of radionuclide "i" in gCi/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 (mrem -sec/DFG'ico(e)

= the site-specific critical organ dose rate factor m. m -yrc) for an elevated gaseous release (See Table B. 1-12).B.3-15 ODCM Rev. 32 1 For an off-site receptor and ground level release, the critical organ dose rate can be determined as follows: Dco(g) = (0, DFG'c, 0 ())(3-5b)where Qco(g)the off-site critical organ dose rate (mrem/yr) due to iodine, tritium, and particulates in a ground level release,= as defined previously, and DFG'io(g)

=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: DcoE(e)= 0.00 14

Y (Qi DFGiCOE(e))

(3-5c)where I5coE(e) = 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, Q.i = 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. 32 I For an on-site receptor at the Science & Nature Center and ground level release conditions, the critical organ dose rate is:)coE(g) = 0.0014 * (

DFG'coE(g))

(3-5d)i where IcoE(g) 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, Qi = as defined previously, and DFG'icoE(g)

= 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: DcoR(e) = 0.0076

O (Q

DFGicoR(e))

(3-5e)where co R(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(e)

= the "Rocks"-specific critical organ dose rate factor for an elevated discharge (see Table B. 1-15).For an on-site receptor at the "Rocks" and ground level release conditions, the critical organ dose rate is: IcoRfg) = 0.0076

i (Q

DFG:COR(g))

(3-5f where IbcoR and 0, are as defined previously, and DFG'icoR(g)

= the "Rocks"-specific critical organ dose rate factor for a ground level discharge (see Table B. 1-15).B.3-17 ODCM Rev. 32 I Equations 3-5a through 3-5f can be applied under the following conditions (otherwise, justify Method I or consider Method II): 1. Normal operations (not emergency event), and 2. Tritium, 1-131 and particulate releases viamonitored 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 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 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. 32 1 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 mrad in any year per unit. Dose evaluation is required at least once per 31 days.Use Method I or Method IA first to calculate the gamma air dose from the station gaseous effluent releases during the period.Use Method II if a more refined calculation is needed (i.e., use of actual release point parameter with annual or actual meteorology to obtain release-specific X/Qs), or if Method I or Method IA predicts a dose greater than the Part A Control limit to determine if it had actually been exceeded.

See Part B, Section 7.2.4 for basis.3.7.1 Method I The general form of the gamma air dose equation is: D =r = 3.17E-02 * [XIa * * -(Qi

DF') (3-6)1ri (mrad) = pCi -yr (sec,, y ) ( mrad -M3/=uCi-sec (n ) pCi -yr where Dyair is the gamma air dose.3.17E-02 is the number of pCi per gCi divided by the number of second per year,[X/Q]h is the 1-hour gamma atmospheric dispersion factor, ta is a unitless factor which adjusts the 1-hour [X/Q]y value for a release with a total duration of t hours, Qi is the total activity in gCi of each radionuclide "i" released to the atmosphere from the station gaseous effluent release point during the period of interest, and DFM is the gamma dose factor to air for radionuclide "i" (see Table B. 1-10).Incorporating receptor location-specific atmospheric dispersion factors ([X/Q]Y), adjustment factors (fa) 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. 32 1

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

Dayq) = 3.2E -07 t t-0.275* (Qi

DFiy) (3-6a)= rad pCi -yr ] mrad -m'(mrad) = ___Ci_ * ( )

E(/Ci) -p yr )b. Maximum off-site receptor location, ground-level release conditions:

Dai(g) = 1.6E- 06

t-°'2 9 3 * (Qi

DF') (3-6b)i= (ad pCiTY -yrmrad -m'.(mrad) =./.ICiim3

( )p* Z(uCi) trpd- yr J c. Science & Nature Center receptor; elevated release conditions:

DairE(e) = 4.9E-10

t 0 2 5 2 *Z(Qi*DV)

(3-6c)i (mrad) =pCi- yr )*( )E(Ci* mrad- m3 Pd Ci-m- pCi- yr d. Science & Nature Center receptor; ground-level release conditions:

DUrE(g) =4.4E- 09

t-0 3 2 1 *-(Qi*DF)

(3-6d)pCi-yr ,( )Y(PCi* mrad- m 3 P Ci- mi pCi- yr e. Receptor at the "Rocks"; elevated release conditions:

DairR(e) = 5.1 E- 09

t-0 1 5 5

X(Qi

DFJ) (3-6e)pCi-yr ,( )E(/Ci* mrad-m 3 (m ( Ci- M3 pCi- yr B.3-20 ODCM Rev. 32 I

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

DOyirR(g)

= 4.1 E- 08

t-0 2 0 4 * -(Qi

DFI) (3-6f)(mrad) =pCi- yr ).( ) Z(PCi* mrad- m 3 P Ci- mn- pCi- yr Equations 3-6a through 3-6f can be applied under the following conditions (otherwise justify Method I or consider Method ID): 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).3.7.2 Method H1 Method II consists of the models, input data (dose factors) and assumptions 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 (B-4 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.

B.3-21 ODCM Rev. 32 1 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: Dar= 3.17 E- 02* (X/Q)lhr

ta * -(Qi

DFip) (3-7)(mrad)= pCi-yr ,(sec( ) .Ci*mradm PCi-m 3)Y m) 3 pCi-yr )where D is the beta air dose, Dair 3.17E-02 is the number of pCi per tCi divided by the number of seconds per year, (X/Q)Ihr is the 1-hour undepleted atmospheric dispersion factor, ta is a unitless factor which adjusts the 1-hour X/Q value for a release with a total duration oft hours, Qi is the total activity (giCi) of each radionuclide "i" released to the atmosphere during the period of interest, and DFij 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-a) 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. 32 I

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

Dae) =4.1 E- 7

t-.3 * (Qj

DFP) (3-7a)pCi- yr ),*() Ci* mrad- m 3 pr) Ci- M3 pCi- yr b. Maximum off-site receptor location, ground-level release conditions:

DP (g) = 6.0 E- 06* t.3 1 1 * (Q* DF) (3-7b)(mrad) pCi- yr )( .mrad-m 3 paCi_dM 3) )(/Ci* pCi-yr)c. Science & Nature Center receptor; elevated release conditions:

DaIE(e) = 1.8 E- 09

t 0 3 5* j(Q

DFip) (3-7c)pC i- yr mrad- m 3 (mrad) =(y *i 3) *( ) E (/. ipCi -y )p Ci-rM3 pCi- yr d. Science & Nature Center receptor; ground-level release conditions:

DarE(g) 2.4 E-08 * *0 3 -*(Qi*DFP)

(3-7d)pCi- yr mrad- m 3 p Ci- M3 pCi- yr e. Receptor at the "Rocks"; elevated release conditions:

DairR(e) = 3.9 E- 08

t-°.2 4 9

X(Qi

DFiP) (3-7e)i3 pCi- yr mrad- m 3 (mrad)( (--m3)*( ) Z(Ci* pCi- yr)f. Receptor at the "Rocks"; ground-level release conditions:

DairR(g) =4.6 E- 07

t 0 2 6 7 *Z(Qi

DFip) (3-7f)pCi- yr mrad-m 3 p Ci_ M3 pCi- yr B.3-23 ODCM Rev. 32 Equations 3-7a through 3-7f can be applied under the following conditions (otherwise justify Method I or consider Method H): 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 H Method H consists of the models, input data (dose factors) and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific data or assumptions have been identified in the ODCM. The general equations (B-4 and B-5) taken from Regulatory Guide 1.109, and used in the derivation of the simplified Method I approach as described in the Part B Bases Section 7.2.5, are also applied to Method H 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. 32 I 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 mrem 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 mrem in a year.Use Method I or Method IA first to calculate the critical organ dose from gaseous effluent releases as it is simpler to execute and more conservative than 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=(X//Q)'depl/(X/Q)depl

t-a * (Qi

DFGiCo) (3-8)sec sec ,( (mrem)(mrem)=( (--)m/(---)( ) *Z.(puC i)* )lC mn 3 rn3 Ci where Dco is the critical organ dose from iodines, tritium, and particulates, (-/..dp is the 1-hour depleted atmospheric dispersion factor.(X/Q)dpl is the annual average depleted atmospheric dispersion.

ta is a unitless adjustment factor to account for a release with a total duration oft hours, Qi is the total activity in gCi of radionuclide "i" released to the atmosphere during the period of interest (for strontiums, use the most recent measurement), and DFGico 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 and (X/Q).a')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. 32 1

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

Dco(e) = 14.8 *t 0 1 9 7 * -(Qi* DFGico(e))

(3-8a)i (mrem)=( )*( )Z(/uCi* mrem IzCi b. Maximum off-site receptor location, ground-level release conditions:

Dco(g) = 17.7

t 0 3 1 6 * -(Qi

DFGico.g))

(3-8b)i (mrem)=()*( )(ci* mrem)/tCi c. Science & Nature Center receptor; elevated release conditions:

DcoE(e) = 3.3 E- 02

t-°0 34 9

Z(Qi

DFGicoE(e))

(3-8c)i (mrem)=()*( )Z Ci*mrem)pCi d. Science & Nature Center receptor; ground-level release conditions:

DcoE(g) = 3.3 E- 02

t-0 34 7

E(Qi

DFGicoE(g))

(3-8d)i (mrem)=()

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

DcoR(,) = 7.3 E- 02

t 0.24 8 * (Q DFGicoR(e))

(3-8e)*,mrem (mrem)=( )* ( )(ZC1 -),UCi f. Receptor at the "Rocks"; ground-level release conditions:

DcoR(g) = 8.6 E- 02

t 0 2 6 7

E(Qi

DFGicoR(g))

(3-8f)i (mrem)=()*( )Z(eCi* m )/UCi B3.3-26 ODCM Rev. 32 I Equations 3-8a through 3-8f can be applied under the following conditions (otherwise, justify Method I or consider Method H1): 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 H1 consists of the models, input data and assumptions in Appendix C of Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific data or assumptions have been identified in the ODCM (see Tables B.7-2 and B.7-3). The critical organ dose will be determined based on the location (site boundary, nearest resident, or farm) of receptor pathways, as identified in the most recent annual land use census, or by conservatively assuming the existence of all pathways (ground plane, inhalation, ingestion of stored and leafy vegetables, milk and meat) at an off-site location of maximum potential dose. Concurrent meteorology with the release period may be utilized for determination of atmospheric dispersion factors 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.3-27 ODCM Rev. 32 1 3.10 Method to Calculate Direct Dose from Site Operations 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 plant and Dry Fuel Storage (DFS) facilities 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 plant operations and DFS environmental monitoring programs, and subtracting out the dose contribution from background.

Realistic occupancy factors should be applied to the estimation of annual dose to MEMBERS OF THE PUBLIC from Seabrook site operations.

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 Science and Nature 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.

For TLD assessments, the direct dose from plant area sources and DFS operations is determined by comparing the expected annual data for environmental TLD locations near the site boundary or public access areas to pre-operational data for the same locations.

The expected measurement for each indicator TLD location is determined as a function of the observed change in exposure rate at the control location TLD data (i.e., TLD locations more than 20 km from the site) for the current year and for the pre-operational monitoring program as follows: Xe = Xp xmc PC where: Xe = the expected TLD measurement for a given location, independent of any direct radiation from station facilities, Xp = the average pre-operational TLD' measurement response for the location of interest, 1 For the DFS, the pre-operational period includes those DFS environmental monitoring program TLD measurements made prior to fuel assemblies being moved into the DFS facility.B.3-28 ODCM Rev. 32 Xmc = the average TLD measurement for the control TLD locations (TL-36 through TL-38 and TL-40 through TL-42)2 in the current year, and X = the average TLD measurement for the control TLD locations (TL-36 through TL-38 and TL-40 through TL-42) in the pre-operational period 2 All doses are expressed in mR/91days.

This is the length of a standard quarter which places all quarterly TLD field measurement on equivalent time intervals for comparisons.

The current year annual average measurement for each TLD location (Xo) is compared to the expected TLD measurement (Xe) by taking the difference between the current measurement value and the expected value (Xo -Xe). A direct dose component due to plant or DFS operations is assumed if the difference between the current annual average TLD measurement (Xo) and the calculated expected value for a TLD location (Xe) is greater than 20% (unless an evaluation can show that this difference is not plant related):[(X -Xe)/Xe > 0.2]For those measurements determined to reflect an increase due to plant or DFS operations, the net dose is estimated as the difference between the observed location value and the expected value (Xo -Xe).The 20% criterion in increased dose is selected based on its similar use as a significance criterion in NUREG-1301 (Land Use Control 3/4.12.1, Action b). The 20% increase criteria is also related to the ability of the TLDs to differentiate a true positive increase above a fluctuating background in the low dose ranges associated with environmental TLD measurements (20% is in the range of the typical 3-sigma counting statistics on the TLD readouts).

2 Control location TL-39 was not included since it has been moved from its original pre-operational location.Control locations can be added or removed from the assessment if their locations or nearby environments are altered from their baseline conditions.

B.3-29 ODCM Rev. 32 1 3.11 Dose Projections Part A Controls C.6.3.1 and C.7.4.1 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 31-day period to the appropriate dose limit for use of radwaste equipment, adjusted if appropriate for known or expected differences between past operational parameters and those anticipated for the next 31 days.3.11.1 Liquid Dose Projections The 31-day liquid dose projections are calculated by the following:

a. Determine the total body Dtb and organ dose Dmi, (Equations 3-1 and 3-2, respectively) for the last typical completed 31-day period. The last typical 31-day period should be one without significant identified operational differences from the period being projected to, such as full power operation vs. periods when the plant is shut down. For periods with identified operational differences, skip to subsection 3.11. .e. below.b. Calculate the ratio (R 1) of the total estimated volume of batch releases expected to be released for the projected period to that actually released in the reference period.c. Calculate the ratio (R 2) of the estimated gross primary coolant activity for the projected period to the average value in the reference period. Use the most recent value of primary coolant activity as the projected value if no trend in decreasing or increasing levels can be determined.

d. Determine the projected dose from: Total Body: Dtb pr = Dtb .R 1 .R2 Max. Organ: Dmo pr Dm, .RI .R2 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.6-1. 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.3-30 ODCM Rev. 32 1 For these periods outside the bounds of steps 3.11.1 .a. when significant operational differences exist from the last reference period, the projected dose to the total body Dtb and organ dose Din 0 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 Projections I1. For the gaseous radwaste treatment system, the 31-day dose projections are calculated by the following:

a. Determine the gamma air dose Dyir (Equation 3-6a), and the beta air dose Dar (Equation 3-7a) from the last typical 31-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 of curies 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: r = Dir Beta Air: Dl pr -Dfr. 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 Dco (Equation 3-8a) from the last typical 3 1-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 public are not exceeded.)*

.The last typical 3 1-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.3-31 ODCM Rev. 32 1

b. Calculate the ratio (R 4) of anticipated primary coolant dose equivalent 1-131 for the next 31 days to the average dose equivalent 1-131 level during the reference period.Use the most current determination of DE 1-131 as the projected value if no trend can be determined.

c. Calculate the ratio (R 5) 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: Dc, pr = Dc. .R4. R5 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.3-32 ODCM Rev. 32 I In place of the dose conversion factors found in Table B. 1-12, and critical organ dose equation 3-8a for Dco, Chemistry Department technical evaluation CHSTID 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 alternate 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.3-33 ODCM Rev. 32 1

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 hour0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months 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.B.3-34 ODCM Rev. 32 I 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 1, 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.6.2.1 .b, C.7.2.1 .a, C.7.2.1 .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 if the 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 [equivalent limits to 10 CFR 72.104(a) when considering Dry Fuel Storage].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, including direct radiation from the Dry Fuel Storage facility, 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 I models are derived from the methods given in Regulatory Guide 1.109, Rev. 1, October 1977.The direct radiation component from the plant and the Dry Fuel Storage facility can be determined using environmental TLD results as noted in Section B.3.10.1 (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.3-35 ODCM Rev. 32 4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP)The Radiological Environmental Monitoring Program consist of two interconnected sample collection and measurement schedules that look for environmental influences from: (1) plant operations which release to the environment radioactive materials in liquid and gaseous effluents, and direct radiation from plant facilities inside the power block Protected Area, and (2) direct radiation from used fuel placed in the Dry Fuel Storage (DFS) facility located in the West Southwest sector approximately 0.38 miles from the Containment Building.

Several monitoring locations provide data that are shared or used in the assessment of both plant and DFS operations 4.1 Plant Operations REMP The plant operations related 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.All 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 1 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 upon all measurements.

This will prevent the positive bias associated with the calculation of the mean and range based upon detectable measurements only. Secondly, the Lower Limit of Detection column will specify the LLD required by ODCM Table A.9.1-2 for that radionuclide and sample medium.4.2 Dry Fuel Storage Facility Monitoring Program The DFS radiological environmental monitoring stations are listed in Table B.4-2. The measurement locations with respect to the Seabrook site area are shown on Figures B.4-4 and B.4-7.The results of the Dry Fuel Storage environmental monitoring are summarized and reported in the Annual Radiological Environmental Operating Report.B.4-1 ODCM Rev. 34 TABLE B.4-1 RADIOLOGICAL ENVIRONMENTAL MONITORING STATIONS(a)

Exposure Pathway and/or Sample Sample Location and Designated Code Distance From Unit I Containment (km)Direction From the Plant 1. AIRBORNE (Particulate and Radioiodine)

AP/CF-01 AP/CF-02 AP/CF-03 AP/CF-04 AP/CF-05 AP/CF-07 AP/CF-08 AP/CF-09 2. WATERBORNE PSNH Barge Landing Area Harbor Road SW Boundary W. Boundary Winnacunnet H.S.(b)(PSNH Substation)

E&H Substation(b)

Georgetown Electric Light (Control)Hampton-Discharge Area Ipswich Bay (Control)Hampton-Discharge Area(b)Hampton Beach(b)Seabrook Beach Ipswich Bay (Control)(b)

Plum Island Beach (Control)(b) 2.6 2.5 1.0 1.2 4.0 5.7 3.4 21.4 ESE E SW W NNE NNW SSE SSW a. Surface b. Sediment WS-01 WS-51 SE-02 SE-07 SE-08 SE-52 SE-57 5.3 16.9 5.3 3.1 3.2 16.9 15.9 E SSE E E ESE SSE SSE 3. INGESTION a. Milk TM-15 Hampton Falls, NH (d)6.9 NW b. Fish and Invertebrates(c)

FH-03 FH-53 HA-04 HA-54 MU-06 MU-09 MU-56 MU-59 Hampton -Discharge Area Ipswich Bay (Control)Hampton -Discharge Area Ipswich Bay (Control)Hampton -Discharge Area Hampton Harbor(b)Ipswich Bay (Control)Plum Island(b)4.5 16.4 5.5 17.2 5.2 2.6 17.4 15.8 ESE SSE E SSE E E SSE SSE B.4-2 ODCM Rev. 34 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-08 Site Boundary 1.05 W TG-09 Site Boundary .97 SW TG-10 Georgetown Light 21.4 SSW 4. DIRECT RADIATION TL-1 Brimmer's 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 2.5 E Beach 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 (e) 1.2 SSW TL-I 1 Site Boundary Fence (e) 1.0 SW TL-12 Site Boundary Fence (e) 1.2 WSW TL- 13 Inside Site Boundary (e) 1.2 W TL-14 Trailer Park, Seabrook (e) 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. 34 TABLE B.4-1 RADIOLOGICAL ENVIRONMENTAL MONITORING STATIONS(a)(Continued)

Exposure Pathway and/or Sample Sample Location and Designated Code Distance From Unit 1 Containment (km)Direction From the Plant S SSW TL-23 TL-24 TL-25 TL-26 TL-27 TL-28 TL-29 TL-30 TL-31 TL-32 TL-33 TL-34 TL-35 TL-36 TL-37 TL-38 TL-39 TL-40 TL-41 TL-42 Ferry Rd., Salisbury Ferry Lots Lane, Salisbury Elm St., Amesbury Route 107A, Amesbury Highland St., S. Hampton Route 150, Kensington Frying Pan Lane, Hampton Falls Route 27, Hampton Alumni Drive, Hampton Seabrook Elementary School Dock Area, Newburyport Bow St., Exeter Lincoln Ackerman School Route 97, Georgetown (Control) (e)Plaistow, NH (Control) (e)Hampstead, NH (Control) (e)Fremont, NH (Control) (e)Newmarket, NH (Control) (e)Portsmouth, NH, (Control)(bxe)

Ipswich, MA (Control)(b)(e) 8.1 7.2 7.6 8.1 7.5 7.5 7.2 7.6 3.8 2.0 9.8 12.0 2.3 22.6 21.5 27.7 27.0 21.6 21.0 22.8 SW WSW W WNW NW NNW NNE S S NW NNW SSW WSW W WNW NNW NNE SSE B.4-4 ODCM Rev. 34 TABLE B.4-1 RADIOLOGICAL ENVIRONMENTAL MONITORING STATIONS(a)(Continued)(a) Sample locations are shown on Figures B.4-1 to B.4-6.(b) This sample location is not required by monitoring program defined in Part A of ODCM;program requirements specified in Part A do not apply to samples taken at this location.(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 after 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 of milk locations.(e) Indicates locations shared with DFS monitoring.

B.4-5 ODCM Rev. 34 TABLE B.4-2 DRY FUEL STORAGE RADIOLOGICAL ENVIRONMENTAL MONITORING STATIONS Exposure Sample Location Distance From Direction Pathway and Designated Code DFS Pad From Containment (km) DFS Pad DIRECT RADIATION TL-44 On-site, outside Science & Nature Center (1) 0.21 ESE SB-36 On-site, inside Science & Nature Center 0.24 SE TL-67 On-site, outside near Fitness Center parking (1) 0.05 S SB-35 On-site, inside Fitness Center 0.08 S TL-68 Nearby site boundary (dump) to DFS 0.45 W TL-69 Nearby site boundary (Rocks Rd) to DFS 0.47 W TL-10 Site Boundary Fence (2) 0.81 S TL-11 Site Boundary Fence (2) 0.52 SSW TL-12 Site Boundary fence (2) 0.53 WSW TL-13 Inside Site Boundary (2) 0.61 WNW TL-14 Trailer Park, Seabrook (2) 0.94 NW TL-36 Rt 97, Georgetown (Control)

(2) 22 SSW TL-37 Plaistow, NH (Control)

(2) 21 WSW TL-38 Hampstead, NH (Control)

(2) 27 W TL-39 Fremont, NH (Control)

(2) 27 WNW TL-40 Newmarket, NH (Control)

(2) 22 NNW TL-41 Portsmouth, NH (Control)

(1)(2) 22 NNE TL-42 Ipswich, MA (Control)

(1)(2) 22 SSE (1) This location is not part of the required DFS radiological monitoring program defined in Table A.9.4-1 of the ODCM.(2) Shared environmental monitoring locations for both plant REMP (see Table B.4-1) and DFS monitoring.

B.4-6 ODCM Rev. 34 FIGURE B.4-1 RADIOLOGICAL ENVIRONMENTAL MONITORING LOCATIONS WITHIN 4 KILOMETERS OF SEABROOK STATION B.4-7 ODCM Rev. 34 FIGURE B.4-2 RADIOLOGICAL ENVIRONMENTAL MONITORING LOCATIONS BETWEEN 4 KILOMETERS AND 12 KILOMETERS FROM SEABROOK STATION 0 5 KILOMiETERS f-i A, AP/CF-.07 TM-iS &SEE ENLARG2EMENT IM FIGURE 5.4-i z~ -I-A. AP/CF-05 BEACa DISCHARGE SITE SE-02 MU-06 Fli-03 HA--04 BEACH SALISBURY BEACH ATLANTIC OCEAN B.4-8 ODCM Rev. 34 FIGURE B,4-3 RADIOLOGICAL ENVIRONMENTAL MONITORING LOCATIONS OUTSIDE 12 KILOMETERS OF SEABROOK STATION 0 5 10 15 KILOMETERS NEWMARKET S RAYMOND V EPPING 0\,.~ \SEE ENLARGEMENT I N FIGIJRE 8.4-2 KINGSTON *S EXETER t I!I I I I I I N HAMPTON HARBOR I D1,ISCHARGE SITE PLAISTOW 0 I ASI, ANTIC OCEAN TG-10,&AP/CF-09 &PLUM ISLAND FH-53 SE-52 A6 WS-5I IPSWICH 0 B.4-9 ODCM Rev. 34 FIGURE B.4-4 DIRECT RADIATION MONITORING LOCATIONS WITHIN 4 KILOMETERS OF SEABROOK STATION B.4-10 ODCM Rev. 34 FIGURE B.4-5 DIRECT RADIATION MONITORING LOCATIONS BETWEEN 4 KILOMETERS AND 12 KILOMETERS FROM SEABROOK STATION B.4-11 ODCM Rev. 34 FIGURE B.4-6 DIRECT RADIATION MONITORING LOCATIONS OUTSIDE 12 KILOMETERS OF SEABROOK STATION Nw s 10 is KILOME -:ST\ , TL-40 RAYNONI I!W TL-38 A ESE QC!-AM SSW S B.4-12 ODCM Rev. 34 FIGURE B.4-7 DRY FUEL STORAGE RADIOLOGICAL ENVIRONMENTAL MONITORING STATIONS TRUE NORTH T TL-14 500 250 0 500 1090 1590 GRAPHIC SCALE: FEET ODCM Rev. 34 B.4-13 5.0 SETPOINT DETERMINATIONS Chapter 5 contains the methodology for the calculation of effluent monitor setpoints to 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 Liquid 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. l) is defined as 2E-04 liCi/mI total for all dissolved and entrained gases. Therefore, DFmin = Ci or CNG , whichever is larger. (5-3)10ECLi 2E -04 Where: DFmin 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.5-1 ODCM Rev. 33 Ci Activity concentration of each radionuclide

'T' (except noble gases) determined to be in the test tank (pCi/ml).

This includes tritium and other non gamma emitting isotopes either measured or estimated from the most recent composite analysis.Cng The sum of all dissolved and entrained noble gases identified in each test tank (pgCi/ml).

ECLi Effluent concentration limit (ECL) for radionuclide I"i (except for dissolved and entrained noble gas) in jiCi/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 (gCi/ml).Next, the available dilution flow through the discharge tunnel (Fd), or a conservative estimate for it, is divided by the minimum dilution factor (DFmin) to determine the maximum allowable discharge flow rate (Fmax) 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, Fmax -Fd DFmin Where: Fmax = 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. 33 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 (DFmin) 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.DFmin, = Y (Ci,/10ECLi)

(5-3a)Where: DFminy 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).

Qi Activity concentration of each detectable gamma-emitting radionuclide "i" in the mixture (liCi/ml).

ECLi As defined previously.

As in the determination of Fmax 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, Fmax'y, is obtained by dividing the discharge tunnel flow, Fd, by DFminy/. 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, Fmax.y, = Fd/ DFminy Where: Fmaxy 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. 33 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 (Fmax) and the maximum allowable discharge rate based on only gamma-emitting radionuclides that are present (Fmax-y) must satisfy the following:

Fm < Fmax

ftt < 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: Rsetpoint

= fl x Fd xzC)i (5-1)Fm x DFmi, Where: Rsetpoint The maximum allowable alarm/trip setpoint for an instrument response ([tCi/ml) that ensures the limiting concentration at the point of discharge is not exceeded.f, The fraction of the total contribution of ECL at the discharge point to be associated with the test tank effluent pathway, where f 2 , 1f 3 , f 4 , and f 5 , are the fractions for the Turbine Building Sump, Steam Generator Blowdown, Primary Component Cooling and Water Treatment Effluent (Condensate Polishing System) pathways contribution to the total, respectively (fl+f 2+f 3+f 4+f 5 < 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.5-4 ODCM Rev. 33 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 Ci (gCi/ml) ECLi (tCi/ml) ECLI (giCi/ml)Cs-134 2.15E-05 9E-07 9E-06 Cs-137 7.48E-05 1E-06 IE-05 Co-60 2.56E-05 3E-06 3E-05 H-3 1.50E-01 1E-03 1E-02 The minimum required dilution factor for this mix of radionuclides (including beta-emitters) is: DFmin = .' Ci 2.15E-05 +-7.48E-05

+ 2.56E-05 + 1.50E-01 = 26 10ECLi 9E -06 IE-05 3E-05 IE-02 The release flow rate (Fm) 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 (Fmax) is: Fd F m ax = --DFmin 412,000= --gpm 26= 15,846 gpm Next, the required dilution factor for only gamma emitters in the mix is: 2.15E -05 7.48E -05 2.56E -05 DFminy Z 9E-06 E -05 3E-05 The maximum allowable discharge flow rate (Fmaxy) considering only gamma emitters is given as: Fm~v Fd 412,000 Fmaxy -= F-412,00gpm

= 37,455 gpm DFminy gp1 B.5-5 ODCM Rev. 33 With the selected release rate from the test tank set at 150 gpm, and the administrative flow fraction (ftt) assumed in this example to be 0.7, the condition for the control concentration limits is met since: Fm (equal to 150) < Fmax (equal to 15,846 gpm) x ftt (set at 0.7)< Fmaxy (equal to 37,455 gpm x Ftt (set at 0.7)150 < 11092 < 26219 and the monitor response due to the mix of the gamma emitters is: Cyi (jtCi/ml)Cs-134 2.15E-05 Cs-137 7.48E-05 Co-60 2.56E-05:C .= 1.22E -04 pCi/ml Under these conditions, the alarm/trip setpoint for the liquid radwaste discharge monitor is: Rsetpoint

= fl x Fd x ICyi (5-1)Fm X DFmjy gaCi/ml ( ) gCi/ml 412,000 Rsetpoint

= 0.4 x -1 0 0 0 x 1.22E -04= 1.22E -02 jCi/ml In this example, the alarm/trip setpoint of the liquid radwaste discharge monitor can be put at 1.22E-02 ptCi/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. 33 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 (tCi/ml) f 2 (DF') ("unidentified mix ECL" (jtCi/ml))

(5-21)where: Circulating water flow rate (gpm)DF' = Flow rate pass- monitor (gpm)unidentified mix ECL = most restrictive ECL value (jiCi/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 -(fl + f 3 + f 4 + fs); 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 Alarm Monitor Setpoint = (Monitor Sepoint) (0.25)(PCi/ml)5.1.3 Steam Generator Blowdown Liquid Sample Monitor (RM-6519)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.B.5-7 ODCM Rev. 33 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 10-8 ýtCi/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.6.1-1 of the ODCM and is used to determine the setpoint.The rate-of-change alarm setpoint is calculated from: RCset = lxl0 8 9 SWF

1 (5-23)PCC (gal) = (m DCi) (gal) ( ml hr m hr Pc where: RCset The setpoint for the PCCW head tank rate-of-change alarm (in gallons per hour).lx10-s The minimum detectable activity level in the Service Water System due to a PCCW to SWS leak (ItCi/ml).

SWF = Service Water System flow rate (in gallons per hour).PCC Primary Component Cooling Water measured (decay corrected) gross radioactivity level (itCi/ml).

As an example, assume a PCCW activity concentration of lx10-5 PtCi/ml 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: RCt= I -UC' 0 1.Ox106 gph (1/lxl0-5

)ml ml RCs -= 1000 gph As a result, for other PCCW activities, the RCset which would also relate to a detection of a minimum service water concentration of 1x10-8 jCi/ml can be found from: lxl0-5xl1000gph RCset -(5-24)PCC B.5-8 ODCM Rev. 33 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 3.3-6.5.1.6 Water Treatment Liquid Effluent (CPS Rad Monitor RM-6473)The Water Treatment Liquid Effluent monitor is used to detect abnormal activity concentrations in waste liquid discharges to the Transition Structure from tank Tk-32 (CPS Neutralization Tank or equivalentl), tank Tk-274 (Low Conductivity Tank) and megarinse wash water from the Condensate Polishing System (CPS) demineralizer regeneration cycle.The CPS Rad monitor provides alarm and automatic termination of release prior to exceeding the concentration limits specified in 1OCFR20, 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 (see Section 5.1.1.1) if the total activity expected to be discharged is greater than 10 percent of ECL, as determined by the most recent grab sample isotopic analysis.

If the total activity is less than 10 percent of ECL, the setpoints of RM-6473 can be calculated as follows: High Trip Monitor Setpoint (jiCi/ml) f5 (DF') ("unidentified mix ECL" (pLCi/ml))

where: DF' Circulating water flow rate (gpm)Flow rate pass- monitor (gpm)unidentified mix ECL = most restrictive ECL value (RCi/ml) for an unidentified mixture or a mixture known not to contain certain radionuclides as given in 1OCFR20, Appendix B, Notes.f5 1 -(fl + f 2 + f 3 + f 4); where the f values are described in Section 5.1.1.1 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).' The Condensate Polishing System (CPS) has the provision to discharge temporary tanks as batch releases through the same discharge path as the Neutralization Tank (TK-32). The requirements for discharges from a temporary tank through this CPS pathway are equivalent to those of the Neutralization Tank.B.5-9 ODCM Rev. 33 Warning Alarm Monitor Setpoint (/uCi/ml)( High Trip (0.25)

Sepoint)B.5-10 ODCM Rev. 33 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 not exceeded.5.2.1 Plant Vent Wide-Range 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 jiCi/sec) is set by limiting the off-site noble gas dose rate to the total body or to the skin, and is denoted Rsetpoint-Rsetpoint is the lesser of: Rb=5 8 8 fV (5-5)DFB,/ Ci/sec =(mrem-tpCi-m) ( pCi- yr 3)yr- pCi- sec mrem-m and: 1 Rsk.= 3,000 1 fv (5-6)DF',;ptCi/sec =(mrem) ( Ci- yr yr mrem-sec where: Rtb Response of the monitor at the limiting total body dose rate (jiCi/sec) 500 (mrem-p Ci- m 3 588 (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 ýtCi (pCi/gCi)8.5E-07 = [X/Q]Y, maximum off-site long-term average gamma atmospheric dispersion factor for primary vent stack releases (sec/mi 3)B.5-11 ODCM Rev. 33 DFBc C = Composite total body dose factor (mrem-m 3/pCi-yr)_ ZQiDFBi (5-7)f= The fraction of the offsite limiting total body dose rate to be administratively assigned to the plant vent (f, < 1 -fg, where fg is the fraction of the limiting dose rate to be assigned to monitored ground level releases)Qi = The relative release rate of noble gas "i" in the mixture, for each noble gas identified or postulated to be in the off-gas (jtCi/sec)

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

Z i DF'ij(e) (5-8)DF'i(e) Combined skin dose factor for elevated release point (see Table B. 1-10)(mrem-sec/jtCi-yr)

B.5-12 ODCM Rev. 33 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.I-10 for DFBi(e) and DF 'i(e)): Q0 DFBi DF'i(e)(Ci mrem-_m 3 3 mrem-sec (_sec pCi- yr a Ci- yr Xe-138 2.52E+02 8.83E-03 1.20E-02 Kr-87 7.90E+O1 5.92E-03 1.38E-02 Kr-88 1.15E+02 1.47E-02 1.62E-02 Kr-85m 4.49E+O1 1.17E-03 2.35E-03 Xe-135 6.82E+I01 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 1.11 E-03 Xe-131m 1.15E+00 9.15E-05 5.37E-04 Xe-133m 4.67E+01 2.5 1E-04 1.12E-03 Xe- 135m 6.64E+O 1 3.12E-03 3.74E-03 B.5-13 ODCM Rev. 33 0i QDFBi j DF -(5-7)" Q, DFB, = (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 (gCi-mrem-m 3/sec-pCi-yr)

' 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+O1 + 6.64E+O1 1.OOE+03 gCi/sec 4.86 E+ 00 DFBC 1.00 E+ 03 4.86E-03 (mrem-m 3/pCi-yr)and therefore:

1 Rtb = 588 fv (5-5)DFB.(588) 1 0.7 (4.86 E- 03)8.47E+04 tCi/sec B.5-14 ODCM Rev. 33 and next;DFI' -D i DF' (5-8)' Qi ODF' = (2.52E+02)(1.20E-02)

+ (7.90E+01)(1.38E-02)

+ (1.15E+02)

(1.62E-02)

+ (4.49E+01)(2.35E-03)

+ (6.82E+01)(3.33E-03)

+ (3.23E+02)

(5.83E-04)

+ (4.13E+00)(1.1 1E-03) + (1.15E+00)(5.37E-04)

+ (4.67E+01)

(1.12E-03)" (6.64E+01)

(3.74E-03)

= 6.80E+00 (tCi-mrem-sec/sec-tCi-yr) 6.80E + 00 DF'C -1.00E + 03= 6.80E -03 (mrem -sec/,uCi -yr)and therefore:

1 Rski.= 3,000 f" (5-6)DF,'= (3,000) ( 1 ) 0.7 6.80E -03 3.09E+05 jiCi/sec The setpoint, Rsetpoint, is the lesser of Rtb 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.5-15 ODCM Rev. 33 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 jiCi/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.7.1.1 .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 condenser air evacuation monitor is the most sensitive indicator of a primary to secondary leak in the plant steam generators.

Therefore, the operational setpoints for the air evacuation monitor are typically administratively set at small fractions of the values necessary to maintain the site boundary dose rate limits of Part A Control C.7.1.1. Station procedures for determining the condenser air evacuation monitor setpoints recognize the not to be exceeded ODCM determined maximum values. Typically when primary to secondary leakage is not present, the alert setpoint is 2 to 3 times background, with the high alarm set at 2.OE+03 cpm or a higher value corresponding to 30 gallons per day of primary to secondary steam generator leakage.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 3/4Ci, (the AR-41 response value determined by HIPSTID 00-021), 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.5-16 ODCM Rev. 33 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: 1 RAE =D147 fg fgland where: RAE Release rate equivalent to the assigned fraction of the limiting offsite total body dose rate (p.Ci/sec) 147 500 (mrem -ptCi -m 3 i (1E + 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 gtCi (pCi/[tCi) 3.4E-06 Maximum off-site long-term average gamma atmospheric dispersion factor for ground level releases (sec/mi 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).1 RAE4.86E -03 (0.3) (0.7)-- 6.36E+03 pCi/sec release rate limit and for the 10,000 cfm (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/iCi 4.72E + 06cm 3/ sec= (6.36E+03)

(1.87E+08)

/ (4.72E+06)

= 2.520E + 05 cpm B.5-17 ODCM Rev. 33 Case 2: As an extension of Case I 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 same release rate limit as calculated above (i.e., 6.36E+03 iiCi/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 = RAE 1.87E + 08 cpm-cm 3/4Ci 2.36E + 04cm 3/ sec= (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: 1 RAE = 588 f, fgland DFBc where: RAE= Release rate equivalent to the assigned fraction of the limiting offsite total body dose rate ([tCi/sec) 500 (mrem -ltCi- 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/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-m3 /pCi-yr] for the limiting fuel gap activity mix)fgland Same as listed above (i.e., 0.7)fv 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 (fv) 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 fv maybe set at 1.1 RAE = 588 0.7 4.86E -03 8.47E+04 gCi/sec release rate limit B.5-18 ODCM Rev. 33 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-cm 3/gCi 2.36E + 06cm 3/ sec= (8.47E+04)

(1.87E + 08) (2.36E+04)

= 6.71E + 08 cpm 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 bf 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.5-19 ODCM Rev. 33 6.0 LIQUID AND GASEOUS EFFLUENT STREAMS, RADIATION MONITORS AND RADWASTE TREATMENT SYSTEMS Figure B.6-1 shows the liquid effluent streams, radiation monitors and the appropriate Liquid Radwaste Treatment System. Figure B.6-2 shows the gaseous effluent streams, radiation monitors and the appropriate Gaseous Radwaste Treatment System.For more detailed information concerning the above, refer to the Seabrook Station Final Safety Analysis Report, Sections 11.2 (Liquid Waste System), 11.3 (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. 28 Figure B.6-1 Liquid Effluent Streams. Radiation Monitors, and Radwaste Treatment System at Seabrook Station CVCS Letdown Diversion Equipment Drainage Equipment Leakage PAB Floor Drainage O Non-Recyclable and Misc Containment Sumps Laboratory Drainage Decontamination Water Turbine Building Sump RM Radiation Monitor O Tritium Control Release (O Secondary Side Steam Generator Blowdown O Service Water System (O Water Treatment Liquid Effluent (includes condensate polishing system)CCLT Level Transmitter ODCM61 5/26/04 B.6-2 ODCM Rev. 28 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)~Containment

, Building SsVentilators Turbine tt Building oTrWae Vacuum" " atr Vacuum Effluent Secondary TPump System CntaSnaentc Delaah Taer Gaseous Waste Processing System I[-I .~Typical of Three WatBudig -GFuxilard Vent Air I t erry-6 lant "CLooler' [-] Tank--Charcoal Beds SCompressorlrmr

~Vent Stack Degasiffier Auxiliary Building Vent Air literm^,,^.ODCM62 LEGEND: H -HEPA Filter C -Charcoal Fi RM- Radiation IMlnitor ,U B.6-3 ODCM Rev. 28 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 1OCFR50, Appendix I Subsection 11 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" (10CFR5O, 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 MethodI. Method H1 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 11 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 11 evaluations.

The 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 DFLitb (mrem/jiCi)]

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 swimming and boating near the discharge point. A nominal operating plant discharge flow rate of 918 ft 3/sec was used with a mixing ratio of 0.10.The mixing ratio of 0.10 corresponds to the minimum expected prompt dilution or near-field mixing zone created at the ocean surface directly above the multiport diffusers. (Credit for additional dilution to the outer edge of the prompt mixing zone which corresponds to the I OF surface isotherm (mixing ratio .025) can be applied in the Method 11 calculation for shoreline exposures only since the edge of this isotherm typically does not reach the shoreline receptor points during the tidal cycle. The mixing ratio for 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.13.7-1 ODCM Rev. 24 The requirements for the determination of radiological impacts resulting from releases in liquid effluents is derived from 1OCFR50, Appendix I. Section HI.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 I.Seabrook utilizes an offshore submerged multiport diffuser discharger for rapid dissipation and mixing of thermal effluents in the ocean environment.

The 22-port diffuser section of the Discharge System is located in approximately 50 to 60 feet of water with each nozzle 7 to 10 feet above the sea floor. Water is discharged in a generally eastward direction away from the shoreline through the multiport diffuser, beginning 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 I 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 hours5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months ) 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 to 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 11 aquatic food pathway dose calculations shall conservatively assume credit for prompt dilution only with an Mp = 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 assumption 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 IOCFR50, Seabrook has determine site-specific mixing ratios which factor in its plant design.B.7-3 ODCM Rev. 24 The transit time used for the aquatic food pathway was 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and for shoreline activity 0.0 hours0 days 0 hours 0 weeks 0 months . 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. 1-1 1. 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: ADtb = kQi DFLitb (mrem) 0 (Ci) m (7-1)where: DFLitb Site-specific total body dose factor (mrem/jtCi) for a liquid release. It is the highest of the four age groups. See Table B.1-11.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 ft 3/sec).Method I is more conservative than Method II in the region of the Part A dose limits because the dose factors DFLitb 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 (mrem/jiCi) was determined for each of seven organs and four age groups. The largest of these was chosen to be the maximum organ dose factor (DFLimo)for that radionuclide.

DFLimo 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 DFLimo (mrem) ( (j~i) (7-2)(mre) P .Ci B.7-4 ODCM Rev. 24 where: DFLimo = Site-specific maximum organ dose factor (mrem/gtCi) for a liquid release. See Table B.1-11.Q Total activity (liCi) 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 ft 3/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 (KG/YR) (KG/YR) (LITER/YR) (KG/YR) (KG/YR) (KG/YR) (LITER/YR) (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 Il-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: I)tb = I E+ 06 [X/Q ] DFBj (7-3)mrem/= r pCi ")(sec ( -tCi )(mrem-m 3 yr jiCi A m 3 A sec pCi-yr )where:[X/Q]y Maximum off-site receptor location long-term average gamma atmospheric dispersion factor.Q Release rate to the environment of noble gas "i' (IiCi/sec).

DFBi Gamma total body dose factor, .pi-m 3r .See Table B.I-10. (Regulatory Guide 1.109, Table B-i).Elevated and ground level gaseous effluent release points are addressed separately through the use of specific [X/Q]Y For an elevated gaseous effluent release point and off-site receptor, Equation 7-3 takes the form: Dtb(e) = (1 E+ 06) * (8.5 E- 07)* Z (0i* DFBi)(mrem' (pCi ,(sec) '(uCi .mrem-m3'yr) tm3) 3 sec pCi-yr )B.7-7 ODCM Rev. 24 which reduces to: (3-3a)Otb(e) = 0.85

Z (Qi* DFBi)mrem ) -r pCi- sec)Z ( ,u Ci) * (mrem- m 3 yr) k sec) .pCi- yr )For a ground level gaseous effluent release point and off-site receptor, Equation 7-3 takes the form: Otb(g) = (1 E+ 06) * (3.4 E- 06) *. (o,

DFBi)which reduces to: Itb(g)= 3.4 * ((i

DFBi) (3-3b)(mrem"= (pCi- sec ,u(Ci)* (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 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. 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 = 1.11 Dar + 3.17 E+ 04 X QI[X/Q] DFSj (7-4)mrem =(mrem) (mrad)(pCi-yr)

Ci (sec mrem-m yr mrad yr J/Ci-sec yr ,m3 )( 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

'T' which includes the attenuation by the outer "dead" layer of the skin.Dair = 3.17 E+ 04 0, [X/Q] DFiy (7-5)mrad _ pCi- yr (Ci) (sec)(mrad-m 3 J yr j Ci- sec ) yr )m3 n .3pCi- yr)DF.' = Gamma air dose factor for a uniform semi-infinite cloud of radionuclide "i".Now it is assumed for the definition of(X/QY) from Reference 8 that: Dfinite = Dair [X/Q ]y /[X/Q] (7-6)m rad =mrad (sec 3 yr H yr )(, m3 )ýsc and Qi = 31.54 (7-7)Ci) = Ci-sec -Z('Ci~yr) iuCi -yr) sec)B.7-9 ODCM Rev- 24 so: IDskn =1.11 1E+06 [ ]' E i

DFr (7-8)mrem .(mrem) (pCi) (sec( _Ci"(mrad-m 3 yr ) .mrad j.Cin) m.- sec pCi-yr+1E+06 X/Q EO0DFSi i (pCi)(sec)]

C -u~i (mrem -m3'uCi) im) sec ); pCi-yr Substituting atmospheric dispersion factors for an elevated gaseous effluent release point, Equation 7-8 takes the following form: Iskin(e) = [1.11

1 E+ 06

8.5 E- 07* (0

DFj)] + [1 E+ 06

8.2 E- 07* (Q

DFSj)]which yields: Dskin(e)=

[0.94 (0,

DF 1 Y)] + [0.82 Z (Qi* DFSj)]mrem) =(pCi-sec-mrem) , ('ECi *mrem- m 3 +pCi-sec uCi*mrem-m 3 yr uCi-m'-mrad sec pCi-yr ) ,uCi-m 3 i .(sec pCi-yr (7-9a)defining: DFi(e) = 0.94 DFI' + 0.82 DFS, (7-10a)Then the off-site skin dose rate equation for an elevated gaseous effluent release point is: Dskin(e) = i

DF'i(e) (3-4a)mrem r=

mrem- sec)yr ik sec pCi--yr For an off-site receptor and a ground level gaseous effluent release point, Equation 7-8 becomes: Dskin(g) [1. 11

1 E+ 06

3.4 E- 06

Z(Q 1* DF)] + [1 E+ 06

1.0 E- 05 * (0i

DFSj)]B.7-10 ODCM Rev. 24 which yields: Oskin(g) = [3.8 (

DFN')] + [10 (ii

DFS)] (7-9b)=j0i[3.8DF'+

10DFSi]i defining: DFi(g) = 3.8 DF'" + 10 DFS, (7-10b)Then the off-site skin dose rate equation for ground level gaseous effluent release points is: Dskin(g) = Z i

DFI(g) (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.l-10 must be considered.

7.2.3 Critical Organ Dose Rate from lodines, 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 f),co 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 0 gCi/sec: I)co= 3.15 E+ 07* (Q

DFGjCO) (7-12)mrem~ sec (Ci

mrem Lyr) X yr) sec U~ jCi)B.7-11 ODCM Rev. 24 Equation 7-12 is rewritten in the form:= -(i

DFG )mrem= (uCi* mrem-sec)

(7-12a)yr ksec u yr where: (7-13)DFGi.o = 3.154 E+ 07

DFGi(7 Smrem- sec (sec r om)em pCi-yr ,yr-) .PCi )The dose conversion factor, DFGico, has been developed for both elevated gaseous effluent release points and ground level gaseous effluent release points (DFGico(e) and DFGico(g)), 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 (DFG'ico(e) and DFG'ico(g)), 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: C. = *DFG co ) (3-5a)Dc(e)= i*D io(e))mrem= (,, Ci

mrem- sec)yr sec p.Ci-yr )For an off-site receptor and ground level effluent release point, the critical organ dose rate equation is: (3-5b)Dco(g) Z (Qi

DFG(~o~g))

i mremr = _ _i mrem- secj yr j sec /pCi-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 11-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 1OCFR50, 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 hours0.0088 days 0.211 hours 0.00126 weeks 2.8918e-4 months ). The CHI/Q 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 B-4 and B-5 of Regulatory Guide 1.109 with the added assumption that DYfinite = D7 [X/Q] 7 /[X/Q]: ADYe) = 3.17E+4 [X/Q]Y IQiDF(i (mrad) = (pCi- yr sec)(Ci) (mrad-m3 J (7-14)(Ci .pCi -yr where: 3.17E+04=

Number of pCi per Ci divided by the number of seconds per year.[X/Q] v Annual average gamma atmospheric dispersion factor for the receptor location of interest.Q = Number of curies of noble gas "i" released.DFYi Gamma air dose factor for a uniform semi-infinite cloud of radionuclide "i".Incorporating a unitless release duration adjustment term t-a (where "a" is a constant and "t" is the total release duration in hours), and the conversion factor for Ci to jtCi (to accommodate the use of a release rate Q in [tCi), 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 D', = 3.17E 02* [X/ r*ta*i '<iD (mrad) = pCi= Yr (sec) (,,C,

mrad-m' (3-6)(. Ci-s-ec W -.~ * (-6 For an elevated release, the equation used for an off-site receptor is: Dyair(e) =3.17E'02*[1.oE-05]*t-'

7 5*Z (Qi

DFir)which leads to: 0.t-275 (Q i r)-a Dyr(e) =3.2 E- 07 * * ( DF 1) (3-6a)i (mrad) = pCi-yr

mrad-m'Ci-M-3 * ,

pCir- yr For a ground-level release, the equation used for an off-site receptor is: Dair(g) = 3.17 E- 02 *[4.9E- 05]* t 2 9 3 (Q which leads to: Dair(g) =1.6 E- 06* t" 2 9 3.* (Qi*DF/) (3-6b)(m rad) p -m- -* y L c i

m r*_M 3 The major difference between Method I and Method II is that Method II would use actual or concurrent meteorology with a specific noble gas release spectrum to determine

[X/Q]7 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 identify the general equations, parameters and approaches to Method H-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 1OCFR50, Appendix I, Subsection B.1, which limits the estimated beta air dose in off-site unrestricted area locations.

For any noble gas release, in any period, the increment in dose is taken from Equations B-4 and B-5 of Regulatory Guide 1.109: A ir Da 3.17E-02 *X/QY_ (Q* DFLP) (7-15)i (mrad) = ( pCiYr (sec (,Ci) (mrad-m3'(/.Ci-sec)

Km~ ) ( pCi-yr )B.7-14 ODCM Rev. 24 where: DFiP= Beta air dose factors for a uniform semi-infinite cloud of radionuclide "i'T.Incorporating the term t-a into Equation 7-15 leads to: Dar = 3.17E-02

X/Qlhr

t' *Y (Q,

DFi) (3-7)(mrad) = (pCi-y. * (se j. )* mrad -m'(./PCi -s) ,m 3 )~ ()X /,i* p-_ y Where X/Qlhr = average 1-hour undepleted atmospheric dispersion factor.For an elevated release, the equation used for an off-site receptor is: D~r(e) 3.17E-02

1.3E-05

t-0 3

Y (Qi

DFP)(mrad) = pCi -yr * (sec j ( Y

mrad-rM 3 (,/ad) i-sPc ) -.3 * (3 Z ~ pCi -yr which leads to: Dal(e) = 4. 1E -07

t-0 3

Z (Qi

DFfl) (3-7a)(mrad) -pCi-yr P ( mrad-m-(mrd) ./Ci~m * ( * , ~ pCi -yr)For a ground-level release, the equation used for an off-site receptor is: D6 -- 3.17E-02

1.9E-04

t-°0 3 1 9 Y_ (Q,

DFP)(pCiY-yr * (sec) ( _M3ad-m3 (mrad) = .--sec) .m 3 (C )* i~

piyr B.7-15 ODCM Rev. 24 which leads to: D r(g) 6.OE -06

t-0 3 1 9 * (Qi

DFP) (3-7b)(mrad) = PCi -yr 3

mrad- M3* -2 /~ pCi -yr)7.2.6 Dose to Critical Organ from Jodines, 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 II-type dose assessments.

Method I may be used to show that the Part A Controls which limit off-site organ dose from gases (C.7.3.1 and C.8.1.1) have been met for releases over the appropriate periods. Part A Control C.7.3.1 is based on the ALARA objectives in IOCFR50, Appendix I, Subsection II C.Part A Control C.8.1.1 is based on Environmental Standards for Uranium Fuel Cycle in 40CFR190, which applies to direct radiation as well as liquid and gaseous effluents.

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" (lOCFR50, Appendix I). The use below of a single "critical receptor" provides part of the conservative margin to the calculation of critical organ dose in Method I. Method H 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 H 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 DFGio 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; Equation C-4 for doses associated with inhalation of all radionuclides to different organs of individuals of different age groups; and Equation C-13 for doses to organs of individuals in different age groups resulting from ingestion of radionuclides in produce, milk, meat, and leafy vegetables in Reference A). Tables B.7-2 and B.7-3 outline human consumption and environmental 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 "i" is: A Di,. = QiDFGico (7-16)where DFGico is the critical dose factor for radionuclide I"i and Qi is the activity of radionuclide

'T' released in microcuries.

Applying this information, it follows that the general form for the critical organ dose equation is: Dc. = (XhQ)ha/(X/Q)np'*

ta * (Qi

DFGiCO) (3-8)(sec' )*.Ci mrem)mremr J/ J*(@UCi

P(i Substituting specific values associated with the maximum off-site receptor location and elevated release condition yields: Dwo(e) = (1.12 E- 05)/(7.55 E- 07)

t-2 9 7

j (Q

DFGico(e))

which reduces to: Dco~e) = 14.8

tf 2 9 7* Y (Qi

DFGico(e))

(3-8a)B.7-17 ODCM Rev- 24 For the maximum off-site receptor location and ground-level release conditions, the equation is: Dco.(g) = (1.71E- 04)/(9.64 E- 06)

t 0 3 1 6", * (Qj

DFGicog)i which reduces to: Dco(g) = 17.7

t

Z (Qi

DFGio(g))

(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: I~tb 1E+ 06 [X/Q]yQ DFBi (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 hour7.75463e-4 days 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months s/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 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months each.Tour groups who schedule visits to the facility stay approximately 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . For conservatism, it was assumed that an individual in a tour group would return five times in a year, and stay 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months on each visit. These assumptions, when multiplied together, provide the occupancy factor of 12.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months s/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- 12.5hrs/yr)

_ 0.0014 8760 hrs/yr The "Rocks" -67 hrs/yr)0.0076 8760 hrs/yr substituting in the annual average gamma X/Qs:[X/Q] 7 = 1.1E-06 sec/m 3 (Science & Nature Center) for primary vent stack releases.= 5.3E-06 sec/m 3 (Science & Nature Center) for ground level releases.= 5.0E-06 sec/mr 3 (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: OtbE(e) = 0.0015

O (Qi

DFBi) (mrem/yr)

(3-3c)ItbE(g) = 0.0074 * (Oi

DFBi) (mrem/yr)

(3-3d)OtbR(e) = 0.038

Y (Qi

DFBi) (mrem/yr)

(3-3e)i ItbR(g) ý 0.2 * (0i

DFBi) (mrem/yr)

(3-3)where: DtbE(e), D DbE(g) Dt(e) and DtbR(g) total body dose rates to an individual at the Science & Nature Center and the "Rocks" (recreational site), respectively, due to noble gases in an elevated (e) and ground level (g)release, o Taken from Seabrook Station Technical Specifications (Figure 5.1-1).B.7-19 ODCM Rev. 24 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): Iskin = 1.11 1E+06 [X/QI Q 1 DFr+ (7-8)IE+06X/Q IQiDFSi i substituting in the annual average gamma X/Qs:[X/Q]y = 1. 1E-06 sec/m 3 (Science & Nature Center) for primary vent stack releases.= 5.3E-06 sec/m 3 (Science & Nature Center) for ground level release points.= 5.OE-06 sec/m 3 (The "Rocks") for primary vent stack releases.= 2.6E-05 sec/m 3 (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/m 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/m 3 (The "Rocks") for ground level release points.and multiplying by: OF = 0.00 14 (Science & Nature Center)= 0.0076 (The "Rocks")gives: DskinE(e)

=0.00 14 Q. [1.22 DFW + 1.60 DFSj ] for an elevated release point.DskinE(g)

= 0.0014 0i [5.88 DKr + 23 DFSi ] for a ground level release point.IskinR(e)

= 0.0076 0i [5.55 DFJ" + 17.0 DFSi] for an elevated release point.DskinR(g)

=0.0076 (. [28.9 DF17 + 160 DFSj ] for a ground level release point.B.7-20 ODCM Rev. 24 and the equations can be written: DskinE(e)

= 0.00 14

O (i

DFiE(e)) (3-4c)DskinE(g)

= 0.00 14

Qi

DFRE(g)) (3-4d)DskinR(e) o 0.0076

O (Q,

DFiR(e)) (3-4e)lskinR(g)

= 0.0076

O (0i

DFiR(g)) (3-4f)where: DskinE(e)

DskinE(g)

DskinR(e) and DskinR(g) 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, Oi defined previously, and DFjE(e) , DFiE(g), DFi'R(e), and DFiR(,) the combined skin dose factors for radionuclide Ti" 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 lodines, Tritium and Particulates with Half-Lives Greater Than Eight Days The equations for Dc. are derived in the same manner as in Part B, Section 7.2.2, except that the occupancy factors are also included.

Therefore:

DCoE(e) =0.0014

O (i

DFGco()) for an elevated release. (3-5c)1DdE(g) =0.0014 * (0i

DFG CoE(g)) for a ground level release. (3-5d)i I)coR(e) = 0.0076

O (Q

DFGicoR(e))

for an elevated release. (3-5e)D)coR(g) = 0.0076

O (0,

DFGicoR(g))

for a ground level release. (3-50 B.7-21 ODCM Rev. 24 where: DcoE(e), DcoE(g) DoR(e)' and DooR(g) 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,= as defined previously, and DFG:COE(e) , DFG'coE(g)

DFG'coR(e), and DFGcoR(g) 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): Dr3.17E-2*[X/Q]lrh*

t *- (0i* DF) (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: DairE(e) = 3.17 E- 02

1. l E- 05 t 0 2 1 2

0.0014 * (Q* DFY)which reduces to: DrE(e) =4.9E- 10

t-0 2 5 2 * (Q

DFD') (3-6c)(mrad) pCi-yr*( )*CY

mrad-C*itCi-m3) pCi- yr)For ground-level releases, the gamma air dose equation for the Science & Nature Center becomes: DIr(E(g) = 3.17 E- 02

1.0 E- 04 t-31 2* 0.0014 -(Q*

DFY)B.7-22 ODCM Rev. 24 which reduces to: DarE(g) = 4.4E- 09

t-0 3 2'

1 (Qj

DFI) (3-6d)(mrad) ( pCi-yr )*, (Ci* mrad-Ci*/p Ci-m3) pCi- yr)Incorporating the specific OF and atmospheric dispersion factors for the "Rocks" yields the gamma air dose equation for elevated releases: DairR(e) = 3.17 E- 02

2.1 E- 05 *

0.0076 * (Q

DFO')which reduces to: DarR(e)5.1E-09* t-1'5 5 * (Qi *DF) (3-6e)(mrad) -pCi-yr) )* (PCi* mrad- m3 p./-Ci-rM 3) pCi- yr )For ground-level releases, the gamma air dose equation for the "Rocks" becomes: DairR(g) = 3.17E- 02

1.7E- 04 t-2 0 4

0.0076 * (Q 1

DFI')which reduces to: DairR(g) = 4.1 E- 08 * -0 2 0 4 * (Qi

DF() (3-6f)i3 (mrad) -pCi-yr)Ui**

mrad- Mý. uCi- m3') pCi- yr 7.2.7.5 Beta Dose to Air from Noble Gases Method I was derived as described in Part B, Section 7.2.5. The general form of the dose equation is: Daý = 3.17 E- 02

X/unaepl*t-aW(37

"-rhE/ *ta *Z(Q*DF/)

(3-7)i where all terms in the equation are as defined in Part B, Section 7.2.5.B.7-23 ODCM Rev. 24 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: D~rE()=3.17E-02*4.OE-05

t-0 3 5 0.0014 (Q*F)which reduces to: DIrE(e) = 1 .8E- 09 t

t (Qi

DFP) (3-7c)(mrad) -C pCi- yr. ( ), i (rd- Ci-m3) pCi-yr)For ground-level releases, the beta air dose equation for the Science & Nature Center becomes: DrE(g = 3.17 E- 02

5.5 E- 04

t-0 3 4 7

0.0014*Z (Q D

DFP)which reduces to: DarE(g) = 2.4 E- 08

t-0 3 4 7

Z (Qi

DFiP) (3-7d)(mrad) j pCi-_yr ., aCi

mrd-ym 3ý,/Ci-m3) pCi- yr Incorporating the specific OF and atmospheric dispersion factors for the"Rocks" yields the beta air dose equation for elevated releases: Dr = 3.17RE-02

1.6 E- 04

t-°2 4 9

0.0076* (Q*DFPO)* irR(e) ( i 1 which reduces to: DrR(e)= 3.9 E- 08 t

t (Qi.* DFP) (3-7e)(mrad) pCi-yr () , Ci* mrad-m/,UCi- m3 pCi- yr For ground-level releases, the beta air dose equation for the "Rocks" becomes: DarR(g = 3.l7 E- 02

1.9 E- 03

t.2 6 7

0.0076 * (Qi* DFiP)i B.7-24 ODCM Rev. 24 which reduces to: DairR(g) 4.6 E- 07

t-0 2 6 7

Z (Qi* DFiO) (3-70 (mirad) pCi-yr *Z mrad- m3n (mad= -Ci-m3) ) ¢i*pCi- yr 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: D. = (X/Q)del /(X/Q)depl

t-a

OF * (Q

DFGico)i (sec/ (sec) mrem/(mrem)= --1 /1 3 1*( )*( )*Y /Ci* )imnm 3 pCi)Applying the Science & Nature Center-specific factors for elevated release conditions produces the equation: DcoE(e) = (3.72 E- 05)/(1.56 E- 06)

t-0 3 4 9

0.0014 * (Qi

DFGi6CoE(e))

which reduces to: DcoE(e) =3.3 E- 02

t 0 3 4 9 * (Qi

DFGico E(e)) (3-8c)(mrem)=()*( )*ZCi* mremi For a ground-level release, the equation for a receptor at the Science & Nature Center is: DcoE(g) = (5.21 E- 04)/(2.23 E- 05)

t-0 3 4 7 *0.0014

z (Q

DFGico E(g))i which reduces to: DcoE(g) = 3.3 E- 02

t-034* 7 (Qi

DFGico E(g)) (3-8d)(mrem)=( )*( ZpCi* mrem-)B.7-25 ODCM Rev. 24 The specific Critical Organ Dose equation for a receptor at the "Rocks" under elevated release conditions is: DcoR(,) = (1.54E- 04)/(1.61 E- 05)

t-04,* 0.0076 * (Qi

DFGio R(e))which reduces to: DcoR(e) = 7.3 E- 02

t-0.2 4 8

Z (Qi

DFGicoR~e))

(3-8e)(mrem)=()*( )* /Ci* mremj For a ground-level release, the equation for a receptor at the "Rocks" is: DcoRA(g) = (1.80 E- 03)/(1.59E-

04) *t-267

0.0076 * (Q

DFGicoR(g))

which reduces to: DcoR() =8.6 E- 02

t-0 2 6 7 * (Q

DFGico R(g)) (3-8f)(mrem)()*()

)*Z Ci* P The special receptor equations can be applied under the following conditions (otherwise, justify Method I or consider Method II): 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/M 2) 2. 2. 0.70 2. 0.70 2. 0.70 2.P Soil Surface Density (Kg/M 2) 240. 240. 240. 240. 240. 240. 240. 240.T Transport Time to User (HRS) 48. 48. 48. 48. 480. 480.TB Soil Exposure Time(a) (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(') (gm/n 3)* 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 11 dose/dose rate analyses of identified radioactivity releases of less than one year, the soil exposure time for that release may be set at 8760 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months (1 year) for all pathways.(2) For Method II dose/dose rate analyses performed for releases occurring during the first or fourth calendar quarters, the fraction of time animals are assumed to be on pasture is zero (nongrowing season). For the second and third calendar quarters, the fraction of time on pasture (FP) will be set at 1.0. FP may also be adjusted for specific farm locations if this information is so identified and reported as part of the land use census.(3) For Method II analyses, the fraction of pasture feed while on pasture may be set to less than 1.0 for specific farm locations if this information is so identified and reported as part of the land use census.(4) For all Method II analyses, an absolute humidity value equal to 5.6 (grrmm 3) shall be used to reflect conditions in the Northeast (

Reference:

Health Physics Journal, Vol. 39 (August), 1980; Page 318-320, Pergammon 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: Age Leafy Group Vegetables Vegetables Milk Meat Inhalation (kg/yr) (kg/yr) (1 /yr) (kg/yr) (m 3/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 The "Rocks" and Science & Nature Center: Age Leafy Group Vegetables Vegetables Milk Meat Inhalation (kg/yr) (kg/yr) (1/yr) (kg/yr) (m 3/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 3700.00 Infant 0.00 0.00 0.00 0.00 1400.00* Regulatory Guide 1.109 B.7-29 ODCM Rev. 24 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 .1109 (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 Receptor 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 H1 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/QY](Reference B, Section 6), and the replacement of X/Q in infinite cloud dose equations by the[X/Qfl.7.3.3 Average Atmospheric Dispersion Factors for Receptors The calculation of Method I and Method II atmospheric diffusion factors (undepleted CHI/Q, depleted CHI/Q, D/Q, and gamma CHI/Q values) utilize a methodology generally consistent with US NRC Regulatory Guide 1.111 (Revision

1) criteria and the methodology for calculating routine release diffusion factors as represented by the XOQDOQ computer code (NUREG/CR-2919).

The primary vent stack is treated as a "mixed-mode" release, as defined in Regulatory Guide 1.111. Effluents are considered to be part-time ground 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 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/Turbine 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 1 primary vent stack to "The Rocks" (244m NE) and the Science & Nature Center (488m SW) as measured from a recent site aerial photograph were used. For the site boundary, the minimum distances from the Unit 1 primary vent stack to the 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 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 Lone-Term Averaue Dispersion Factors*Primary Vent Stack Dose Rate to Individual Dose to Air Dose to Critical Organ Total Skin Critical Gamma Beta Thyroid Body Organ X/Q depleted (sec 7.5E-07 -7.5E-07 X/Q undepleted (sec) 8.2E-07 8.2E-07 DQ _ 1.5E-08**

-1.5E-08 D/Q (M (sec> 8.5E-07 8.5E-07 8.5E-07 X (QY m )m 3 )* West site boundary, 974 meters from Containment Building** Northwest site boundary, 914 meters from Containment Building B.7-33 ODCM Rev. 24 Table B.7-5 Seabrook Station Lone-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 ( sec) 1.5E-06 1.5E-06 X/Q undepleted(sec)]

1.6E-06 1.6E-06 D/Q(_2 -2.7E-08 (sec' 1.IE-06 1.LE-06 1.1E-06 X/QY/. m3 The "Rocks": (ENE -244 meters)X/Q depleted (seC- 1.6E-05 1.6E-05 (sc ~ 5- -1.7E-05 -X/Q undepleted

/-0-0 D/Q (i22 -1.1E-07 -X/Q1, sec) 5.OE-06 5.OE-06 -5.OE-06 -B.7-34 ODCM Rev. 24 Table B.7-6 Seabrook Station Lone-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/m 3 1.6 x 10-4 2.3 x 105 1.0 x 10-5 (244m ENE) (406m SW) (823m W)Depleted CHI/Q, sec/m 3 1.5 x 10-4 2.1 x 10-' 9.6 x 10-6 (244m ENE) (406m SW) (823m W)D/Q, m-2 5.1 x 10-7 1.0 x 10-7 5.1 x 10.8 (244m ENE) (406m SW) (823m W)Gamma CHI/Q, sec/m 3 2.6 x 10-' 5.3 x 10-6 3.4 x 10-6 (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 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 = Sh Ci S1i(8-1)(cps) (cps-ml)where: R Response of the monitor to radioactivity (cps).Sli Detector counting efficiency for radionuclide "i" (cps/(J.Ci/ml)).

Ci= Activity concentration of each gamma emitting radionuclide "i" in the mixture that the monitor has a response efficiency sufficient to detect (gCi/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 S 1 for Sli, where S, is the detector counting efficiency determined from the calibration procedure.

Therefore, Equation 8-1 becomes: R = Si Y Cy (8-2)(cps) =~ ~csm) (P4ci)B.8-1 ODCM Rev. 28 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: Cdi < 10 (8-3)ECLi where: Cdi = Activity concentration of radionuclide

'i" determined to be present in the mixture at the point of discharge to the environment (j.Ci/ml).

ECLi = Effluent concentration limit (ECL) for radionuclide "i" (except for dissolved and entrained noble gas) in ltCi/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 pCi/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: C -Fd(C 1+/-c)( gpm i and with equivalence ofCi = (Cy 1 + Co3i), Equation 8-4 can be written as C =Fm Cdi F. Ci Fd where: Fm = Flow rate past monitor (gpm)Fd = Flow rate out of discharge tunnel (gpm)CJ3i = Activity concentration of non gamma emitting radionuclide "i" in the mixture at the monitor for which the monitor response is inefficient to detect (piCi/ml).

Ci = 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. 28 Substituting the right half of Equation 8-4 for Cdi in Equation 8-3, and solving for Fd/Fm yields the dilution factor needed to complete Equation 8-3: DFmin < Fd >- Ca (8-5)Fm *1OECLi gpm) mi-pi)where: ECLi = Effluent concentration limit (ECL) for radionuclide "i" (except for dissolved and entrained noble gas) in piCi/ml as specified in 10 CFR 20, Appendix B, Table 2.For noble gases, a value of 2E-04 paCi/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: DFmi. < Ed (8-5)Fm The maximum allowable discharge flow rate past the monitor can be found by setting Fm to Fmax and its equivalents, i.e: Fd Fmax --DFmin Usually Fd/Fm is greater than DFmin (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:

Rsetpoint

= f x Fd x Si Z Cyi Fm x DEtin (cps) ( ) ( cps-mljlgCici (8-6) 0 (cp) () .gtCi ),MI)B.8-3 ODCM Rev. 28 or with Fmax substituted into Equation 8-6 for the maximum allowable discharge flow rate Fd ), the setpoint equation can be stated also as: DFmin)Rsetpoint

= fl X Fmax X SYS Cyi Fm where fi 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 five liquid discharge pathways is equal to or less than one (f, + f 2 + f 3 + f 4 + f 5 < 1). The five monitored liquid effluent pathways are derived from: (fl) the Liquid Waste Test Tanks, (f 2) the Turbine Building Sump, (f 3) Steam Generator Blowdown, (fQ)Primary Component Cooling, and (f 5) Water Treatment Liquid Effluent (the Condensate Polishing System).The monitoring system is designed to incorporate the detector efficiency, SI, into its software.This results in an automatic readout in ptCi/ml or [iCi/cc for the monitor response.

Since the conversion for changing cps to tiCi/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 jtCi/ml determined by the laboratory analysis.

Therefore, the setpoint calculation for the liquid waste test tank is: Rsetpoint

= f, x Fd x Y- Cyj (5-1)F. x DFmin ml ml All liquid effluent pathway monitors are similar to the Liquid Waste Test Tank Monitor in design and function as described above for ensuring the plant site's effluent concentration limits are not exceeded in combination with each other.B.8-4 ODCM Rev. 28 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. 1.1 .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 = ISgi Cmi (8-7)i (cpm) = 3cpm-cm3 where: R = Response of the instrument (cpm)Sgi Detector counting efficiency for noble gas "i" (cpm/(gCi/cm 3))Cmi = Activity concentration of noble gas "i" in the mixture at the noble gas activity monitor (gCi/cm 3)Cmi, the activity concentration of noble gas "i" at the noble gas activity monitor, may be expressed in terms of Qi by dividing by F, the appropriate flow rate. In the case of the plant vent noble gas activity monitors the appropriate flow rate is the plant vent flow rate..1 Cmi = Qi -F (8-8)F= (sci) (cm3s 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 Cmi yields: R = ESgi j , 1 (8-9)i F/cpm-_cm3 13.8-5 ODCM Rev. 28 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, Sg 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= Sg 1 (8-10)F i (cpm) = cPm-cm3- Csec 3 sc The total body dose rate due to noble gases is determined with Equation 3-3a: DItb(e) = 0.85 * (i

DFBi) (3-3a)i mremn CpCi-sec (.i mrem-m 3 I yr P=./Ci-_M3) s-eci) pCi -yr where: Dtb(e) Total body dose rate (mrem/yr)0.85 = (1.OE+06) x (8.5E-07) (pCi-sec/jLCi-m 3)1E+06 Number of pCi per ptCi (pCi/gCi)8.5E-07 = [X/Q]Y, maximum off-site average gamma atmospheric dispersion factor (sec/m 3) for primary vent stack releases Q= The relative release rate of noble gas "i" identified or postulated to be in the gas mix (jtCi/sec).

DFBi = Total body dose factor (see Table B.I-10) (mrem-m 3/pCi-yr)B.8-6 ODCM Rev. 28 A composite total body gamma dose factor, DFBc, may be defined such that: DFBC XQi i mrem -m3 pCi -yrDFBi i (8-11)pUCi~Csec )pci sec )mrem -m'pCi- yr )Solving Equation 8-11 for DFBc yields: Q 1 jDFBj DFBc Q7 ZQ (5-7)Part A Control C.7. 1.1 .a limits the dose rate to the total body from noble gases at any location at or beyond the site boundary to 500 mrem/yr. By setting I)tb equal to 500 mrem/yr and substituting DFBc for DFBi in Equation 3-3, one may solve forZQ0 at the limiting whole body noble gas dose rate: C-'iD ='Ui 588 1 DFBc (8-12)mrem-_uCi-m 3 pCi-yr yr-- pCi-- sec ), mrem-m3)Substituting this result for ZQ0 in Equation 8-10 yields Rtb, the response of the monitor at the limiting noble gas total body dose rate: R tb 588 Sg 1 F 1 DFBc (8-13)(cpm)= rmrem-pCi-m 3 (cpm-cm3 yr-pCi-sec ) pCi -sec cm 3 C pCi-yr mrem -m3)B.8-7 ODCM Rev. 28 The skin dose rate due to noble gases is determined with Equation 3-4a: Dskin(e) = L (i

DFite) (3-4a)i mrem, Pc (/~mremn-sec/yr sec uCir -cyr where:!skin(e) Skin dose rate (mrem/yr)Q 1 = As defined above.DF'i e)= Combined skin dose factor (see Table B.I-10) (mrem-sec/[tCi-yr)

A composite combined skin dose factor, DF'C, may be defined such that: DF'C * =O` : Z (1 i

DF'i(e)) (8-14)mrem-sec) (PCi) (s-Ci mrem -sec pi-yr sec ~ sec) pCi -yr)Solving Equation 8-14 for DF'c yields: ZQiDF'i (e)DF', -(5-8)(5-8 B.8-8 ODCM Rev. 28 Part A Control C.7. 1. .a 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 setting DhkJ equal to 3,000 mrem/yr and substituting DF'c for DF'i in Equation 3-4 one may solve for ZQi at the limiting skin noble gas dose rate: 3,000 (8-15)i DF'c (sAciD (mrem)( pCi-yr sec yr mrem -sec)Substituting this result for Oin Equation 8-10 yields Rskin, the response of the monitor at the limiting noble gas skin dose rate: Rskin = 3,000 Sg 1 F DF'c (cpm) (mrem] cpm -cm'/ sec' p. i, C-yr /(8-16)\yr ' .,Ci ).cm 3J mrem -sec)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 (p.Ci/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: Rtb(e) 588 (5-5)DFBc ([tCi) = (torero -RtCi -m3.( pCi -yr 3 sec yr -pCi -sec mrem -m R skin(e) 3,000 (5-6)DF'(lICi) = (mrem) ( Ci- 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. 28 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 10-8 jiCi/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. 28 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/cm 3) 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.7.1.1 .a, 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. 1-10, a conservative activity release rate from the plant vent for both whole body and skin dose rate conditions can be calculated.

From Table B.1-10, Kr-89 is seen to be the most restrictive 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: Rtb 588 1/DFB, (5-5)where: Rtb = plant vent maximum release rate (liCi/sec) based on the whole body does rate limit of 500 mrem/yr DFBc = 1.66E-02 (mrem-m 3/pCi-yr), whole body dose factor for Kr-89 588 = conversion factor (mrem-gtCi-m 3/yr-pCi-sec)

Therefore:

Rtb = 588 1/1.66E-02

= 35,421 [tCi/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 1/DF'c (5-6)where: Rskin = plant vent maximum release rate (jtCi/sec) based on skin dose rate limit of 3000 mrem/yr.DF'c = 2.45E-02 mrem-sec/jtCi-yr skin dose factor for Kr-89 3000 = Site boundary skin dose rate limit (mrem/yr)B.8-11 ODCM Rev. 28 therefore:

Rskin = 3000 (mrem/yr) 1/2.45E-02(mrem-sec/liCi-yr)

= 122,449 [tCi/sec from the plant vent Comparing the release rate limit for the whole body to that for the skin (i.e., 35,421 gCi/sec vs 122,449 [iCi/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(ltCi/sec)

= Rwg(g.Ci/cm

3) F wg(Cm 3/sec)or solving for Rwg: Rwg(gtCi/cm

3) 35,421(gtCi/sec)

/ Fwg(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 cfm)therefore:

Rwg(ptCi/cm

3) 35,421(jtCi/sec)

/ 566.4(cm 3/sec)62.5 ltCi/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. 28 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) (8-13)(cpm) = (mrem-[tCi-m 3/yr-pCi-sec) (cpm-cm 3/4.Ci) (sec/cm 3)(pCi-yr/mrem-m 3)where: Rtb(e) = 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-[tCi-m 3/yr-pCi-sec)

Sg = the detector response efficiency (cpm-cm 3/[tCi) as determined from monitor calibration.

For the air evacuation monitor, a typical value is 1.87E+08 cpm-cm 3/4iCi.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-m 3/pCi-yr).

For different gas mixes, the composite can be found from: DFBC = Y, OQDFBli 0 / 1 (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. 28 Therefore, Rtb(e) = 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: Rskin(e) = 3000 Sg (I/F) (1/DF'c) (8-16)(cpm) = (mrem/yr) (cpm-cm 3/Ci) (sec/cm 3) (gCi-yr/mrem-sec) where: Rskin(e) = count rate (cpm) for a plant vent maximum release rate based on the skin dose rate limit of 3000 mrem/yr DF'c = 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/pCi-yr).

Therefore, Rskin(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 6.71 E+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. 28 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 Gland Seal Condenser exhaust, which is a potential release pathway which the air evacuation monitor response must also take into account.For ground releases, the general equation 8-10 is used to represent the monitor count rate.R = (Sg) (1/F) ZQ 1 i (8-10)(cpm) = (cpm-cm3/4iCi) (sec/cm 3) (jtCi/sec) where: R = detector count rate (cpm)Sg = the detector efficiency (cpm-cm 3/4iCi)F = release flow rate (cm 3/sec)= the release rate of noble gas "i" in the mixture, for each noble gas listed in Table B. 1-10.For a ground release, the off-site total body dose rate is based on: Itb(g) = 3.4 Y (Q DFBi) (3-3b)i A composite total body dose factor, DFBc can be defined such that: DFBc YQi = Z(QDFBi) (8-11)i i B.8-15 ODCM Rev. 28 By substituting 8-11 into 3-3b and rearranging to solve for 1 the following equation is obtained:-0 i = (,b~tg / 3.4) (1/DEB,;)By inserting-a limiting value of 500 mrem/yr as btb(g) this simplifies to: Y0 = 147 (1/DFBJ)i Insertion of this equation into equation 8-10 yields: Rtb(g) = 147 S(g) (1/F) (1/DFBJ)(cpm) -(mrem-jiCi-m 3/yr-pCi-sec) (cpm-cm 3/g1Ci) (sec/cm 3) (pCi-yr/mrem-m 3)where: Rtb(g) = 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-[tCi-m 3/yr-pCi-sec)

Sg = the detector response efficiency for the air evacuation monitor (a typical value of 1.87E+08 cpm-cm3/[LCi is applied in this example).F release flow rate. During the hogging mode of operation, a value of 4.72E+06 cm 3/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 cfm 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 unmonitored flow, an administrative fraction (fgland) 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, fgland = 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. 28 DFBc = Composite total body dose factor which weights the combination of total body dose factors (from ODCM Table B. l-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-m 3/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 (fQ) and ground fraction (fg) is less than or equal to I (fg < 1 -fQ). The second release reduction factor (fgland) 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:

Rtb(g) = (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: Dskin(g) i E(Q jDFi(g)) (3-4b)A composite skin dose factor, DF'c(g) can be defined such that: DF'c(g) 0i : 1(QDFi~g)

(8-17)By substituting 8-17 into 3-4b and rearranging to solve for ZQ 1 the following equation is obtained: Yj = DSkin(g)0 (I/DF'(g))

i By inserting a limiting value of 3000 mrem/yr as Dskin(g) this simplifies to: Yi 3000 (l/DF'c(g))

Insertion of this equation into equation 8-10 yields: Rskin(g) = 3000 Sg (1/F) (1/DF'c(g))(cpm) = (mrem/yr) (cpm-cm3/4tCi) (sec/cm 3) (ýtCi-yr/mrem-sec)

B.8-17 ODCM Rev. 28 where: Rskin(g) Count rate (cpm) for the maximum ground release rate based on the skin dose rate limit of 3000 mrem/yr.DF'c(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 fgland 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. 28 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 1OCFR50, Appendix I", U.S.Nuclear Regulatory Commission, Revision 1, October 1977.B. Hamawi, J. N., "AEOLUS-2

-A Computer Code for the Determination of Continuous and Intermittent-Release Atmospheric Dispersion and Deposition of Nuclear Power Plant Effluents in Open-Terrain Sites, Coastal Sites, and Deep-River Valleys for Assessment of Ensuing Doses and Finite-Cloud Gamma Radiation Exposures," Entech Engineering, Inc., March 1988.C. Regulatory 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-1 ODCM Rev. 21 APPENDIX A DOSE CONVERSION FACTORS A-1 ODCM Rev. 21 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 (1 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

Uap

Mp , Z Qi *Bip

Daipj

e-A*tp 1 F The full assessment for the ODCM dose factors indicated that for i = Co-60, the maximum dose (mrem/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. ft./sec effluent flow rate for circulating water system Q 1.0 curies/year released of Co-60 assumed Bip 100 equilibrium bioaccumulation factor for Co-60 in salt water fish, in liters/kg Daipj 4.02

10- mrem/pCi.

adult GI-LLI ingestion dose factor from RG-l.109, table E-11.1.501

10 5 decay constant for Co-60 in M/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 pCi/liter.

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

Uap* Mp

Q *Bip

Daipj *ee-)*t P = 0.01032 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: Uap := 5 kg/yr usage factor Bip 1000 1/kg bioaccumulation factor all other variables the same as above therefore the dose from invertebrates is (mrem/yr):

1119.7 *

Mp Qi

Bip

Daipj

e-'tP = 0.0245 3 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 111970* Uap* Mp *W Qi *T* Daipj* e-*tp * [1- e-".tl]4 F 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 1 curie per year release of Co-60 are: Uap := 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* 10' radioactive half life in days for Co-60 Daipj : 1.70* 108 dose factor for Co-60 due to deposits in sediments, units of (mrem/hr)/(pCi/m2) tp 0.0 transit time to point of exposure, hrs tb := 131400 period that sediment is assumed to be exposed to water contamination for long term buildup, set at 15 years for Method I DCF's Q 1.0 curies per year, Co-60 assumed 111970 conversion factor to convert (Ci/yr)/(ft 3/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: 111 9 7 0* Uap*MP*W *Qi*T*Daipj*

e-'*tp* [1- e-**tb]= 0.0573 5 F 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-4 ODCM Rev. 21 a near field prompt dilution factor for the multiport 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 * , Qi

DFim 6 (mrem/yr)Fa 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*106 mrem-liters per hrs-pCi, dose factor for Co-60 for water immersion taken from HERMES.1.0* 1012 constant for pCi/Ci Therefore the swimming dose for a 1 curie release of Co-60 is (mrem/yr):

1.0* 1012

Up* MP

Q *DFim = 3.155* 10-57 Fa 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.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: 0.0103 + 0.0245 + 0.0573 = 0.0921 mrem/yr A-5 ODCM Rev. 21 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.I-1 1 (9.22 E-08 mrem/jICi) 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 II. 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 of all 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

L i

i

DFAija = Dja 8 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 Ra := 8000 the breathing rate for age group a (adults) in m 3 /yr.-9 := 7.5

10 7 the long term average depleted atmospheric dispersion factor, in Q sec/m 3 , at the maximum exposure point off site (S.B.)Qi I1 the release rate of nuclide i to the atmosphere in Ci/yr A-7 ODCM Rev. 21 DFAija := 9.67*10-6 the inhalation dose factor for nuclide i (Mn-54), organj (GI-LLI), and age group a (adult) taken from RG 1.109, table E-7, in mrem/pCi inhaled.Therefore, the inhalation dose to the maximum potential off site individual isgiven as: 3.17*104 *Ra * *LQi*DFAija=

0.00184 mrem/yr per Ci 10 PART B: GROUND PLANE DIRECT DOSE CONTRIBUTION The general equations for ground plane external direct dose in RG 1.109 are equations C-I and C-2 which together give the dose DG as: 8760*1.0*1012.

S* L Q* Qi* 1-e-~t---

DFG-11 Where for the case of Mn-54 releases, the variables in the above equation are defined as: 1.0*1012 is the number of pCi per Ci SF : 0.7 D 12 := 1.5*10-8 Q Xi := 0.8105 tb : 15 DFGii:= 5.80*10-9 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.the long term average relative deposition factor at the maximum site boundary location, in 1/m 2 is the radiological decay constant for Mn-54 (nuclide i in this case) in 1/yr.is the time in years over which accumulation is evaluated (approx.midpoint of plant operating life)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/m 2 is the unit release quantity assumed for each nuclide i, in Ci/yr.Qi 1.0 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*10'1*SF*

F *1*Q -le*1b *DFGij = 0.658 13 mrem per yr per Ci IýQý]* Q i A-9 ODCM Rev. 21 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*10 8* L

r* -Biv*-e *e-A*1h 14*i Yv

AUi P *'Ji 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*108 is the number of pCi per Ci divided by the number of hours in a year (8760).D -1.5*10.8 15 is the relative deposition factor, in 1/m 2 , at the maximum exposure point off site Q (S. B.)Q i:=1 r := 0.2 XEi := 0.00219 tb := 131400 Y := 2.0 Biý "= 2.9*102 the release rate of nuclide i to the atmosphere in Ci/yr fraction of deposited activity retained on crops, leafy vegetables, or pasture grass (1.0 for iodines)effective removal rate constant for Mn-54 from crops due to decay and weathering, in hr-I soil exposure time to deposition, in (equal to 15 yrs, or mid plant life)agricultural productivity (yield) for produce, in kg/m-2 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 xi 9.252* 10-5 radioactive decay constant for Mn-54, in hrs-l P 240 effective surface density of soil, in kg/m2 th := 1440 crop holdup time after harvest and before ingestion, in hrs te 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: 1.4,08 [ 1 F] Q Ir 1- te-it 1 -e- tb..14 r-- + Bie *

e-A*t = 67.379 16 pCi/kg QYv

AEi 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 leafy vegetables at the maximum deposition point due to a unit activity release is given as: 1.14*108*LD1*Qi*r*ei

+ B* 1- e-A,*tb

e =*th76.81117 pCi/kg 1"14108 Y *Q* r* E+Bi* A, 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: Yv 0.70 for agricultural productivity of pasture grasses, kg/m 2 te : 720 for grass exposure time to plume, hrs th := 0.0 for holdup time after harvest A-11 ODCM Rev. 21 Using these variables in the above equation gives the concentration in pasture grass as: 1.1 4*1 8 D*Q*r*1- t+ Biv

I -e -' jtl e_;-.th = 179.227 18 QI I Yv * /IEi P

Ai 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: Yv : 2.0 te : 1440 th := 2160 agricultural productivity for stored feed feed crop exposure time to plume in hrs 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.[Dl E 1-1-E"a e' i 0 t 1 1.14"10 8* [ *Qi* r*-- +Bi'~y , ei' *-Ai *teit =63.03719 pCi/kg 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: Fm

CV

QF

e'Ai *tf 20 = conc. in milk, pCi/liter where the variables are defined as: Fm := 2.5*104 QF : 6.0 tf : 2.0 average fraction of animal's daily intake of Mn-54 which appears in each liter of milk, in days/liter amount of feed consumed by a goat per day, in kg/day (50 kg/d for meat)average transport time of activity from feed into milk and to receptor, in days.A- 12 ODCM Rev. 21 Xi 2.22*10-3 decay constant of Mn-54, in days-I In addition, the C, term for the concentration of a nuclide in the animal's feed is given from RG 1.109 general equation C-11 as: Cv=fp*fs*Cp+[i-fp]*c

+fp*[1-fs]*C.,21 where the following equals: fP 0.5 fraction of the year that animals graze on pasture fs 1.0 fraction of daily feed that is pasture grass when the animal grazes on pasture CP 179.227 concentration of Mn-54 in pasture grass as calculated from above, pCi/kg C, := 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 + [1-fp]* C + fp * [1Cfs* = 121.132 pCi/kg 22 When this value of 121.132 is put back into the above general equation for nuclide concentration in milk, we get:[Cv "= 121.132 pCi/kg ]and Fm* Cv

QF

e-A*tf = 0.18123pCi/liter of Mn-54 in goats milk PART C.3.d.: Concentration in Meat: Cf Sinilar to milk, the concentration of the nuclide in animal meat is calculated.

RG 1.109 general equation C-12 is given as: Cf = Ff

Cv

QF* ez'% 24 Here the variables are set as: Ff := 8.0*10-4 fraction of animals daily intake of Mn-54 which appears in each kg of flesh, in days/kg QF 50.0 animal's daily feed intake, in kg/day A- 13 ODCM Rev. 21 20.0 average time from slaughter to consumption, in days Cv 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: Ff *Cv *QF

e-" = 4.635 25 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: Z DFIija * [ Uva

fg

C, + U.

Cm + UFa

Cf +/- UL.

fI

CL] 26 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)Uva := 520.0 vegetable ingestion rates for adults, kg/yr fg 0.76 fraction of stored vegetables grown in the garden fl 1.0 of leafy vegetables grown in the garden Uma "= 310.0 milk ingestion rate for adults, liter/yr UFa "= 110.0 meat ingestion rate for adults, kg/yr UL " 64.0 leafy vegetable ingestion rate for adults, kg/yr Cv 67.379 concentration of Mn-54 in stored vegetables, in pCiikg (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 * [Uva

fg

Cv + Uma

Cm + UFa

Cf + ULa

fl

CL] = 0.4495 27 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 of stored vegetables Dose for ingestion of goat's milk Dose for ingestion of meat Dose for ingestion of leafy vegetables DFIija *Uva *fg *Cv = 0.373 DFIija *Uma *Cm = 7.855* 104 DFIija *UFa *Cf= 0.00714 DFIija *ULa *f! *CL = 0.0688 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)Ground plane dose (Part B)Ingestion dose total (Part D)Total dose all pathways (critical organ is GI-LLI of an adult for Mn-54)0.00184 mrem/yr per Ci 0.658 mrem/yr per Ci 0.449 mrem/yr per Ci 1.11 mrem/yr per Ci A-15 ODCM Rev. 21 APPENDIX B ANNUAL AVERAGE EFFLUENT CONCENTRATION LIMITS TAKEN FROM 10 CFR 20, APPENDIX B B-1 ODCM Rev. 25 App. B App. B PART 20

STANDARDS FOR PROTECTION AGAINST RADIATION Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. I Col. 2 Col. 3 Col. 1 Col. 2-Oral Monthly Ingestion Inhalation Average Atomic Radionuclido Class ALI ALI .Air *Water Concentration No. (pCi) (pCi) (PCi/ml) (pCI/ml) (pCi/el) (pCi/el)I Hydrogen-3 Water, DAC includes skin absorption 8E+4 Gas (HT or T2)Submersloni:

Use above 8E-4 2E-5 1E-7 1E-3 1E-2 values as HT and Tz oxidize in air and in the body to HT0.2E+4 9E-6 3E-8 6E-4 6E-3 4 Berylliun-7 W, all compounds except those given for Y Y, oxides, halides, and nitrates 4 BerylI ium-1O W, see 7Be Y, see 7 Be Carbon-1l 2 Monoxide Dioxide Compounds 6 Carbon-14 Monoxide Dioxide r Compounds 9 Fluorine-18 2 D, fluorides of H. Li, Na, K, Rh, Cs, and Fr 4E64 1E+3 LLI wall (16E3).4E+5 ZE-3 5E÷4 St wall (5E+4)4E+2 4E÷3 7E+2 2E+4 8E-6 3E-8 2E+2 6E-8 ZE-10 1E+1 6E-9 2E-11 1E+6 5E-4 2E-6 6E+5 3E-4. 9E-7.4E+5 2E-4 6E-7 2E+6 7E-4 2E-6 2E+5 9E-5 3E-7 2E+3 1E-6 3E-9 7E+4 3E-5 1E-7 2E-5 6E-3 3E-5 7E-4 2E-4 6E-2 3E-4 7E-3 W, fluorides of Be, Mg, Ca, Sr. Bar- Ra,.Al; Ga, In, TI, As, Sb, B8. Fe, Ru, Os, Co, Ni, Pd, Pt, Cu, Ag, Au, Zn, Cd, Hg, Sc, Y. Ti, Zr, V, Kb, Te, Mn, Tc, and Re Y, lanthanum fluoride I.11 Sodiumr22

0. all compounds 11 Sodium-24 0, all compounds 12 Magnesuom-28 D, all compounds except those given for W W, oxides, hydroxides, carbides, halides, and nitrates 13 Aluminum-2 6 0, A)l compounds except those given for W W, oxides, hydroxides, carbides, halides, and nitrates 14 Silicon-31 D, all compounds except those given for W and Y W, oxides, hydroxides, carbides, and nitrates Y, aluminosilicate glass 14 Silicon-32 0, see 3 1 St Wd se 31 31 *Y, see Si 15 Phosphorus-32 D, all compounds except phosphates given for W W, phosphates of Zn2+, S3+, Mg 2+. Fe3+, 8i 3+, and lanthanides 32 15 Phosphorus-33

0. see 32p W: see p 9E+4 8E÷4 6E+2 5E63 4E-5 3E-5 3E-7 2E-6 1E-7 1E-7 9E-10 7E-9 6E-6 6E-5 5E-5 5E-4 2E+3 7E-7 2E-9 9E-6 9E-S 1E+3 5E-7 2E-9 4E+2 66+1 3E-8 9E-11 6E-6 6E-5 9E+1 4E-8 1E-10 9E-3 3E-4 1E-5 4E-8 .1E-4 1E-3 3E*4 3E+4 2E+3 2E-2 LLI wall (36.3) 1E*2-5E÷0 1E-5 1E-5 16-7 5E-8 2E-9 5E-8 4E-8 3E-10 2E-10 7E-12 4E-5 4E-4 6E-2 9E+2 4E-7 1E-9 9E-6 9E-5 4E÷2 2E-7 SE-10 6E+3 8E3 4E-6 1E-8 3E+3 1E-6 4E-9 BEý5 8E-4 B-2 ODCM Rev. 25 App. B PART 20 e STANDARDS FOR PROTECTION AGAINST RADIATION App .B Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Cal. 1 Col. 2 Oral Monthly Ingestion Inhalation Average Atomic Radionuclide Class ALI ALl OAC Air Water Concentration No. (pCi) (PCi) (pCi/ml) (pCi/ml).. (pCi/ml) (pCi/ml)16 Sulfur-35 Vapor 11E4 6E-6 2E-8 --0, sulfides and sulfates except those given for W 1E+4 LLI wall (8E+3)2E14 7E-6 2E-8 1E-4 1E-3 P-UL Go u3 W, elemental sulfur, sulfides of Sr, Ba, Ge, Sn, Pb. As, Sb, B1, Cu, AgJ Au, Zn, Cd. Hg, W, and Mo. Sulfates of Ca, Sr, Be, Re, As. Sb, and B8 17 Chlorine-36 0, chlorides of H, Li, Ha. K. Rb, Cs, and Fr W, chlorides of lantha-nides, Be, Mg, Ca, Sr, Be, Ra, Al..Ga, In, TI, Ge, Sn, Pb, As, Sb, BI, Fe, Ru, Os, Co, Rh, IO.NI, Pd, Pt, Cu, Ag, Au, Zn, Cd, Hg, Sc, Y,. TI, Zr, Hf, V, Nb, Ta, Cr, No. W, Mn, Tc, and Re 17 Chlorine-38 2 0, see 3 6 C0 W,Isee 3 6 C)17 Chlorine-39 2 D, see 3 6 Cl W, see 3 6 C1 i8 Argon-37 Submersion 1 18 Argon-39 Submersion 1 is Argon-41 Submersion1 19 Potassiua-40 D, all compounds 19 Potassiua-42 D, all compounds 19 Potasslum-43 D, all compounds 19 Potassium-44 2 0, all compounds 2E+3 9E-7 3E-9 -2E+3 ZE+4 St. wall (31+4)2E+4 St. wall (4E+4)3E+2 51+3 6E+3 2E+4 St. wall'4E14)3E+4 St. wall 5E+4)3E+3 Bone surf (4E+3)2E-3 8E+2 7E+3 Sf+2:41+.)9E+2 2EI3 141 wall 3Et,)2Ef4 2E+3 1E-6 3E-9 21-5 2E-4 19 Potassium-45 2 D, all compounds 20 Calcliu-41 W, all compounds 20 20 21 21 21 21 21 21 21 22 Calctam-45 Calciur47 Scandium-43 Scandlum-44m Scandium-44 Scandiua-46 Scandium-47 Scandium-48 Scandium-49 2 Titanium-44 6E+3 W, all compounds W, all compounds Y, all compounds Y, all compounds Y, all compounds Y, all compounds Y, all compounds Y, all compounds Y, all compounds 0. all compounds except those given for W and Y W, oxides, hydroxides, carbides, halides, and nitrates Y, SrT103 2E+2 4E-4 5E+4 SE+4 6E-4 4E*2 SE+3 9E+3 7E+4 IE-5 4E-3 Bone surf (4E13)2E+4 7E-2 1E+4 2E+2 3E+3 1E+3 5E+4 1E-7 ZE-5 2E-5 21-S 2E-5'2E-4 3E-6 2E-7 2E-6 4E-6 3E-5 SE-S 2E-6 4E-7 4E-7 9E-6 3E-7 51-6 10-7 1E-6 6E-7 2E-5 3E-10 6E-B 6E-8 7E-8 8E-8 6E-3 SE-7 IE-8 1E-B 7E-0 1E-8 9E-8 5E-9 3E-9 1E-9 2E-8 31E-10 4E-9 3E-4 5E-4 4U-6 6E-5 9E-5 5E-4 3E-3 SE-3 4E-5 9E-4 5E-3 7E-4 7E-3 6E-5 6E-4 21E-5 2E-4 1E-5 1E-4 1E-4 1E-3 7E-6 7E-5 5E-5 5E-4 1E-5 1E-4 4E-5 21E-9 1E-5 8E-8 31E-4 4E-4 1E-4 3E-3 1El1 SE-9 2E-11 4E-6 4E-5 3E+1 6E+O 3E-5 4E-11 2E-9 8E-12 B-3 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Atomic Radionuclide Class No.22 Titanium-45 D, see 4 4 Ti W, see 4 4 Ti Y, see Ti 23 Vanadium-47 2 0, all compounds except those given for W W, oxides, hydroxides, carbides, and halides 23 Vanadium-48 0, see 47v W, see V 23 Vanadium-49

0. see 47V Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. 1 Col. 2 0ral Monthly Ingestion Inhalation Average ALl ALI BAC Air Water Concentration (pCi) (pCi) (pCi/ml) (pCi/ml) (pCi/mi) (pCi/ml)9E-3 3E-4 1E-5 3E-B 1E-4 1E-3-4E14 iE-5 5E-8 -3E+4 1E-S 4E-B -3E+4 8E+4 3E-5 1E-7 -St. wall (3E+4) 4E-4 4E-3 1E+5 4E-5 iE-7 6E+2 1E+3 5E-7 2E-9 9E-6 9E-5 6Et2 3E-7 tE-DO -7E+4 3E+4 1E-5 LLI wall Bone surf W, see 4 7 V -2E+4 8E-6 2E-B --24 Chromiwm-48 0, all compounds except those given for W and Y 6EB3 1E+4 5Et- 2E-8 BE-5 8E-4 W, halides and nitrates 7E+3 3E-6 1BE-8 Y, oxides and hydroxides 7E-3 3E-6 1E-8 24 Chromium-45 2 0, see 48Cr 3E-4 8E+4 4E-5 iE-7 4E-4 4E-3 W, see 4Cr 1tE5 4E-5 1E-7 -.Y, see Cr 9E+4 4E-5 iE-7 -e48 Cr 4E÷4 5E+4 2E-5 6E-8 5E-4 5E-3 24 Chromiun-Si 0, see 4B8,4t E-A 2- t-rE E W,-see Cr 2E-4 1E-5 3E-8. -Y, see 48Cr -2E+4 8E-6 3E-8 -25 Manganese-512 0, all compounds except those given for W 2E+4 5E+4 2E-5 7E-8 3E-4 3E-3 W. oxides, hydroxides.

halides, and nitrates -E-4 3E-5 8E-B -25 Manganese-52m 2 D, see 5 1Mn 3E÷4 9E44 4E-5 1E-7 -St. wall 51 (4E+4) ---SE-4 5E-3 V. see Mn 1E+5 4E-5 iE-7 -25 Manganese-52. , see fln 7Ei2 IE+3 5E-7 2E-9 iE-5 lE-4 W, sees 'nn 9E+2 4E-7 iE-9 --25 Manganese-S3 D. see 511n SE+4 IE-4 SE-6 -7E-4 7E-3 Bone surf o (2E+4) -3E-8 -W. see 5 1 Mn iE+4 SE-6 2E-8 -25 Manganese-54

0. see 5 1 Mn 2E+3 9E+2 4E-7 1E-9 3E-5 3E-4 W, see 5 1 n -8E2 3E-7 1E-9 -25 Manganese-SB 0 ee 5 1 Mn 5E+3 2E-4 6E-6 2E-8 7E-5 7E-4 I, see 5 2E+4 9E-6 3E-8 --R-4 ODCM Rev. 25 App. B App. B PART 20 STANDARDS FOR PROTECTION*AGAINST RADIATION Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2. Col. 3 Col. 1 Col, 2 Oral Monthly Ingestion Inhalation Average Atomic Radionuclide Class ALl ALL DAC Air Water Concentration No. (tiCi) (pCi) (pCi/ml) (pCi/nl) (pCi/ml) (pCi/ml)26 Iron-S2 D, all compounds except those given for W 9E+2 3E13 W, oxides, hydroxides, and halides 2E+3 26 Iron-55 0, see 5 2 Fe 9E+3 2E-3 W, see 5 2 Fe -4E+3 26 Iron-S9 D, see 5 2 Fe %E+2 3E+2 W,. see 5 2 Fe 5E-2 26 Iron-60 0, see 5 2 Fe 3E#1 6E.0 W, see 5 2 Fe 21+1 27 Cobalt-55 W, all compounds except those given for Y 1E+3 3E+3 Y, oxides, hydroxides, halides, and nitrates 3E÷3 27 Cobalt-B6 W, see 5 5 Co 5E+2 3E+2 Y, see 5 5 Co 4E+2 2E+2":55 Co :8E+3 3E+3 27 Cobalt-57 W, see 5 5 Y, see 5 5 Co .4E3 7E+2 27 Cobalt-58m W, see'5 Co 6U44 9E+4 Y. see Co -6E+4 27 Cobalt-58-" W, see 5 5 Co 21E3 1Ee3 Y. see 5 5 Co lE+1 7E2 27 Cobalt-60m 2 W. see 5 5 Co0 1E-6 4E+6 St. wall Y. see 5 5 Co (1E+6) -27 Cobalt-60 W, see 5 5 Co 5E+2 2E+2 Y, see 5 5 Co 2E+2 3E+1 27 Cobalt-61 2 W: see 5 5 Co 2E+4 6E+4 Y, see 5 5 0Co 2E+4. 6E+4 27 Cobalt-62m 2 W. see 5 5 Co 4E14 2E+5 St. wall (5E+4).-Y. see 5 5 Co -2E+5 28 Nickel-56 D, all compounds except those given for W hE+3 2E-3.W, oxides, hydroxides, and carbides 1E+3 Vapor IE-. .lEe3 28 Nickel-57 0, see 5 6N1 2E#3 5E+3 W, see 5 6 N1 -I 3E+3 Vapor 6E13 1E-6 4E-9 U-5 1E-4 IE-6 3E-9 BE-7 3E-9 1E-4 1E-3 2E-6 6E-9 1E-7 5E-10 1E:B 1E-4 2f-7 7E-10 --3E-9 9E-12 4E-7 4E-6 8E-9 3E-11 1E-6 4E-9 2E-S ZE-4 1E-6 4E-9 1A-7 4E-10 6E-6 6E-5 BE-8 3E-10 -1E-6 4E-9 6E-S 6E-4* 3E17 9E-10 --4E-5 1E-7 8E-4 8E-3 3E-5 9E-8 -BE-7 2E-9 2E-5 2E-4 3E-7 1E-9 --2E-3 6E-6 --2E-2 2E-1 1E-3 4E-6 -7E-8 2E-10. 3E-6 3E'S 1E-B 5E-11 ." 3E-5 9E-8 3E-4, 3E-3 2E-5 8E-8 7E-5 ZE-7 7E-4 7E-3 6E-5 2E-7 -8E-7 3E-9 2E-5 ZE-4 BE-7 21-S SE-7 2E-9*..2E-6 7E-9 2E-5 2E-4 iA-6 4E-9 3E-6 9E-9 B-5 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average Atomic Radionuclide Class ALI ALl OAC Air Water Concentration No. (pCi) (pCi) (pCi/ml) (pCi/ml) (pCi/ml) (PCi/ml)*t L 28 Nickel-59 28 Nickel-53 28 Nickel-65 28 Nickel-66 29 Copper-60 2 29 Copper-61 29 Copper-64" 29 Copper-67 30 Zinc-62 30 Zinc-63 2 30 Zinc-6S 30 Zinc-69a 30 Zinc-69 2 30 Zinc-71m 30 Zinc-72 31 Gallium-65 2 56.D, see Mi W, see 56NHi Vapor 56N 0. see 5 6 Nl W, see NI Vapor 0, see 5 6 N1 W. se5N Vapor 0, see 5 6 N W. see 5 6 Ni 6E+21 Vapor 3E+3 0, all compounds except those given for W and Y 3E+4 9E+4 St. wall (3E+4) -W, sulfides.

halides, and nitrates -1E-5 Y, oxides and hydroxides 1E+S 0. e 6 0 Cu 1E+4 3E84 W. see 6 0 Cu Ysee 60 Cu 4E+4 see Cu 4E+4 D, see 6 0 Cu 1E84 3E+4 W. see 6 0 Cu Y. see 6CU2E+4 Y, see Cu 2E+4 D, see 6 0 Cu 5E-3 8E-3 W, see 60Cu Y, see Cu 5E+3 Y, all copoulnds 1E+3 3E+3 V, all compounds 2E+4 7E+4 St. wall (3E+4) -2E+4 4E+3 7E+3-ZE+3 9E.3 2E+3 3E-3 8E+2 8E+3 2E+4 3E-4-2E+4 4E+2 2E+3 LLI! wall 2E-6 5E-9 3E-4 3E-3 3E-6 1E-8 -8E-7 3E-9 -7E-7 2E-9 1E-4 1E-3 1E-6 4E-9 -3E-7 1E-9 -1E-5 3E-8 1E-4 18E3 1E-5 4E-8 --7E-6 2E-8 7 2E-9 --6E-6 6E-5 3E-7 9E-10 -1E-6 4E-9 -4E-5 1E-7 -4E-4 4E-3 5E-5 2E-7 -4E-5 1E-7 --.11-5 4E-8 2E-4 2E-3 2E-5 6E-8 -1E-5 SE-8 5 4E-8 ZE-4 2E-3 1E-5 3E-8 -9E-6 3E-8 -3E-6 1E-8 6E-5 6E-4 2E-6 7E-9 --2E-6 68-9 -1E-6 .4E-9 2E-5 2E-4 3E-5 9E-8 --3E-4 3E-3 1E-7 4E-10 5E-6 SE-5 3E-6 1E-8 6E-5 6E-4 6E-5 2E-7 8E-4 8E-3 7E-6 2E-8 8E-5 8E-4 5E-7 2E-9 1E-5 1E-4 7E-5 2E-7 -9E-4 9E-3 Y, all compounds Y, all compounds Y, all Compounds Yi all compounds Y, all compounds 0, all compounds except those given for W W, oxides, hydroxides, carbides, halides, and nitrates 4E+2 3E+2 4E#3 7E-3 6E+4 1E+S 6E+3 2E+4 1E+3 1E+3 5E+4 2E85 St. wall (6E+4) -2E+S 8E-5 3E-7 B-6 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average Atomic Radionuclide Class ALl ALI .Air Water Concentration No. (pC0 ) (pCi) (pCI/el) (pCi/ml) (pCi/ml) (pCI/ml)31

Gallium-66 D, see 6 5 Ga W, see 65Ga 31 Gallium-67.

D, see 6 5 Ga WU see 6 5 Ga 31 Gallium-682

1. see 6 5 Ga W, see 5 a 31 Gallium-70 2 0, see 65Ga W, see 6 5 Ga 65G8 31 Gallium-72 D. see 6 5 Ga W, see ia 31 Gallium-73 0, see 6 5 Ga W, see 6 5 Ga 32 Germanium-66 0, all compounds except those.given for W W, oxides, sulfides, and halides 32 Germanium-67 2 0, see 6 6 Ge W. see 66G.32 Germanium-68

-- , see W, see 6 6 Ge 32 Geraanium-69 0, see 6G9 W, see 66Ge 32 Germanium-71 0, see 66e W, see 6 6 Gm 32 Germanium-75 2 D.r see 6 6 Ge W, see 666, 32 Germanium-7 7 a, see 6 6 Ge W, se 66Ge 32 Gersanium-78 2 0, see 66Ge W, see 66Ge 1.E3 4E03 3E+3.7E03 10E4-I.E.4 2E+4 4E+4-SE.4 5Et4 2E+5 St. wall (7E04)2E+5 10E3 4E+3-3E+3 50E3 2E.4 2E-4 2E+4 3E+4 2E+4 3E.4 9E÷4 St. wall *(4E+4)5E-3 4E+3 1E+2 1E+4 2E+4 5E+5 40.5 4E+4 4E+4 8E-4 St. wall (7E+4)8E+4 9E43 10÷4 6E+3 2E+4 20.4 St. wall (2E+4)ZC+4 1E-6 SE-9 1E-5 .1E-4 1E-6 4E-9 --6E-6 2E-8 1E-4 1E-3 4E-6 1E-8 --2E-5 6E-8 2E04 2E-3 2E-5 7E-8 -7E-5 2k-7 -"--1E-3 1E-2 8E-5 3E-7 1E-6 SE-9 2E-5 2E-4 1E-6 4E-9 --6E-.6 2E-8 7E-5 7E-4 6E-6 2E-8 -U.-5 4E-8 3E-4 3E-3 8E-6 3E-8 -4E-5 1E-7 ---6E-4 6E-3 40-S 30-7 --2E-6 5E-9 6-5 6E-4 4E-8 1E-10 --6E-6 2E-8 2U-4 2E-3 3E-6 10-8 2E-4 6E-7 7E-3 7E-2.2E-5 6E-8.3A-5 1E-7--9E-4 9E-3 4E-5 1E-7 4E-6 10-8 1E-4 1E-3 2E-6 8E-9 -9E-6 3E-8 ---3E-4 3E-3 9E-6 3E-8 B-7 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. I Col. 2 Oral Monthly Ingestion Inhalation Average-Atomic Radionuclilde Class ALl "ALI AC Air Water Concentration ,o. (pCi) (pCO) (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml)33 Arsenic-69 2 33 Arsenic-702 33 Arsenic-71 33 Arsenic-72 33 Arsenic-73 33 Arsenic-74 33 Arsenic-76 33 Arsenic-77 W, all compounds W, all compounds W, all compounds W, all compounds W, all compounds W, all compounds W. all compounds W, all compounds 3E+4 10E5 St. wall (4E+4)1E+4 5E-4 4E03 5E+3 9E02 11-3 8E*3 2E+3 11.3 8E+2 1E3 1E+3 4E+3 5E+3 LL wall (5E+3).-.to to 33 Arsenic-78 2 W. all compounds 8E+3 34 Selenium-702 0, all compounds, except those given for W 2E+4 W, oxides. hydroxides, carbides, and elemental Se 1E04" 3WSelenla-73 2 se: 7OSe 6E+4 W, see Se 3E+4.70 34 Selenfum-73 70 see oSe H, see Se 70 34 Selenium-75

0. see 7 0 Se 5E+2 W. see Se 34 Selenium-79 D, see 7 0 Se 6E+2 W. see 70 Se 34 Selenium-81 2 0. see 704E+4 S, see 2E+4 34 SeleniLn-81 2 D, see 70Se 6E+4 St. wall W see 70S (+4)34 Selenium-83 2 D, se 70 W, see 3E+4 2E+4 4E+4 5E-s 2E-7 -6E-4 6E-3 2E-S 7E-8 2E-4 2E-3 2E-6 6E-9 5E-5 SE-4 6Eý7 2E-9 1E-5 IE-4 7E-7 2E-9 1E-4 1E-3 3E-7 1E-9 2E-5 2E-4 6E-7 2E-9 1E-5 1E4 2E-6 7E-9. --6E-5 6E-4 9E-6 3X-B 1E-4 11-3 2E-5 5E-8 1E-4 11-3 2E-S 6E-8 -6E-5 2E-7 4E-4 4E-3 6E-5 2E-7 5E-6 2E-8 4E-5 4E-4 7E-6 2E-B --3E-7 1E-9 7E-6 7E-5 3E-7 8E-10 --3E-7 10-9 8E-6 BE-5 2E-7 8E-10 3E-5 9E-8 3E-4 3E-3 3E-S 1E-7 9E-5 3E-7-10-3 1E-2 3E-4 3E-7 -5E-5 2E-7 4E-4 4E-3 5E-5 2E-7 -4E+4*2E+5 1E-S IE-4 2E-4 7E-2 6E-2 8E+2 6E-1'7E-4 7E+,4 2E.5 2E-5 1E25 7E+2 B-8 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION Table 1 Table 2 Table Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average Atomic Radionuclide Class All AAL OAC Air Water Concentration No. (pCI) (pCi) (pCi/ml) (pCi/Ml) (pCi/ml) (pCi/ml)App. B 35 Bromine-74m' 0, bromides of H, Li, Na, K, Rb, Cs, and Fr 1KE4 4E-4 2E-S 5E-8 St. wall (2E-4) --3E-4 3E-3 W, bromides of lantha-nides, Be, Mg, Ca, Sr, Ba, Ra, Al, Ga, In, TI, Ge, Sn, Pb, As, Sb, Bi, Fe, Ru, Os, Co, Rh, Ir, Ni, Pd, Pt, Cu, Ag, Au, Zn, Cd, Hg, Sc. Y, TI,.Zr, Hi, V, Nb, Ta, Mn, Tc, and Re -4E-4 2E-5 6E-8 -35 Bromine-74 2 0, see 74mBr 2EK4 7E-4 3E-5 1E-7 St. wall 74m (4E+4)- -5E-4 5E-3 W, see Br 8E+4 4E-5 1K-7 35 Bromine-75 2 0, see 7 4mBr 3E+4 SE+4 2E-5 7E-8 -St. wall w, see 7 4'%r (4E-4) -5E-4 SE-3 Wse mr5E+4 2E-5 7E-8 --35 Bromine-76 , see 4 mBr 4E-3 5E+3 2E-6 7E-9 5E-5 SE-4 , see 7 4%r 4E+3 2E-6 6E-9 r. 35 Bromine-77 0, see 7 4"'r 2KM 2E44 1E-S 3K-B 2K-4 2E3 S, see 7 4 Br E24 8KE-6 3E-8 -" Go 357 Bromine-60m.

0, se B 2KE4 2E+4 7E-6 2E- 3 KE-4 3E-3 an W; see 'ir 1E+4 6E-6 2E-8 35 Bromine-80 2 0, see 74mBr 5E+4 2E+5 8E-5 3E-7 -St. wall (9E-4) ---1K-3 1E-2 W, see 7 4"Br 2E+5 9E-5 3E-7 -35 Sromine-82 D, see 7 4 mr SE3 4KE+3 2K6 6K-9 4K-S 4E-W. so7m-4E+3 2E-6 6E-9 4-5 *E4 H, see. ibr. E+ 4K+3. 2K-B 5K- --35 Bromine-83 D, see. 74mBr SE+4 6E+4 3E-5 9E-8 -St. wall 74%f (7E+4) --9K-4 9E-3 W, see 6E+4 3E-S 9E-8 -35 Bromine-84 2 0, see 7 4'Sr 2E+4 6E44 ZE-5 BE-8 -St. wall (3E-4) -- -4K-4 4K-3 U, see 7 4Br ( 6E4 3E-5 9E-8 --36 Krypton-74 2 Submersion 1 --3E-6 1E-8 --36 Krypton-76 Submersion1 9 BE-6 4E-B -'36 Krypton-T/

2 Submersioni 4E-6 2E-8 -36 Krypton-79 Submersioni 2E-5 7E-B -36 Krypton-81 Submersion 1 7E-4 3E-6 -B-9 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Table I Table. 2. Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. I .Col. 2 Col. 3 Col. I Col. 2 Oral Monthly ingestion Inhalation Average Al ALl DAC Air Water Concentration (pCl) .(pCi) (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml)Atomic Radionuclide Class No.ul U-.a: 36 Krypton-83a 2 Submersion 1 36 Krypton-85m Submersion 1 36 Krypton-85 Submersion 1 36 Krypton-87 2 Submersion' 36 Krypton-88 Submersion1 37 Rubidium-79 2 0, all compounds* 37 Rubidium-81m 2 0, all compounds 37 Rubidin-8I D, all compounds.37 Rubidium-82m 0, all compounds 37 Rubidium-83 0, all compounds 37 Rubidium-84 0, all compounds 37 Rubidium-86 0, all compounds 37 Rubidium-87 D, all compounds 37 Rubidiuma-88 2 0, al-. compounds 37 Rubidium-89 2 0, all compounds 38 Strontiim-80 2 0, all.soluble compounds except SrTiO 3 Y, all insoluble com-: pounds and SrTl03 38 Strontium-81 2 D, see 8 0 Sr Y. see.8 SrO 38 Strontium-82 D, see 80Sr Yj see B0Sr 38 Strontium-83

0. see 8 0 Sr Y, sae 8 0 Sr 38 Strontium-85.

2 0, see 8 OSr Y.s ee Sr 38 Strontium-.5 0, :see 8 0 r Y, see soSr 38 Strontium-87m

.0, see 8 0 Sor Y, see 80Sr-E-2 5E-5 -2E-5 1E-7 -1E-4 7E-7" 5E-6 2E-8 2E-6 9E-9 4E-4 16E5 5E-5 2E-7 St. wall (6E-4) -8E-4 8E-3 2E65 3E-5 1E-4 SE-7 St. wall (3E+5) -4E-3 4E-Z 4E+4 5E*4 26-5 7E-8 5E-4 5E-3 1E64 2E+4 7E-6 2E-8 2E-4 2E-3 6E+2 11+3 4E-7 1E-9 9E-6 9E-5 5E+2 8E62 3E-7 1E-9 7E-6 7E-5+SE.2 8E62 3E-7 1E-9 7E-6 7E-5 1E+3 26E3 6E-7 2E-9 IE-S 1E-4 2E÷4 6E44 3E-5 9E-8 St. wall (3E+4) --4E-4 4E-3 4E+4 1E+5 6E-5 2E-7 St. wall (6E+4) 9E-4 9E-3 4E+3 1E+4-1E+4 3E+4 8E64 2E+4 8E+4 3E62 4E+2 LLI wall (2E+2). -2E+2 9E61 3E+3 7E+3 2E+3 4E-3 2E+5 6E+5 8E+5 3E+3 3E63 2E+3 5E64 I- 1E.5 4E+4 2E+5 5E-6 2E-8 6E-5 5E-6 2E-8 -3E-5 1E-7 3E-4 3E-5 1E-7 -2E-7 6E-10 -S -3E-6 4E-8 16-10 -3E-6 1E-8 3E-5 1E-6 5E-9 3E-4 9E-7 3E-3 4E-4 1E-6 .1E-6 4E-9 4E-5 6E-7 2E-9 5E-5 2E-7 6E-4 6E-5 2E-7 -6E-4 3E-3 3E-5 3E-4 3E-2 4E-4 6E-3 B-10 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. I Col. 2 Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average Atomic Radionuclide Class ALl ALI OAG Air Water Concentration No. (pCi) (pCi) (pCi/ml) (pCl/ml) (pCi/ml) (pCi/ml)38 Strontium-89 0, see 8 0 Sr 6E-2 8E+2 4E-7 IE-9 LLI wall (6E-2) ---E-5 Y, see 80Sr 5E-2 IE0Z 6E-8 2E-10 -App. B r, U3 cc FE 38. Strontium-90 O, see 8oSr" 3E*1 2E-*I Bone surf Bone surf (4E-1) (2E+1)Y , see 8040r S.4E+O 38 Strontium-91

0. see osr 20÷3 so Sr2E+3 6E+3 Y. see: Sr 4E+3 80 38 Strontiu.-92 D, see 8:Sr 3so3 90.3 Y, see Sr. 7E-3 39 YttriuwU86s 2 W, all Compounds except those given for Y 2E+4 6E04 Y, oxides and hydroxides 39 Yttrium-r86 W, see 86my13 33 Y. see B -3E03 39 Yttrium-87 .W see 86!. 2E+3 3E-3 Y, see " .3E-3 39 Yttriuum-88 W. see 860 1E+3 3E*2 Y. see -2E-2 39 Yttrlum-90-..

W: .see 86m 8E+3 10E4 Y, see 'Y -.1E04 39 Yttriumrgo W, see 86a 4E+2 7E02 LLI wall (SE+2)Y, see 8- 6E+2 39 Yttrium-g9n 2 W, see 8 1E.5 2E*5 Y, see 8 2E05 39 Yttrium-91 W, see 860Y 5E-2 2E02 LLI wall Y. see .(82) 111+39 Yttrium-92 W, see .86my 3E+3 9+3 Y.'. -8Ee3 39 Yttrfum-93 W, see 8Y1E03 3E+3 Y. see 8a .- 2E+3 39 Yttriua-94 2 W, see 8"my 20+4 80*4 St. wall (30.4)Y, see 86yy(3 -4 8-E+4 39 2 W, see 86mY 4E+4 2E+5 St. wall 86.y (E+4)L -Y, see 1E+5 B-1lI 8E-9 3E-11 5E-7 50-6 2E-9 6E-12 2E-6 8E-9 ZE-5 ZE-4 1E-6 5E-9 --4E-6 1E-8 4E-5 4E-4 3E-6 9E-9 2E-5 BE-8. 3E-4 3E-3 2E-S 8E-8 1E-6 5E-9 2E-5 2E-4 1E-6 5E-9 1E-6 5E-9 3E-5 3E04 1E-6 5E-9 --1E-7 .3E-10 1E-5 10-4 1E-7 3E-10 --5E-6 2E-8 1E-4 IE-3 5E-6 2E-8 --3E-7 9E710 ---71-6 7E-5 3E-7 9E-10 --1E-4 3E-7 2E-3 20-2 7E-5 2E-7. --7E-8 2E-10 --*- 0E-6 E-5 5E-8 2E-10 --4E-6 1E-8 .4E-5 4E-4 3E-6 1E-8 -1E-6 4E-9 2E-5 2E-4 1E-6 3E-9 -3E-5 1E-7 ---4E-4 4E-3 3E-5 1E-7 -..6E-5" 2E-7 ---7E-4 7E-3 6E-5 2[-7 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Table I lable 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. I Col. 2 Col. 3 Col. I Col. 2 Oral Monthly Ingestion Inhalation Average Atomic Radionuclide Class ALl '.'TLI OAC Air Water Concentration No. (pCi) (pCi) (pCi/ml) (pCi/mi) (pCi/m1) (pCi/ml)40 Zirconium-86 0, all compounds except q those given for W and Y 1E+3 4E-3 2E-6 6E-9 2E-5 ZE-4 W, oxides, hydroxides, halides, and nitrates 3E83 1E-6 4E-9 Y, carbide 2E+3 1E-6 3E-9 40 Zirconium-8 , see 8Zr 4E+3 2E-2 9E-8 3E-10 5E-5 5E-4 W, see-6 6 Zr SE-2 2E-7 7E-10 -Y, see Zr -3E82 1E-7 4E-10 -40 Zirconiuo-O9 D; see 866Zr 2E+3 4E83 1E-6 5E-9 2E-5 2E-4 W, see 8 6 Zr -2E+3 1E-6 3E-9 -Y see 86Zr 2E+3 1E-6 3E-9 -40 Zirconium-93 0,.see 8 6 rZ 1E+3 6E-0 3E-9 --Bone surf Bone surf W, see 86 (3E÷3) (2 ) --2E-1_ 4E-5 2E-4 Bone surf (6E8.) -9E-11 -Y. see 8 6 Zr 6E-1 2E-8 -, Bone surf (78+1). -9E-11 " 40 Zirconium-95 0, see 8 6 Zr 1E+3 1EZ 5E-B 2E-5 2E-4 6 .Bone surf S- (3E8 2 ) 4E-10 .Or. see 86 : Zr 4E82 2E-7 5E-10 --U- Y .e Zr- 3E+2 1E-7 4E-10 --40 Zircoqium-97 D, see 8 6 Zr 6E.2 2E+3 BE-7 .3E9 9E-6 9E-S W, se 8 6 Zr 18E3 6E-7 2E-9 -.Y see Zr 1E+3 5E-7 2E-9 -41 Niobium-S8 2 W,. all compounds except those given for Y 58E4 2E*5 9E-5 3E-7 St. wall (7E84) -1E-3 IE-*Y. oxides and hydroxides

-2E8S 9E-5 3E-7 -".41 Niobiu-89m2 W, see 8 8Nb 1E-4 4E-4 2E-S 6E-8 1E-4 1E-3 (66 min) Y, see 8 8Nb --.4E84 2E-5 5E8 -41 Niobium-B9

W, see 8 8Nb .*5E+3 2E-4 BE-6 3E-8 7E-5 7E-4 (122 min) , see -TMNb -2E+4 6E-6 2E-8 41. Niobium-g90 W, see 8-Nb 1E+3 3E+3 1E-6 4E-9 1E-5 1E-4 Y, see 8 8Nb 2E+33 1E-6 3E-9 -41 " Niobium-ln H, ee 8 W4Nb 9E+3 2E83 E8E7 3E-9 -LLI well (18E4) E- 2-4 2E-3 Y, see.8 8Nb " 2E-2 7E-8 2E-10 .-43 Hiebium-94 W, see 88 Nb 9E+2 2E42 BE-8 3E-10 1E-S 1E-4 YV see 8Nb -2E81 6E-9 2E-11 41 Nioblum-95.m W, see 8Nb 2E83 3E-3 1E-6 4E-9 --LLI wall 3 3 88 (2E83) E+ --3E-5 3E-4 Y, see 8 Nb -2E-3 9E-7 3E-9 --B-12 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Atomic Radionuclide Class No.41 Niobium-95 W, see 8Nb Y, see Nb 41 Njobium-96 W, see asNb Y. see aNb 41 Niobium-97 2 W, see h88Nb Y, see -Nb 41 Niobium-982 2§6 wY, see 42 Molybdenun-90

0. all compounds except those given for Y Y, oxides, hydroxides, and MoS2 42 Molybdenum-93a D. see nMO Y, see 'Mo 42 Molybdenum-93 0, see 9-k Y, se 42 Molybdenum-g9 0, see 90,o Y, see 90No 42 Molybdenum-101 2 O, see 90 e Y. see 90MNo 43 Technetium-93W D, all compounds except those given for W W. oxides, hydroxides, halides, and nitrates 43 Technetliu-93 0 see 9 T3Tc V, see 43 Technetiimi94m 2 0, see9_0 Tc 43 Technetium-94 0, sea 9_Tc W. see --ic w; see 43 Technetiuw-9S D, seea Tc W, see 43 Technetium-96W 2 0, see 93 W. see 43 Technetium-96 0, see 9.Tc W, see 9 Pc 43 Technetium-gIm 0, see 93.Tc W, see 93MTc Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Co .1 Col. 2 Col. 3 Col. I Col. 2 Oral Monthly Ingestion Inhalation Average ALl ALl SAC Air Water Concentration (pCi) (pCi) (pCi/ml) (pCi/ml). (pCi/ml) (pCi/ml)2E-3 1E*3 5E-7 2E-9 3E-5 3E-4 I.E-3 SE-7 2E-9 1E+3 3E-3 1E-6 4E-9 2E-5 2E-4 2E-3 1E-6 3E-9 2E+4 8E+4 3E-5. IE-7 3E-4 3E-3 7E+4 3E-5. IE-7 1E.4 SE+4 2E-S 8E-8 2E-4 2E-3 5E+4 2E-5 7E-8 4E+3 7E+3 3E-6 1E-8 3E-5 3E-4 2E+3 5E+3 ZE-6 6E-9 9E+3 2E+4 7E-6 2E- 6EE-S 6E-4 4E-3 1E"4 6E-6 2E-8 4E+3 5E+3 2E-6 8E-9 5E-5 SE-4.2E+4 2E-2 8E-8 2E-10 2E+3 3E+3 1E-6 4E-9 *LLI wall (1E+3) --2E-5 2E-4 1E+3 1E+3 6E-7 2E-9 --4E+4 IE.5 6E-5 2E-7 --St. wall (5E.4) ---7E-4 7E-3 I E-5 6E-5 .2E-7 -7E+4 2E+5 6E-5 2E-7 AE-3 1E-2 3E+5 1E-4 4E-7 3E+4 7E+4 3E-5 1E-7 4E-4 .4E-3 1E+5 4E-5 1E-7 --2E+4 4E-4 2E-5 6E-S 3E-4 3E-3-S6E4 2E-5 BE-8 -9E+3 2E+4 BE-6 3E-8 1E-4 I£-3 2E+4. 1E-5 3E-8 4E-3 5E+3 2E-6 8E-9 5E-S 5E-4.ZE+3 8E-7 3E-9 -1E54 2E+4 SE-6 3E-8 1EA4 1E-3-2E+4 SE-6 3E-8 --2E+5 3E+5 1E-4 4E-7 2E-3 2E-2 2E+5 1E-4 3E-7 --ZE+3 3E-3 1E-6 5E-9 3E-S 3E-4 2E*3 9E-7 3E-9 5E+3 7E+3 3E-6 6E-5 6E-4 St. wall (7E+3) -1E-8 1E+3 SE-7 2E-9 B-13 ODCM Rev. 25 App. B PART 20. STANDARDS FOR PROTECTION AGAINST RADIATION App. B Atomic Radionuclide Class ALI No. .(ptC)43 Technetiuw-97 D, see 93Te 4E-4 W, see93T 43 Technetium-98 D, see 93'Tc 1E+3 W, see 93mTr 43 Technetium-99m U, see 9 3 MT 8E+4 W, see 9 3 Tc 43 Technetium-99 D, see 93mTC 4E+3 W, see 93mTc 43 Technetium-101 2 D, see 9 3 m"Tc 9E+4 St. wall W, see 93.T¢ (1E+5)43 Technetium-104 2 0, see 9 3"Te 2E+4 St. wall W, see 93M(34)44 Ruthenium-94 2 D, all compounds except those given for W and Y 2E+4 W, halides Y, oxides and hydroxides 44 Ruthenium-97 D, see H94Ru 8E-3 W, see 94Ru 94 Y, see Ru 44 Ruthenium-103 0, see 9 4 RU 2E" W 94 Ru2E-3 W, see 9Ru Y, see Ru 44 Ruthenium-105 0 see 9 4 Ru 5E-3 W, see 9 4 Ru Y, see 9 4 Ru -44 Ruthenium-i06 0, see 9 4 Ru 2E+2 LLI wall (2E-2)W, see 9 4 Ru m Y, see 94Ru.45 Rhodium-99m D, all compounds except those given for W and Y 2E+4 W, halides Y, oxides and hydroxides

-45 Rhodiumr99 D0 see 9 9mh 2E+3 W, s 9R Y, see9 Rh Table I Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col, 2 Col. 3 Col. 1 Col. 2 Monthly on Inhalation Average AE A Air Water Concentration (pCi) (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml).5E+4 2E-9 7E-B 5E-4 5E-3 6E+3 2E-6 BE-9 --2E+3 7E-7 2E-9 1E-5 1E-4 3E+2 1E-7 .4E-10 -2E+5 6E-5 2E-7 1E-3 1E-2 2E-5 1E-4 3E-7 5E+3 ZE-6 6E-5 6E-4 St. wall (6E+3) -8E-9 7E÷2 3E-7 9E-10 3E+5 1E-4 5E-7---2E-3 2E-2 4E+5 2E-4 5E-7 --7E+4 3E-5 1E-7 --4E-4 4E-3 9E-4 4E-5 iE-7 --4E-4 6E+4 6E-4 2E+4 1E+4 1E+4 2E+3 IE-3 6E+2 1E,4 1E.'4 1E+4 911-1 5E+1 1E.1 6E+4 BE+4 7E+4 3E+3 2E+3 2E+3 2E-5 6E-8 2E-4 2E-3 3E-5 9E-8 --2E-5 8E-8 -8E-6 3E-8 iE-4 1E-3 SE-6 2E-8 -SE-6 ZE-8 -7E-7 2E-9 3E-5' 3E-4 4E-7 1E-9 -3E-7 9E-10 -BE-6 2E-8 7E-5 7E-4 6E-.6 2E-8 --SE-6 ZE-8 --4E-8 iE-10 ----3E-6 3E-5 2E-8 BE-11 -" -SE-9 2E-11 --2E-5 BE-8 2E-4 2E-3 3E-S IE-7 --3E-5 9E-8 --IE-5 4E-9 3E-5 3E-4 9E-7 3E-9 -BE-7 3E-9 ODCM Rev. 25 B-14 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B t f[*Table Occupationa Col. 1 Col. 2 Oral Ingestion In Atomic Radionuclide Class ALl AlI Mo. (tci) (McN )45 Rhodium-100 0, see 9 9mRh 4-53 W: see 9 4-RhE 4+3 Y see 9 9 Rh 4E+3 Y ee Rh 6E÷3 1E+4 45 Rhodium-l0 , 0, see 9 83 E Ysee 9 9"Rh -8E÷3 Y:S:99-:hB+

45 Rhodium-101 D, see oo, Rh 2E-3 5E+2 W, see 8E+Z Y, see 9 9-Rh 2E+2 45 RhodtumrlOZz, D, see 99mRh 1E+3 5E+2 LLI wall W, see 9 MRh (1E+3) -Rh 4E+2 45 Rhodlum-102 0, see 9"mRh 6E12 9E+1 45 ~ ~ Rhdu5-O 990 Rh 2E+2 Y, see ih 6E+1 45 Rhodiurl03m z 0, see 9 9moh 4E+5 1E+6 W, see 9 9 mRh 1E+6 Y, see 9 9 e~h 1E+6 45 Rhodlum-105 0, see 9 9 MRh 4E+3 1E+4 LLI wall_ .W se e 9 9M E- 3 ) 6 3 Y, see 9 9"Rh 6E+3 45 Rhadium-106m 0, see 9 9-Rh 8E-3 3E+4 90Rý 4E+4 U, see 9" Rh -4E+4 Y, see 9 9 MRh 4E+4 45 Rhodium-107 2 D, see 9 9M Rh 7E+4 2E+5 St. wall 9 gM 9 h (9E+4) -W see 9R h 3E+5 Y, see 9-'- 3E+5 46 Palladlam-100 0D all compounds except those given for W and Y 11E3 11+3 W, nitrates 1E+3 Y, oxides and hydroxides

-11+3 46 Pal ladlum-101 0 see lWPd 11+4 3E+4 WU see 1 0 Pd 3E14 Y see 1 0 0 Pd 3E+4 46 Palladium-103

0. see 1 W pd 6E13 6E+3 LLI wall 1 0 0 p (7E+3) -W see Pd 4E+3 Y, see 1 0Pd 4E+3 46 Palladium-107 D, see 1 0OPd 3E14 2E+4 LLI wall Kidneys (4E+4) (2E+4)W see 7 0 0-E13 Ysee Pd 4E+2.1 Table Z Table 3 1 Values Effluent Releases to Concentrations Sewers Col. 3 Col. 1 Col. 2 Monthly halation Average DAC Air *Water Concentration (pCi/ml) (pei/ml) (pCi/ml) (pCi/ml)2E-6 7E-9 2E-5 ZE-4 2E-6 6E-9 2E-6 5E-9 5E-6 2E-8 8E-5 8E-4 4E-6 1E-8 3E-6 1E-B 2E-7 7t-10 3E-5 3E-4 3E-7 1E-9 -6E-8 2E-10 -2E-7 7E-10 --2E-5 2E-4 2E-7 5E-10 5E-8 2E-10 4E-8 1E-10 8E-6 BE-5 7E-8 2E-10 --2E-B 8E-11 5E-4 2E-6 6E-3 6E-2 5E-4 2E-6 -5E-4 2E-6 -5E-6 2E-8 ---5E-5 51-4 3E-6 9E-9 --2E-6 8E-9'1E-5 4E-8. 11-4 1E-3 2E-5 5E-8 --11E5 5E-8 .1E-4 3E-7 ---1E-3 1E-2 1E-4 4E-7 --1E-4 3E-7 --6E-7 2E-9 2E-5 2E-4 5E-7 2E-9 -6E-7 2E-9 5 5E-8 2E-4 2E-3 11-5 5E-8 1E-5 4E-8 3E-6 9E-9-1E-4 1E-3 2E-6 6E-9 -1E-6 5E-9 -9E-6-3E-8 5E-4 5E-3 3E-6 1E-8 -2E-7 6E-10 -B-15 ODCM Rev. 25 (C)co U, App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION.

App. B Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. 1 Col. 2 Oral Monthly, Ingestion Inhalation Average Atomic Radionuclide Class ALL ALl 0AC Air Water Concentration No. (pCi) (PCi) (pCi/ml) (pCi/ml) (pCi/mi) (pCi/ml)46 Palladium-l09 D see 1 0 0 Pd 2E-3 61+3 3E-6 9E-9 3E-5 3E-4 46 alldiu-10 WD se W , see 1 0 0 d 5E+3 2E-6 8E-9 -Y, see d- 5E3 2E-6 6E-9 47 Silver-102 2 0, all compounds except those given for W and Y SE+4 2E+5 BE-5 2E-7 -St. wall (6E-4) 9E-4 9E-3 W, nitrates and sulfides ZE5 9E*S 3E-7 -Y, oxides and hydroxides'.

2E+5 8E-5 3E-7 -47 Silver-103 2 D, see 102A 4E-4 1E+5 4E-5 1-7 5E4 5E-3: S 1 0 2 g 41+5 4E-5 1E-7 --w , s e 1 0 2 A l I E -S -z 7 Y. see Ag A 1E+S 5E-5 2E-7 47 Silver-104m 2 D, see 102g 3E-4 9E+4 4E-5 1E-7 4E-4 4E-3 i. see 1 0 2 A 1E+5 5E-5 2E-7 -V, see Ag 1E+5 5E-5 ZE-7 -47 Silver-104 2 D see 1 0 2 Ag 2E-4 7E+4 3E-5 1E-7 3E-4 3E-3 i. see}nAg 1E+5 6E-5 2E-7 -Y, see Ag -1E+5 6E-5 2E-7 -47 Siver~O5 O see102.47 Silver-OS 0 g see I~Ag 3E+3 1E13 4E-7 1E-9 4E-5 4E-4 V, see 029 2E1EE- 9 2E+ 3E-9 -Y, see 1 Ag -21+3 71-7 2E-9 --47 Silver-106m

0. see 102g 8E+2 7E-2 3E-7 1E-9 1E-5 1E-4 W, see 1 9E+2 4E-7 1E-9 -Y ee 42 0g 9E+2 4E-7 1E-9 -47 Silver-106 2 0, see 1 0 2 Ag 6E+4 2E+S BE-5 3E17 -St. wall n102g (6E+4) --9E-4 9E-3 102 see 2E+5 9E-5 3E-7 -, see "Ag ""2E+S BE-S 31-7 47 Silver-1O0m 0, see 102g 6E+2 2E12 8E-8 3E-10 9E-6 9E-5 W, see 1 0 2 A -3E÷2 1E-7 4E-10 -W. see 102A 2E÷1 1E-B 3E-11 -47 Silver-Ib0m D, see 102Ag 5E+2 1E+2 5E-8 2E-10 6E-6 6E-5 W 1 0 2 A9 2E+2 8E-8 3E-10 -Y., see Ag 9E+1 4E-B 11-10 --.47 Silver-Ill D. see 102Ag 9E12 2E-3 6E-7 " LLI wall Liver 102 (11E3) (2E13) -2E-9 2E-5 2E-4.Y, see Ag 9E12 4E-7 1E-9 -47 Silver-112 0 Ag 31+3 81+3 3E-6 11- 4E-5 4E-4 Ssee 1 0Ag 3 11.4 41-6 11-8 W, E+4 4E-6 IE-8 -Y. see, Ag 9E-3 4E-6 1E-8 B-16 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION Table 1 Table 2 Table 3 Occupational Values Effluent Reeases to Concentrations Sewers Col. I Col. 2 Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average Atomic Radionuclide Class ALI ALI OAC Air Water Concentration No. (pCi) (pCi) (pCl/ml) (pCi/ml) (pCi/ml) (pCi/ml)App. B 1-(0 co ul In 47 Slver-115 2 0, see 102Ag 3E-4 9E+4 St. wall W S 102 (3E-4) -y , s ee 102 Ag 9E-4 Y, see Ag 8E-4 48 Cadmium-104 2 0, all compounds except those given for W and Y 2E+4 7E+4 W, sulfides, halides, and nitrates lE+5 Y, oxides and hydroxides -E+55 48 Cadmium-107 0, see 1 0 4 Cd 2E+4 5E-4 W s 1 0 4 Cd 61E+4 Y* see SEC+- 14 48 Cadmium-109 0, see 1 0 4 Cd 3E+2 4E+1 Kidneys Kidneys (4K.Z) (SK.'1)W, see 104Cd ( 2IE2 Kidneys-(IE+2)Y, see 1 04Cd -1E+2 48 Cadmium-113m D. see 104Cd 2E+1 2E+O Kidneys Kidneys (4E+1) (4E+O)W. see 1 0 4 Cd -8E+0 Kidneys-(11.1)Y, see 1 0 4 Cd -1E*1 48 Cadmium-113 D. see 1 0 4 Cd 2E+l. 2E10 Kidneys Kidneys (3E.1) (3EKO)W, see 1 04Cd 8E.8 Kidneys-(1E+1)Y, see 1 0 4 Cd 1E+I 48 Cadmium-115e U, see 104Cd 3E+2 SE+1 Kidneys-. (8E+1)W, see 1 0 4 Cd 1E+2 Y, see 104Cd 1KE2 48 Cadmium-115 0, see 104Cd 9E+2 1E+3 LLI wall (11E3)Y, see 04Cd E+3 Y, e: 1 0 4 Cd 48 Cadmiuw-l17m D, see 204Cd 5E+3 1E+4 W, see 0 4 Cd 2E14 Y e104Cd -1E+4 4E-5 IE-7 ---4E-4 4E-3 4E-5 1E-7 3E-5 1E-7 3E-5 9E-8 3E-4 3E-3 5E-5 ZE-7 -5E-5 2E-7 -*2E-5 BE-8 3E-4 3E-3 2E-S 8E-8 --2E-5 7E-8. --1E-B " 7E-11 6E-6 6E-5 5E-8-2E-10 5E-8 2E-il 1E-9-SE-12. 5E-7 5E-6 4E--9 4K-S ---5E-9 2E-11 --9E-10-5E-12 4E-7 4E-6 3E-9-2E-11 6E-9 2E-11 ZE-8 -4E-6 4E-S-1K-Il -5E-8 2E-10 -6E-S 2E-10 -6E-7 2E-9 ---1E-5 IE-4 5E-7 2E-9 6E-7 2E-9 -5E-6 2E-8 6E-5 6E-4 7E-6 2E-8 6E-6 2KE-8 B-17 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B.Atomic Radionuclide Class No.E 8 104 48 Cadmium-117 0, see 1 W: se 1 0 4Cd Y ...1 0 4 Y, seeCd 49 Indium-r09 0, all compounds except those given for W W, oxides, hydroxides, halides, and nitrates 49 Indium-110 2 , see109 (69.1 min) W, see In 49 Indium-110 0, see 109. n (449 h) W, see In 49 Indium-ill 0 see 109In W, see 19In 109.49 Indium-112 2 a see 109 W: see In 49 Indium-l13m2 0, see 109.W:se: 1 0 9 ln 49 lndium-ll4m

0. see. 1091In W, see 109 In 49 Indium-115 0 see 9 1n04 W: see In 49 Indiu-11S5 0, see 1 0 9 1n W, see In 49 Indium-116m 2 D see 1091n W, See 1091n 49 Indium-117m 2 0 see 109 n W: see 109 In W 109 I 49 Indium-119 2 0, see 1 0 9 In 4 W, see 1 0 9 In 50 Tin-110 D, all compounds except those given for W 4 W, sulfides, oxides, hydroxides, halides, nitrates, and stannic phosphate 50 Tin-111 2 D see 110nn W: see Sn Table I Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average ALI -XAL FAC Air Water Concentration (pCi) (pCi) (pCi/ml) (pCi/ml) (pCl/ml) (pCi/ml)5E-3 1E-4 5E-6 2E-8 6E-5 6E-4 2E+4 7E-6 2E-8 --11E4 6E-6 2E-8 -.2E÷4 4E+4 2E-5 6E-8 3E-4 3E-3 6E+4 3E-5 9E-8 2E+4 4E+4 2E-S 6E-8 2E-4 .2E-3 6E+4 2E-5 8E-8 --5E-3 2E+4 7E-6 2E-8 7E-5 7E-4 ZE-4 8E-6 3E-8 --4E+3 6E+3 3E-6 9E-9 6E-5 6E-4 6E+3 3E-6 9E-9 --2E+5 6E+5 3E-4 9E-7 2E-3 2E-2 7E+5 3E-4 11-5 -5E+4 1E+5 6E-5 2E-7 7E-4 7E-3 2E+5 8E-5 3E-7 -3E+2 6E+1 3E-8 9E-11 -LL! wall 4E.2) ---5E-6 5E-5 11+2 4E-8 1E-10 -1E+4 4E+4 2E-5 6E-8 2E-4 2E-3 5E+4 2E-5 7E-S --1E1 1E+0 6E-10 2E-12 5E-7 5E-6*E10 2E-9 8E-12 --E+4 8E-4 3E-5 1E-7 3E-4 3E-3 IE15 51-5 2E-7 1E+4 3E.4 1E-5 5E-8 2E-4 2E-3 4E+4 2E-5 6E- --E+4 2E15 7E-5 2E-7 8E-4 8E-3 2E+5 9E-5 3E-7 -E+4 1E+5 5E-5 2E-7 --t. wall E+4) ---7A-4 7E-3 E115 6E-5 2E-7 --E+3 7E+4 5E-5 2E-8 5E-5 5E-4 5E-6 2E-8 9E-5 3E-7 1E-3 1E-2 1E-4 4E-7 --11+4 2E35 3E+5 B-18 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Atomic Radlonuclide No.Class Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average ALl ALl 0AC Air Water Concentration (pCi) (pCi) (pCi/ml) (pCi/ml) (pCi/Ml) (pCi/el)2E+3 1E+3 SE-7 2E-9 -LLI wall (2E+3) --.3E-5 3E-4 5 E-2 21-7 BE-10 --2E-3. 1E+3 SE-7 LLI wall Bone surf (21E3) (2E+3) 3E-9 3E-- 3E-4 10+3 6E-7 2E-9 --N, LO QD tN cc F-CK LO 50 Tin-l3 0, see l10sn W, see 110Sn 50 Tin-117m 0, see 110Sn W, see 110Sn 50 Tin-ligm 0, see ll0sn W, see 110Sn 50 Tin-121m 0, see 110Sn W, see 1105n 50 Tin-121 0, see 110sn W, see 110Sn 50 Tln-123a 2 0 see s Wt s00 1105.so Tin-123 0. see 11n W, see 1105 50 Tin-123 0, see 11sn Wt see 1 1 0 Sn 50 Tin-126 0 see 1103n SO Ti-127, see 11050 W: 11050 50 Tin-12 0, see 11o0n W, see 103Sn 50 Tin-1272 1)se 110sn':110s 51 Antimony-115 2 0, all compounds except those given for W W, oxides, hydroxides, halides, sulfides, sulfates., and nitrates 51 Antimony-116m 2 0 s 115 W 1155 51 Antimany-1162 0, see l 1 S3b W, see 115Sb 51" Antimony-l17 0: see 115S W see 3E-3 2E+3 LLI wall (4E+3) --1+3 3E+3 9E+2 LLI wall (4E+3) -5E-2 6E+3 2E+4 LLI wall (6E+3) -1E-4 5E+4 1E+5 1E+5 5E+2 6E+2 LLI wall (61E2)2E÷2.41-2 9E+2 LLI wall (5E+2) -3E+2 6E+1 7E+1 7E13 2E÷4 2E+4 9E+3 3E+4 4 E+4 8E+4 2E-5 3E+5 2E+4 7E+4-11'S 7E+4 3E#5 St. wall (9E14) -3E+5 7E+4 .2E15 3E+5 1E-6 3E-9 5 61-4 41-7 1E-9 4E-7 1E-9--5E-5 5E-4 2E-7 8E-10 61-6 2E-8 --BE-5 8E-4 5E-6 2E-8 --5E-5 2E-7 71-4 , 7E73 6E-5 21-7 31-7 9E-10 ---9E-6 9g-m 7E-8 21-10 --4E-7 11-9 ----6E-6 6E-5 11-7 5E-10 -" 2E-8 8E-11 4E-6 41-5 3E1a 91-11 --8E-6 31-8 9E-5 9E-4 80-6 3E-8 --1E-5 4E-8 1E-4 1E-3 1E-5 5E-8 1E-4 3E-7 IE-3 1E-2 1E-4 4E-7 3E-5 1E-7 3E-4 3E-3 6E-5 2E-7 4 41-7 ---1E-3 1E-2 1E-4 5E-7 --9E-5 3E-7 9E-4 9E-3 1E-4 4E-7 --B-19 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Table I Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. 1 Col. 2 Oral .Monthly Ingestion Inhalation

_Average ALl All F AC Air Water Concentration (pCi) (pCi) (pCi/ml) (pCi/al) (pCi/ml) (pCi/ml)Atomie Radionuclide Class No.C L L 8 51 Antimony-118i

.0 see W, see 1 1 5 Sb 51 Antimony-119 D see115 Ws 1 1 Sb 51 Antimony-120 2 0, see l 1 5 Sb (16 min)W, see 1155b 51 Antimony-120 D, see 1 1 5 Sb (5.76 d) W, see Sb 51 Antimony-122 0, see ll5sb W, see ll 5 Sb 51 Antimony-124m 2 0 see 1 1 5 Sb W, see llsb 51 Antimony-124 D; see 115 0 W, see Sb 51 Antiaony-125 0 s ,ee15 51 Antimony-126 2 D, see 1 o W, see 1155b 51 Antimony-126 0 see 1155b 51 Antimony-127 0, see 115Sb W. see Sb 51 Antimony-128 2 0, see 1 1 5 Sb (10.4 mis)W, see 1 1 5 Sb 51 Antimony-128 0 see 115Sb (9.01 h) leS%51 Antimony-129 D, se 115 (9 01 h) W, 11e Sb 51 Antimony-lb0 2 0 see 11 5 5b s 115 Sb 51 Antimony-131 2 D, see 115S W, see 1 1 5 Sb 6E13 2E-4 5E13 2E-4 2E+4 SE+4 2E+4 3E14 1E5 4E+5 St. wall (2E15) S 5E+5 1E+3 2E+3 9E+2 1E+3 80-2 21.3 LLI wall (8E+2) -7E+2 1E+3 3E+5 8E+5 2E+5 6E+5 6E+2 9E+2 5E+Z 2E+2 2E-3 2E+3 5E+2 51E4 2E+5 St. wall (7E-4)-.2E-5 6E+2 1E+3 5E.2 5E+2 8E+2 2E+3 LLI wall (8E+2) -7E+2 9E+2 8E+4 4E+5 St. wall (1E+5)4E15 1E-3 4E+13 3E+3 3E-3 9E+3 9E+3 2E.4 6E+4 8E44 1E+4 2E+4 Thyroid Thyroid (2E+4) (4E+4)2E+4 Thyroid (4E-4)8E-6 3E-8 7E-5 7E-4 9E-6 .3E-8 2E-S 6E-8 2E-4 2E-3 1E-5 4E-8 2E-4 6E-7--2E-3 2E-2 2E-4 71-7 --9E-7 3E-9 1E-5 1E-4 5E-7. 2E-9 1E-6 3E-9 ---1E-5 1E-4 4E-7 2E-9 --4E-4 1E-6 3E-3 3E-2 2E-4 8E-7 --4E-7 1E-9 .7E-6 7E-5 1E-7 3E-10 --1E-6 3E-9 3E-5 3E-4 2E-7 7E-10 8E-5 3E-7 ---9E-4 9E-3 81-5 3E-7 --5E-7 2E-9 7E-6 7E-5 2E-7 7E-10 9E-7 3E-9 ----1E-5 1E-4 4E-7 1E-9 2E-4 5E-7--1E-3 1E-2 2E-4 6E-7 -2E-6 6U-9 2E-5 2E-4 1E-6 5E-9 --4E-6 1E-8 4E-5 4E-4 4E-6 1E-8 --3E-5 9E-8 3E-4 3E-3 3E-5 1E-7 -1E-5-6E-8 2E-4 2E-3 1E-S-6E-8 B-20 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. 1 Col. 2 Oral Monthly ingestion Inhalation Average ALI ALl DAC Air Water Concentration (pCi) (pCi) (pCi/ml) (pCi/ml) "(pCi/ml) (pCi/mi)L D L g Atomic Radionuclide Class No.52 Tellurium-1iS D, all compounds except those given for W W. oxides, hydroxides.

and nitrates 52 Telluriaum-121m D, see 116Te-W, see 116Te 52 Tellurium-121 D, see 116 W, see 52 Tellurium-123s

0. see 1 1 5 Te W, see 116Te 52 Teiiuriia-123 D, see a16Te W, see "ýTe 52 Tel~luriua-125m D, see 11fTo W, see 116To 52 Tellurla-127i D. see 116Te W, see 1 1 6 Te 52 Tellurium-r27 0, see 116 T W, see 1 1 6 Te 52 Tel luriu-129m , see 6Te W, see 52 Tellurlu-U-192 0, see 11 W, see 52 Tellurium-131n 0, see Ul6Te W, see 116 Te 52 Telluriur131 2 0, see 116Te W, see ll6Te BE-3 2E+4 9E-6 30-8 1E-4 1E-3 3E+4 1E-5 4E-8 -SE-2 2E+2 BE-B --Bone surf Bone surf (7E12) (4E+2) 5E-10 1E-5 1E-4 4E+2 2E-7 6E-lB 3E+3 4E+3 2E-6 6E-9 4E-5 4E-4 3E+3 1E-6 4E-9 6E72 2E+2 9E-8 -Bone surf Bone surf (1Ee3) (5E+2), -8E-10 1E-5 1E-4-5E+2 2E-7 8E-10 5E+2 2E+2 BE-B -Bone surf Bone surf (1E+3) (5E+2) -7E-10 2E-5 2E-4 4E02 2E-7. -Bone surf (10E3) 2E-9 1E+3 4E+2 2E-7 --*Bone surf Bone surf (1E+3) (1E+3) -1E-9 2E-5 2E-4 7E42 3E-7 1E-9 --6Ee2 3E+2 1E-7 9E-6 9E-5 Bone surf-(4E'2) -GE-10 -3EE2 lE-7 4E-10 -7E03 2E+4 9E-6 3E-8 1E-4 1E-3 2E04 7E-6 2E-8 -5E+2 6E+2 3E-7 9E-10 7E-6 7E-5 2E0-2 1E-7 3E-10 3E+4 6E+4 3E-5 9E-8 4E-4 4E-3-7E+4 3E-5 lE-7 3E+2 .4E02 2E-7 -" Thyroid Thyroid .I (6E+2) (1E+3) -.2E-9 BE-6 BE-5 4E-2 2E-7 -Thyroid (9E02) -.1-E-9 3E+3 5E+3 2E-6 -Thyroid Thyroad (6E+3) (1E+4) -2E-8 8E-5 6E-4 SE+3 20-S -Thyroid (1E+4) -2E-B -B-21 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Table 1 Table2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. I Col. 2 Col. 3 Col.' I Col. 2 Oral Monthly Ingestion Inhalation Average AL! 7L D Air Water Concentration (pCi) (pCi) (pci/ml) (pci/ml), (pci/ni) (pci/ml)Atomic Radionuclide Class NO.52 Tellurium-132 0, see 116Te W, see 116Te 52 Tellrim-133m 2 D. see 116Te W, see 116Te 52 Tellurium-133 2 D, see 116Te W, see i16Teo 52 Tellurlum-134 2 0, see 116Te W, see n6To 53 l0dine-120mW D, all compounds 53 Iodine-120 2 D, all compounds 53 lodine-121 0, all compounds 53 Iodine-123 D, all compounds 53 lodlne-124 0, all compounds 53 Iodine-125 0, all compounds 53 Iodine-126 0, all compounds 53 iodine-128 2 0, all compounds 53 Iodine-129 D, all compounds ZE+2 2E+Z Thyroid Thyroid (7VE2) (8E+2)2E-2 Thyroid-(6E-2)3E+3 SE+3 Thyroid Thyroid (6E+3) (1E+4)5E-3.Thyroid (1E+4)1E+4 2E+4 Thyroid Thyroid (3E+4) (6E÷4)2E+4 Thyroid (6E+4)2E+4 2E+4 Thyroid Thyroid (ZE4) (5E+4)2E+4 Thyroid-(SE+4).1E÷4 2E+4 Thyroid (ME-4) -4E+3 9E-3 Thyroid Thyroid (8E+3) (1E+4)1E+4 2E+4 Thyroid Thyroid (3EI4) (5E+4)3E+3 6E+3 Thyroid Thyroid (1E+4) (2E+4)5E*1 8E+1 Thyroid Thyroid (2E+2) (3E+2)4E-1 6E41 Thyroid Thyroid (IE÷2) (2E+2).2E÷1 4E÷1 Thyroid Thyroid (7E+1) (1E+2)4Et4 1E+S St. wall (6E ÷4) -SE+O 9E+O Thyroid Thyroid (2E-1) (3E+1)9E-8 9E-8 2E-6 2E-6 9E-6 9E-6 1E-5 1E-5 9E-6 4E-6 SE-6 3E-6 3E-8 3E-8 5E-5 1E-9 9E-10 2E-8 2E-8 8E-8 7E-8 7E-8 3E-8j 2E-8 7E-8 2E-B 4E-10 3E-81 2E-1O 2E-7 9E-6 9E-5 9E-5 4E-4 3E-4 2E-4 IE-4 4f-4 1E-4 2E-6 2E-6 IE-6 9E-4 41f-3 3E-3 2E-3 111-3 4E-3 IE-3 IE-5--8E-4 8E-3 4E-9 ---4t-i 2E-7 2E-6 B-22 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. 1 Col. 2 Oral- monthly Ingestion Inhalation Average.Atomic Radionuclide Class ALI ALI DAC Air Water Concentration No. (pCi) (pCi) (pCi/ml) (pCi/m)) (pCi/mI) (pCi/ml)53 Iodine-130 D, all compounds 4E+2 7E÷2 3E-7 --" 53 Iodine-131 0, all compounds 53 Iodine-132m 2 D, all compounds 53 Iodine-132 D, all compounds 53 Iodine-133 0, all compounds Thyroid Thyroid (1E+3) (2E-3)3E+1 5E+1" Thyroid Thyroid (9E11) (2E+2)4E+3 8E+3 Thyroid Thyroid (1E+4) (2E+4)4E+3 BE+3 Thyroid Thyroid (9E÷3). {E+4)1E+2 3E+2 Thyroid Thyroid (5E+2) (9E+2)2E+4 5E+4 Thyroid (3E14) -8E+2 2E.3 Thyroid Thyroid (3E+3) (4E+3)53 Iodine-134 2 3E-9 2E-8 -2E-10 4E-6 0, all compounds 53 lodlne-35 0, all compounds 54 Xenon-120 2 Submersion 1 54 Xenon-121 2___. Submersion1 54 Xenon-122 Submersion1 54 Xenon-123 Submersion1 54 .Xenon-125 Submersion' 54 Xenon-127 Submersion 1 54 Xenon-129m Submersion' 54 Xenon-131m SubmersionI 54 Xenon-133m Submersion 1 54 Xenon-133 Submersion1 54 Xenon-135m2 Submersion1 54 Xenon-135 Submersion1 54 Xenon-1382 Submersion1 55 Cesium-125 2 D, all compounds 55 Cesiom-127 D, all compounds 3E-6 1E-7 2E-5 7E- 7 1E-5 ZE-6 7E-5 6E-6 ZE-5 1E-5 2E-4 4E-4 1E-4 1E-4 3E-8 2E-8 1E-9 6E-8 6E-9 4E-8 1E-8 3E-7 3E-8 7E-8 6E-8 9E-7 2E-6 6E-7 5E-7 4E-8 7E-8 2E-8 2E-7 1E-7 2E-5 2E-4 1E-6 1E-5 1E-4 1E-3 1E-4 1E-3 7E-6 7E-5 4E-4 4E-3 3E-5 3E-4 S -9E-6 S -1E-5 S -4E-6 5E14 1E+5 6E-5 St. wall (9E+4)6E8 4 9E+4 4E-5 1E-3 9E-4 1E-2 9E-3 B-23 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION.

App. B Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average'ALl ALl DAC Air Water Concentration (pCi) (pCi) (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml)cc L-00 In)Atomic Radionuclide Class No.55 Cesium-129 0, all compounds 55 Cesium-130 2 0, all compounds 55 Cesium-131 0, all compounds 55 Cesium-132 0, all compounds 55 Cesium-134e 0, all compounds 55 Cesium,-134 0, all compounds 55 Cesium-135m 2 0, all compounds 55 Cesium-135 0, all compounds 55 Cesium-136 D, all compounds 55 Cesium-137 0, all compounds 55 Cesium-138 2 0, all. compounds 56 Barium-126 2 0, all compounds 56 Barium-128 0, a 1"compounds 56 Barium-131m 2 0. all compounds.56 Baritm-131 D, all compounds 56 Barium-133m D, all compounds 56 Barium-133 D, all compounds 56 Barium-135m 0, all compounds 56 Barium-13U 2 D, all compounds 56 Barium-O140 , all compounds 56 Bariu-1412 D, all compounds 56 Barium-142 2 0, all compounds 57 Lanthanum-131 2 0, all compounds except those given for W W. oxides and hydroxides 2E-4 3E+4 6E-4 2E-5 St. wall (1E5) -2E+4 3E+4 3E13 4E-3 1E-5 1E15 St. wall (1E+5) -7E+1 1E-2 1E+5 2E+S 7E+2 1E+3 4E+2 7E-2 IE12 2E12 ,2E+4 6E+4 St. wall (3E+4)6E+3 2E+4 5E-2 2E13 4E+5 1E+6 St. wall (5E+5)3E+3 8E÷3 2E13 9E+3 LLI wall (3E+3) -2E-3 7E+2 3E+3 1E4 1E+4 3E+4 5E+2 1E13 LLI wall (6E+2)2E+4 7Ev4 5E+4 11E5 1E-5 5E-8 3E-4 3E-3 BE-5 31-7 ----1E-3 1E-2 1E-S 4E-8 3E-4 3E-3 2E-6. 6E-9 .4E-5 4E-4 5E-S 2E-7 ---2E-3 2E-2 41-8 2E-10 9E-7 9E-6 8E-5 3E-7 1E-3 1E-2 5E-7 2E-9 1E-5 1E-4 3E-7 9E-10 6E-6 6E-5 6E-8 2E-10 1E-6 1E-5 2E-5 8E-8 4E-4 4E-3 6E-6 2E-8 8E-5 BE-4 7E-7 2E-9 + 7E-6 7E-5 6R4 2E-6 -S -7E-3 7E-2 3E-6 1E-8 4E-S 4E-4 4E-6 1E-8 4E-S 4E-4 3E-7 9E-10 2E-5 2E-4 51E-6 2E-B 4E-S 4E-4 1H-5 4E-8 2E-4 2E-3 64-7 2E-9 "-8E-6 8E-5 3A-S 1E-7 3E-4 3E-3 6E-5 2E-7 7E-4 7E-3 5E+4 11E5 SE-S 2E-7 6E-4 6E-3 2E-5 7F-5 2E-7 --B-24 ODCM Rev. 25 App. B App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Coi. 1 Col. 2 Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average Atomic Radionuclide Class AI *ALI OAC Air Water Concentration No. (pCI) (ICN ) , (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml)p.I.S I.57 Lanthanum-132 D3 see 1 3 1 La W see 1 3 1 La 57 Lanthanum-135 0: see 131L W see 1 3 1 La 57 Lanthanum-137.

0, see 1 3 1 La W. see 131La 57 Lanthanum-138 0 see 131La' : 131 La 57 Lanthanum-140 1, see 1 3 1 La W: see 1 3 1 La 57 Lanthanue-141 0 1 3 1 a O, see 3L W: 131L: 57 Lanthanum-142 2 0 see1La W, se11La 57 Lanthanum-143 2 D. see 131La W. see 1 3 1 La 58 Cerium-134 W, all compounds except.those given for Y Y, oxides, hydroxides, and fluorides 58 Ceriue-135 W: see 134 Ce C134 58 Cerium-137m W. see 1 4Ce Y. see 13'4Ce 58 Cerium-137 W see 134 Y. ee 1 3 4 Ce 58 Ceriue-139 W,.see '3'.Ce see Ce Se Ceriumr-139 W, see 134C Y, see 134Ce 58 Cerium-141

., see 134Ce Y, see 1 3 4 Ce Se Certuar-143 W. see 134 Ce Y. see 134Ce 4E.4 11E5 11E4 6E+1 Liver (7E+1)3E+2 Liver-(3E+2)9E+2 4E1O-11.1 6E+2 1E+3 1E-3 4E+3! 9E-3 1E+4 8E+3 2E14 3E*4 4E-4 1E.5 St. wall (4E-4) -9E+4 5E+2 7E+2 LLI wall (6E-2) .7E+2 2E13 4E.3 4E+3 2Ev3 4E13 ELI wall (2E+3) --4E+3 SE+4 1E+S 1E+5 SE+3 8E+2-7E+2 2E+3 7E+2 LLI wall (2E+3) -6E+2 1E+3 2E+3 LLI wall (1E+3) -2E+3 4E-5 1E-7 5E-4 5E-3 4E-5 1E-7 --3E-8 -2E-4 2E-3-1E-10 11-7-4E-10 1E-9 .SE-12 1E-5 1E-4 6E-9 2E-11 --6E-7 2E-9 9E-6 9E-5 5E-7 2E-9 4E-6 1-8 5E-5 5E-4 5E-6 2E-8 --9E-6 3E-8 1E-4 1E-3 1E-S 5E-8 -.4E-5 1E-7 -.-E 5E-4 5E-3 41-S 11-7 --3E-7 1E-9 ---. 81-6 8E-5 3E-7 9E-10 -2E-6 SE-9 2E-5 2E-4 1E-6 SE-9 -2E-6 6E-9 ---3E-5 3E-4 2E-6 SE-9 6E-5 2E-7 7E-4 7E-3 51E5 2E17 3E-7 1E-9 7E-5 7E-4 3E-7 9E-10 -3E-7 1E-9 --3E-5 3E-4 2E-7 8E-10 -8E-7 3E-B --2E-5 2E-4 7E-7 2E-9 --3E+3 1E14 4E-6 1E-8 4E-5 4E-4 1E+4 5E-6 2E-8 --B-25 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average ALl .-Ac OR- Air Water Concentration (NCi) (pCi) (pCi/ml) (PCilml) (pCi/ml) (pCi/ml)Atomic Radionuclide Class No.00 5, U, 58 Cerium-144 W, see 134Ce Y, see 1 3 4 Ce 59 Praseodymium-136 2 W, all compounds except those given for Y Y, oxides, hydroxides, carbides, and fluorides 59 Praseodymium-137 2 W see 1 3 6 Pr.Y .... 136 Pr 59 Praseodymium-138m W see 1 3 6 Pr" ,see 1 3 6 Pr* 1 3 6.59 Praseodymium-139 W, see 1 3 6 Pr Y e see 3 6 r S9 Praseodymiium-142m2W see 1 3 6 Pr Ysee 1 3 6 Pr 59 Praseodyinius-142 W, see 1 3 6 r Y, see 16Pr 59 Praseodymium-143 W, see 1 3 6 Pr Y, see 1 3 6 Pr 59 Praseodymium-1442 W, see i16 Pr Y, see 136 Pr 59 Praseodymium-145 W, see 1 3 6 Pr Y s 1 3 6 Pr 59 Praseodymium-147 2 W, see 1 3 6 Pr Y, see 136Pr 60 Neodymium-136 2 W, all compounds except those given for Y Y. oxides, hydroxides, carbides, and fluorides 60 Neodymium-138 , see 1 3 6-':136N"d Y, see d 60 Neodymium-139M W, see 1 3 6 d Y, s 136Nd 2E-2 3E61 LLI wall (3E-2) -SE.4 2E+5 St. wall (7E+4) -2E+5 4E+4 2E+5 1E+5 1E-4 5E+4 4E+4 4E+4 1E+5 BE÷4 2E+5-1E+5 1E-3 2E+3-2E+3 9E+2 8E-2 LLI wall (IE+3) -7E+2 3E+4 1Ev5 St. wall (4E.4) -1E.5 3E+3 9E.3 8E+3 5E+4 2E+5 St. wall (86E4)2Ev5 1E64 6E-4 5E64 2E+3 6E÷3-, 5E÷3 1E-8 4E-11 ---3E-6 3E-5 6E-9 2E-11 1E-4 3E-7" H1-3 1E-2 9E-5 3E-7 ~6E-S 2E-7 5E-4 5E-3 6E-5 2E-7 --2E-5 8E-8 1E-4 1E-3 2E-5 6E-8 --5E-5 2E-7 6E-4 6E-3 5E-S .2E-7 --7E-5 2E-7 1E-3 1E-2 6E-5 2E-7 -9E-7 3E-9 , 1E-5 1E-4 8E-7 3E-9 --3E-7 1E-9 ---2E-5 2E-4 3E-7 9E-10 -5E-5 2E-7 ---6E-4 6E-3 5E-5 2E-7 --46-6 1E-8 04-5 4E-4 3E-6 1E-8 --8E-5 3E-7 ---1E-3 1E-2 8E-5 3E-7 2E-5 8E-8 2E-4 2E-3 2E-5 BE-8 -3E-6 9E-9 3E-5 3E-4 2E-6 7E-9 --5E+3 2E+4 7E-6 2E-8 7E-5 7E-4-1E+4 6E-6 2E-8 --B-26 ODCM Rev. 25 App. B PART 20 -STANDARDS FOR PROTECTION AGAINST RADIATION App. B to to I-Atomic Radionuclide Class No.2 136 .60 Neodymium-1392 W, see 1 3 6 n Y, see 136Nd 60 NeodyniLm-141 W, see 1 3 6 Bd Y, see 136Md 60 Neodymium-147 w, see 136Nd Y. see 1 3 6 Nd" 60 Neodymium-149 2 W, see 136 Y, see 2 136..60 Neodymium-1512 W, see 3 6 N Y , see 136Nd 61 Promethium-141 2.W, all compounds except those given for Y Y, oxides, hydroxides, carbides, and fluorides 61 Promethiur-143 W, see141 Y S:141pm Y, see PM 61 Promethiu.-144 W, see141* ~Y, see 61 Pronethium-145 U, see11p Y, se 141pm see 141P 61 Promethium-145 W, see 1 4 1 Pm Y, see 141 PM 61 Promethfur-147 W, see 1 4 1 p.Y, see 141pM 61 Promethium-14

W, s-e 141e Y. see 61 Promethium-148 W, see 141pm Yse 141pe 61 Promethiuor149 W, see 141pm Y, see 141PIN 61 Proethium-150 W. see 141pm Y, see m 61 Promthfum-151 W, se: 141 pm Y, see 141p Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 .Col. 2 Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average ALI -ALI DA7 Air Water Concentration (pCi) (pCi) (pCi/ml) (pei/mi) (pCi/ml) (pCi/ml)9E-4 3E65 1E-4 5E-7 1E-3 1E-2 3E+5 11-4 4E67 --2E+5 7E.5 3E-4 IE-6 2E-3 2E-2 6E.5 3E-4 9E-7 -1E63 9E÷2 4E-7 1E-9 -LLI wall (1E+3) .. ..2E65 2E-4 BE. 2 4E6-7 1E-9 --1E+4 3E64 1E-5 4E-8 1E-4 1E-3 2E-4 1E-5 3E-8 7E.4 2E65 8E-S 3E-7 9E-4 9E-3-.E*5 8E-5 3E-7 --5E+4 2E.5 8E-5 3E-7 -St. wall (6E.4) -. 8E-4 BE-3-2E+S 7E-S 2E-7 5E63" 6E.2 2E-7 8E-10 7E-5 7E-4 7E62 3E-7 1E-9 1E+3 1E+62 S-B ZE-10 2E-5 2E-4 1E*2 5E-8 2E-10 -1E+4 2E.2 7E7-8 1E-4 1E-3 Bone surf (2E+2) -3E-10 -2E+2 8E-8 3E-10 -.2E+3 56E1 2E-8 7E-11 ZE-5 ZE-4 4E+1 2E-8 6E-11 4E-3 1E62 5E-8 .-LLI wall Bone surf (5E-3) (2E+2) 3E-10 7E-S 7E-4 1E+2 6E-8 2E-0o -7E+2 3E+2 1E-7 4E-10 1E-5 16-4 3E+2 1E-7 5E10 -4E+2 5E+2 2E-7 8E-10 -LLI wall (5E.2) -- 7E-6 7E-5 5E62 2E-7 7E-10 --1E+3 2E+3 8E-7 *. 3E-9. .-ILI wall (63) -- 2E-5 2E-4 3-. 2E+3 8E-7 2E-9 5E+3 2E44 8E-6 3E-8 7E-5 .7E-4 2E+4 7E-6 2E-8 2E+3 4E+3 1E-6 5E-9 2E-5 2E-4 3E+3 1E-6 4E-9 --B-27 ODCM Rev. 25 App. B 'PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App.B Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average Atomic Radionuclide Class ALI A tl 6 Air Water Concentration No. (pCi) (pCi) (pCi/ml) (pjCi/ml) (pCi/ml) (pCi/ml)62 Samarium-141m 2 W, all compounds 3E-4 1E.5 4E-5 i-7 4E-4 4E-3 62 Samarium-141 2 6 2 Samarium-142 2 62 Samarium-145 62 Samarium-146 W, all compounds W, all compounds W, all compounds W, all compounds r-n-u.62 Samariuum-147 W, all compounds 62 Samarium-151 W, all compounds 62 Samarium-153 W, all compounds 62 Samarium-155 2 W, all compounds 5E+4 2E-5 St. wall (6E+4)BE+3 3E+4 6E63 5E+2 1E+1 4E-2 Bone surf Bone surf (3E+1) (6E-2)2E61 4E-2 Bone surf Bone surf (3E+1) (7E-2)1E+4 1E-2 LLI wall Bone surf (1E+4) (2E-2)2E+3 3E+3 LLI wall (2E+3) -6E+4 2E+5 St. wall (8E+4) -56+3 9E63 2E+3 2E+3 1E63 1E-3 3E+3 2E+3 1E+3 4E+2 IE+4 3E+3 3E+3 8E+3 BE-5 2E-7 ---81-4 8E-3 1E-5 4E-8 1E-4 1E-3 2E-7 7E-10 8E-5 8E-4 1E-11 9E-14 3E-7 3E-6 2E-11 4E-B 1E-6 1E-13 4E-7 2E-10 ZE-4 4E-9 -4E-6 2E-3 3F-5 3E-4 9E-5 3E-7 62 Samarium-156 63 Europium-145 63 Europium-146 63 Europium-147 63 Europium-148 63 Europium-149 63 Europiuo-150 (12.62 h)63 Europium-150 (34.2y)63 Europium-152m, 63 Europium-152 63 Europium-154 63 Europlum-155 W, alt compounds W, all compounds W, alt compounds W, all compounds W, all compounds.W, all compounds W, all compounds W, all compounds W, all compounds W, all compounds W, all compounds W, all compounds 4E-6 8E-7 5E-7 7E-7 1E-7 1E-6 4E-6-.1E-3 1E-8 7E-5 3E-9 2E-5 2E-9 1E-5 2E-9 4E-5 5E-10 1E-5 4E-9 2E-4 1E-8 4E-5 1E-2 7E-4 2E-4 1E-4 4E-4 1E-4 2E-3 4E-4 8E+2 2E+1 86-9 3E-11 1E-5 1E-4 3E+3 6E-3 8E+2 2E61 5E6Z .2E61 4E÷3 9E+1 Bone surf (16E2)6E+2 5E62 3E-6 9E-9 4E-5 1E-B 3E-11 1E-5 BE-9 3E-11 7E-6 4E-B -5E-5 2E-10 -2E-7 6E-10 8E-6 4E-4 1E-4 7E-5 5E-4 8E-S 63 Europium-156 W, all compounds B-28 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Atomic Radionuclide Class No.63 Europium-157 W. all compounds 63 Europium-158 2 W, all compounds 64 Gadolinium-1452 D, all compounds except those given for W..W, oxides, hydroxides, and fluorides 64 Gadolinium-146

0. see 145 W, see Gd 64 Gadolinium-147-0, see 1 4 5 Gd W see Gd 64 Gadolinium-148 0, see 145Gd W, see 145Gd 64 Gadolinium-149 0, see 145 W, see 64 Gadolinium-IS1

0. see 145Gd" ., sex 1 4 5 Gd.64 Gadoliniumr-152 0, see 145Gd W, see 145Gd 64 Gadolinium-153 0, see 145Gd W, see 145Gd 64 Gadoliniua-159 D, see 1 4 5 Gd W, see 1 4 Gd 65 Terbium-147 2 W, all compounds 65 Terbium-149 W, all compounds 65 Terbium-150 W, all compounds 65 Terbnum-1S1 W, all compounds 65 Terbium-153 I, all compounds 65 Terbium-154 W, all compounds, 65 Terbium-155 W, all compounds 65 .Terbium-156m W, all, compounds (5.0 h)Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. I Col. 2 Oral Monthly Ingestion Inhalation Average ALl Atl W A Air Water Concentration (pCi) (pCi) (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml)2E53 SE+3 2E-6 7E-9 3E-5 3E-4 2E+4 6E+4 2E-5 RE-B 3E-4 3E-3 56E4 2E65 6E-5 E2-7 --St wall (5E+4) 6E-4 6E-3 2E+5 7E-5 2E-7 1E+3 1.E2 SE-8 2E-10 2E-5 2E-4 3E-2 IE-7 4E-1O -.2E+3 4E+3 2E-6 6E-9 3E-5 3E-4-4E+3 1E-6 5E-9 1E+1 8E-3 3E-12 Bone surf Bone surf (2E-1) (2E-2) -2E-14 3E-7 3E-6 3E-2 IE-11 -Bone surf (6E-2) 8E-14 3E+3 2E63 9E-7 3E-9 4E-5 4E-4 2E+3 1E-6 3E-9 --6E-3 4E+2 2E-7 -9E-5 9E-4-Bone surf (6E+2) 9E-10 -1E63 SET7 2E-9 2E61 1E-2 4E-12 Bone surf Bone surf (3E+1) (2E-2) -3E-14 4E-7 4E-6 4E-2 2E-11 Bone surf (BE-2) -1E-13 5E-3 1E2 6E-8 -6E-5 6E-4 Bone surf-(2E+2) -3E-10 6E+2 2E-7 8E-10 3E+3 8E*3 3E-6 16E- 4E-5 4E-4 6E63 2E-6 BE-9 -9E.3 3E.4 .1E-5 5E-8 1E-4 1E-3 5E+3 7E62 3E-7 1E-9 7E-5 7E-4, 56E3 2E-4 9E-6 3E-8 7E-S 7E-4 4E63 9E+3 4E-6 1E-B 5E-5 5E-4 5E-3 7E+3 3E-6 IE-B 7E-5 7E-4 2E+3 4E+3 2E-6 6E-9 2E-5 2E-4 6E+3 BE+3 3E-6 1E-B BE-5 BE-4 2E+4 3E-4 1E-5. 4E-B 2E-4 2E-3 B-29 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Table I Table 2 Table 3 OccupationalValues

-Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. I Col. 2 Oral Monthly Ingestion Inhalation Average ALt ALI DAC Air Water Concentration (pCi) (pCi) (pCi/ml) (pCi/mI) (pCi/mI) (pCi/ml)7E+3 8E-3 3E-6 1E-8 1E-4 "E-3 U, I-Atomic Radionuclide Class No.65 Terbium-156m W, all compounds (24.4 h)65 Terbium-i56 W, all compounds 65 Terbium-157 W, all compounds 65 Terbium-iS8 W, all compounds 65 Terbium-160 W, all compounds 65 Terbium-161 W. all compounds 66 Dysprosiur-155 W, alu compounds 66 Oysprosium-157 W, all compounds 66 Dysprosium-159 W. all compounds 66 Oysprosum.165.

W, all compounds 66 Dysprosiuom-166 W, all compounds 67 Holmium-155 2 W, all compounds 67 Holmiumr157 2 W, all compounds 67 Holmium-159 2 W, all compounds 67 Holmium-161 W, all compounds 67 Holmium-162m 2 W, all compounds 67 Holmium-162 2 W, all compounds 67 Holmiom-164m 2 W, all compounds 67 Holmium-164 2 W, all compounds 67 Holmiom-166m W, all compounds 67 Holmfus-166 W, all compounds 67 Holmlum-167 W, all compounds 68 Erbium-161 W, all compounds 68 Erbium-165 W, all compounds 1E+3 1E+3 5E+4 3E+2 LL! wall Bone surf (5E+4) (6E-2)1E+3 2E+1 RE+2 2E62 2E÷3 2E+3 LLI wall (2E+3) -9E63 3E+4 2E+4 6E-4 1E64 2E'3 1E+4 5E+4 6iE2 7E62 LLI wall (8E-2)4E-4 2E*5 3E+5 1E+6 2E5' 1HI6 1E+5' 4E+5 SE+4 3E+5 5E+5 2E+6 St. wall (8E*5)1E+S 3E+5 2E+5 6E+S St. wall (2E+S) -6E-2 7E+O 9E-2 2E+3 LLI wall (9E+2) -2E+4 6E+4 2E-4 6E-4 6E+4 2E6S 6E-7 2E-9 1E-5 1E-4 1E-7 6E-10 7E-4 7E-3 BE-9 3E-11 2E-5 2E-4 9E-8 3E-10 1E-5 1E-4 7E-7 2E-9--3E-5 3E-4 1E-5 4E-8 1E-4 1E-3 3E-5 9E-8 3E-4 3E-3 1E-6 3E-9 2E-4 2E-3 ZE-5 6E-8 2E-4 2E-3 36-7 IE-9 ----1E-5 1E-4 6E-5 2E-7 6E-4 6E-3 6E-4 2E66 4E-3 4E-2 4E-4 IE-6 3E-3 3E-2 2E-4 6E-7 1E-3 1E-2-1E-4 4E-7 7E-4 7E-3 1E-3 3E-6 ----1E-2 1E-1 1E-4 4E-7 1E-3 1E-2 3E-4 9E-7 3E-3 3E-2 3E-9 9E-12 9E-6 9E-5 7E-7 2E-9 ---1E-5 IE-4 2E-5 BE-8 2E-4 2E-3 3E-5 9E-B 2E-4 2E-3 8E-5 3E-7 9E-4 9E-3 B-30 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Colt:l Col. 2 Col. 3 Col. I Col. 2 Oral Monthly lngestion Inhalation Average Atomic Radionuclide Class ALI ALI DAC Air Water Concentration No. (pCi) (pCi) (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml)App. B 68 Erbium-169 W, all compounds 3E-3 3E-3 LLI wall (4E.3)68 Erblum-171 W. all compounds 4E.3 16E4 68 Erbium-172 W, all compounds 1E+3 1E+3 LLI wall (1A-3) -69 Thulium-162 2 W, all compounds 7E.4 3E65 St. wall (7E64) -69 Thulium-166 W, all compounds 4E+3 1E+4 69 Thullum-167 W, all compounds 2E63 2E-3 LLI wall (2E+3) -69 Thulium-170 W, all compounds 8E+2 2E62 LLI wall'(1H-3) -69 Thulium-171 W, all compounds 1E-4 3E-2 LLI wall Bone surf (1E+4) (6E62)69 Thullum-172 w, all compounds

.7E+2 1E-3 LLI wall (86E.2)69 Thuliom-173 W. all compounds 4E63 1E+4 69 Thullam-175 2 W,.all compounds 7E+4 3E+5 St.. wall (9E+4) -70 Ytterbiom-162 2 W, all compounds except those given for Y 7E+4 3E+5 Y. oxides, hydroxides, and fluorides 3E+5 70 Ytterbium-166 W, see 1 6 2 Y 163 2E+3 Y, see Yb .2E+3 70 Ytterbiur167 2 W, see 162Y 3E+S BE6S Y, see 1 6 2 7E+5 70 Ytterbium-169 W, see 1621o 2E+3 8E+2 Y, see -7E+2 70 Ytterbium-175 W, see 1 6 2 Yb 3E+3 4E+3 LLI wall (3E.3) -Y. see 162yb( 3E-3 70 Ytterbium-177 2 W. see 1 6 2 6+Y. sea 1 6 2 Yb E+4 5E64 Yse 5E+4 70 Ytterbium-178 2 W, see 162yb 1+4 4E+4 Y, see 1 6 2 Yb -4E4 1E-6 4E-9 -* -SE-5 5E-4 4E-6. IE-9 5E-5 SE-4 6E-7 2E-9 ----2E-S 2E-4 1E-4 4E-7 1E-3 1E-2 6E-6 2E-8 6E-5 6E-4 8E-7 3E-9 3E-5 3E-4 9E-8 36-10 1E-5 1E-4 1E-7 --8E-10 2E-4 ZE-3 5E-7 2E-9 ---1E-5 iE-4 5E-6 2E-8 6E-5 6E-4 1E-4 4E-7 ---1E-3 1E-2 1E-4 4E-7 1E-3 1E-2 1E-4 4E-7 8E-7 3E-9 2E-5 2E-4 8E-7 3E-9 3E-4 1E-6 4E-3 4E-2 3E-4 1E-6 --4E-7 1E-9 2E-5 2E-4 3E-7 1E-9 1E-6 5E-9 -1E-6 5E-9 4-5 5 76-8 26-4 26-3 2E-S 6E-8 --2E-5 6E-8 2E-4 2E-3 2E-5 5E-B --B-31 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. R Table I Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3. Col. I Col. 2 Oral monthly Ingestion Inhalation Average ALl 'AL IT AT Air Water Concentration (pCi) (PCi) (pCi/ml) (pCi/ml) (p0i/em1l (RCf/ml)[(I l Atomic Radionuclide Class No.71 Lutetium-169 W, all compounds except those given for Y Y, oxides, hydroxides, and fluorides 71 LIutetiuma-170 W.see 6 169Lu Y. see Lu 71 Lutetium-171 W, see 169LU Y, see Lu 71 LutetiLm-173 W, see 169i Y see 169 , 71 Lutetium-171 W, see 169 Yse169 L Y, see Lu 169.71 Lutetium-174 W, see 16Lu Y. see 71 Lutetium-173.

W, see 169 Y, see. Lu 71 Lutetium176 W, see 169Lu Y, see 1 6 9 Lu 71 Lutetium-1 7 4 W, see 1 6 9 Lu Y', see 1 6 9 Lu 169 71 Lutetium,177 W, see 169 Lu Y, see Lu 71 Lutetiua-176 2 W, see 169Lu Y, see 169Lu 71 Lutetium'1772 W, see 169 Lu Y, see 169Lu 71 Lutetiumr179 W, see 1 6 9 LU Y. see 169Lu Yse16Lu 3E-3 4E+3 2E-6 6E-9 -1E+3 2E+3 9E-7 3E-9 2E-S 2E-4 2E-3 BE-7 3E-9 -2E+3 2E+3 BE-7 3E-9 3E-5 3E-4 2E+3 SE-7 3U-9 -1E+3 1E-3 5E-7 2E-9 1E-5 IE-4 1E-3 SE-7 2E-9 SE+3 3E-2 1E-7 -7E-5 7E-4 Bone surf (5E12) -6E-10 3E+2 IE-7 4E-10 2E-3 2E+2 1E-7 -LLI wall Bone surf (3E+3) (3E+2) 5E-10 4E-5 4E-4-2E+2 9E-8 3E-10 --5E+3 1E?2 5E-B -7E-5 7E-4 Bone surf (2E-2) -- It-0 --2E-2 6E-8 2E-1O -BEt3 3Et4 2E-5 3E-B 1E-4 iE-3 2t14 9E-6 3E-8 --7Et2 SEvO 2E-9 iE-S 1E-4.Bone surf (iE+l) -2E-11 -8E+O 3E-9 1E-11 -7Ei2 1E+2 5E-B -1E-5 1E-4 Bone surf (it.2) -2E-10 --BE+i 3E-8 1E-10 --2E-3 2E*3 9E-7 3E-9 --LLI wall (lEvI) ---4E-5 4E-4 3 ?2E+3 9E-7 3E-9 E 4 5E-4 2E-5 8E-S 3E-7-St. wall (6E+4) ---BE-4 BE-3 2E+5 7E-5 2E-7 -4E-4 1E+5 5E-S 2E-7 -St. wall (4E+4). -6E-4 6E-3-1E+5 5E-5 2E-7 6E+3 2E+4 BE-6 3I-8 9E-5 9E-4 2E+4 6E-6 3E-8 --4E-3 2E-6 6E-9 3E-5 3E-4 B-32 ODCM Rev. 25 App. B PART 20 STANDARDS-FOR PROTECTION AGAINST RADIATION APP. B Table 1 Table 2 Table 3 Occupational Values, Effluent -Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average ALl ALI 7 AC" Air Water Concentration (pCi) (p(i) (pCi/(l) (pi/ml) (pCi/ml) (pCi/ml)L D C L g Atomic Radionuclide Class No.72 Hafniut-170

.0, all compounds except those given for W W. oxides, hydroxides, carbides, and nitrates 72 Hafniam-172

0. see 170Hf W, see 17IHf 72 Hafnium-173 0, see 170.1 W, see 170Hf.72 Hafnium-175

0. see 1!70 Hi W, see 17011f 72 Hafnium- 177.2 D, see 1 r W, see Hf 72 Hafnium-178 0,. see 1701Hf W, see 170OHf 72 Hafnium--iTi 0, see 17011?W, see 170wi?72 Hafnium-rBO w 0, see 17011 W, see 17011f 72 Hafnium-181 0, see 1701Hf W, see 170f11 72 Hafnium-182m 2 O, see 170%W, see 1701f 72 Hafnium-182 0, see 170Hf W, see 17011, 72 Hafniam-1B3 2 0, see 17011 W, see 17011 72 Hafnlum-iB4 o, see 170111 U, see 17011 3E+3 6E-3 2E-6 8E-9 4E-5 4E-4-5E+3 tE-6 6E-9 1E+3 9E-O 4E-9 2E-5 2E-4 Bone surf* -(2E-1) -3E-11 4E+1 HE-8 Bone surf (6E-1) -tE-11 5E-3 1i-4 SE-6 2E-B 7E-S 7E-4 1E+4 SE-6 12E-8 3E+3 9E.2 4E-7 -4E-5 4E-4 Bone surf (1E+3) -it- -iEt3 .SE-7 2E-9 -2ES4 6E+4 tE-5 BE-8 3E-4 3E-3 9E-4 .SE- 1E--7 3E+2 .tE+O SE-10 -3E-I 3E-5 Bone surf (2E+O) -3E-12 5E+O 2E-9 Bone surf (9EsO) -IE-11 1Ei3 3E+2 iE-7 .-it-5 1t-4 Bone surf (6E42) -BE-10 ---6E+2 3E-7 BE-10 -7E-3 2i+4 9E-6 3E-8 it-4 1E-3 3Et4 iE-5 4t-B "-1Et3 2E-2 7E-B 2E-5 2E-4 Bone surf-(4E.2) -6E-10 ---4E+2 2E-7 6E-10 -.4E+4 9E-4 4SE-5 t-7 SE-4 SE7-3 1E+5 6E-5 2E-7 -2Et2 BE-1 3E-10 Bone surf Bone surf (4E+2) (2E.O) -ZE-12 SE-6 5E-S 3E.O 5j-5-9 --I Bone surf t (7E+O)' -1E-11-2E4 SEE4 2E-S 6E-B 3E-4 .3-3 6E÷4 2E-5 8E-8 i-2E+3 BE+3 3E-6 IE-8 3E-5 .3E-4.6E-3 3E 9E-9 -B-33 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col, 2 Col. 3 Col. 1 Col. 2 Oral Monthly ingestion Inhalation Average ALI tL UXC Air Water Concentration (pCi) (0Ci). (pCi/ml) (pCi/ml). (ijCi/ml) (pCi/ml)Atomic Radionuclide Class No..to U-Io U, 73 Tantalum-172 2 W, all compounds except those given for Y Y; elemental Ta, oxides, hydroxides, halides, carbides, nitrates, and nitrides 73 Tantalum-173 W, see 1 7 2 a Y. see 1 2 Ta Y: a1721'73 Tantalum-174 2 W see172a Ysee'Y 5:172 T 73 Tantalum-175 E see 177 Y , see 17 To 73 Tantalum-176 W, see 172. T Y. see 172T 73 Tantalu-177 W see7 see 172T 73 Tantalum-178 W see 172 Y: see 172 T 73 Tantalu-179 , see 7 a Y. see12T s 172_V : ac 7 Ta 73 Tantalum-1Bm W , see. 1 7 2 Ta Y, see 17 Ta 73 Tantalum-1B2 ,o see 717 Yý se 172 /e 73 Tantalw-rl82m2 W, see 172Ta Y. see 172Ta 73 Tantalum-2 , see 172 Y, see 172Ta 73 Tantalu-l183 W, see 172Ta see 72Ta 73 Tafltaluir184 W, e 172e Y, see:17 Ta 73 Tantalum-185 2 W, see 172 Y , s e e 1 7 2T 73 Tantahar].862 W. see 172 Ta Y. see 172 Ta 74 Tungsten-176

0. all compounds 74 Tungsten-177

0. all compounds 4E+4-1E+5 7E+3 2E+4-2E+4 3E+4 1E-5-9E+4 6E+3 2E+4-1E+4 4E+3 1E+4-1E-4 IE+4 2E+4 2E+4 2E24 9E+4-7E2+4 2E+4 SE+3 9E+2 2E+4 7E+4-6E+4 1E+3 4E+2-2E+1 2E+S SE'5 St. wall (2E+5)* 4E+5 BE+2 3E+2 1E+2 9E+2 1E23 LLI wall (1E+3)S" 1E+3 2E+3 SE*3 5E+3 3E+4 7E+4-:- 6E-4 SE+4 2E+5 St., wall (7E24)2E+5 1E+4 5E+4 4E-5 8E-6 7E-6'4E-5 4E-5 7E-6 6E-6 5E-6 5E-6 BE-6 7E-6 4E- 5 3E-5 2E-6 4E-7.3E-5 2E-5 2E-7 1E-8 2E-4 2E-4 1E-7 6E-8 5E-7 4E-7 2E-6 2E-6 3E-S 3E-5 1E-4 9E-5 2E-S 1E-7 3E-8 2E-8 1E-7 1E-7 2E-8 2E-8 2E-8 2E-8 3E-8 2E-8 1E-7 1E-7 8E-9 1E-9 8E-8 6E- 10 3E-11 8E-7 6E-7 SE-10 2E-10 2E-9 1E-9 7E-9 1E-7 9E-8 3E-7" 3E-.7 7E-8" 9E-5 4E-4 BE-5 5E-5 2E-4 2E-4 3E-4 3E-4 2E-5 3E-3 1E-S 2E-S 3E-S 4E-4 1E-3 1E-4 9E-4 4E-3 BE-4 5E-4 2E-3 2E-3 3E-3 3E-3 2E-4 3E-2 1E-4 2E-4 3E-4 4E-3 1E-2 1E-3 1E÷S 5E-5 .2E-7 5E-4 5E-3 2E+4 9E+4 4E-5 1E-7 3E-4 3E-3 B-34 ODCM Rev. 25 App. B App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION U, fD co Atomic Radionuclide Class No.74 Tungsten-178 0, all compounds 74 Tungsten-179 2 0, all compounds 74 Tungsten-181 0, all compounds 74 Tungsten-185 0, all compounds 74 Tungsten-187 0, all compounds 74 Tungsten-18 0, all compounds 75 Rhenium-177 2 D, all compounds except those given for W W, oxides, hydroxides, and nitrates 75 Rhenium-178 2.0, see 177Re W:s.W, see 177Re 75 henum 18 D:0, 177.W e177:e 75 Rhenium-182 o, see 177Re (12.7 h) W, see Re75 Rhenium-182 D, see 177ie (64.0 h) W, see Re 75 Rheni~u-184m 0, see 1 7 7 Re W, see 1771: 177.75 Rheniuu-184 D, :e, 177 W, see Re 75 Rheniuu-186m 0, see 177Re W. s 177Re 177-75 Rhenium-186 D, see 1 7 7 Re W, see Re 75 Rhenium-187 0, see 1 7 7 Re W, see'17.7 Re 75 Rhenium-18SM 2 ., see 1 7 7 Re W,osee Re W :177.~75 Rheniu-188 O, see 177Re W, see

Re Table 1 u1 Table 2 Table 3 Occupational Values -Effluent Releases to Concentrations Sewers Co1. 1 Col. 2 Col. 3 Col. 1 Col. 2 Oral " Monthly Ingestion Inhalation Average ALI -ALI OAC Air Water Concentration (pCi) (pCi) (pCi/ml) (pCi/m ) (pCi/ml) ( tCi/ml)5E+3 2E-4 8E-6 3E-8 7E-5 7E-4 5E-5 2E-6 7E-4 2E-6 7E-3 7E-2 2E64 3E-4 1E-5 5E-8 2E 2E-3 2E-3. 7E-3 3E-6 9E-9 LLI wall (3E+3) "- -4E-5 4E-4 2E+3 9E.3 4E-6 1E-8 3E-5 3E-4 4E62 1E63 5E-7 2E-9 " LI- wall (5E+2) -7E-6 7E-5 9E+4 3E-5 1E-4 4E-7 St. wall (16E5) .- 2E-3 ZE-2 4E+5 1E-4 5E-7 7E+4 3E.5 16-4 4E-7 --St. wall (1E+5) --" -1E-3 1E-2 3E+5 1E-4 4E-7 .5E+3 9E.3 4E-6 1E-8 7E-5 7E-4-9E.3 4E-6 1E-8 7E+3 1E+4 5E-6 2E-8 9E-5 9E-4-2E+4 6E-6 2E-8 -1E43 2E63 1E-6 3E-9 2E-5 2E-4-2E+3 9E-7 3E-9 --ZE+3 3E63 1E-6 4E-9 3E-5 3E-4 4E+2 2E-7 6E-10 -2E-3 4E+3 1E-6 5E-9 3E-5 3E-4 16E3 6E-7 2E-9 -1E+3 2E+3 7E-7 St. wall St. wall (2E63) (2E+3) -3E-9 2E-5 2E-4 2E62 6E-8 -2E-10 --2E63 3E63 1E-6 4E-9 3E-5 3E-4-2E63 7E-7 2E-9 6E+5 8E+5 4E-4 8E-3 8E-2 St. wall (9E65) -1E-6 -16E5 4E-5 1E-7 -8E+4 1E65 6E-5 2E-7 1E-3 1U-2* 1E+5 6E-5 2E-7 *2E+3 3E+3 1E-6 4E-9 2E-5 2E-4 3E+3 1E-6 4E-9 --B-35 ODCM Rev, 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average Atomic Radionuclide Class ALI A F DAC Air Water Concentration

'No. (pCi) (pCi) (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml)r, Cc)r--U.c 75 Rhenium-189 76 Osmium-180 2 76 Osmium-181 2 76 Osmium-182 76 Osmium-185 76 Osmium-189M 76 Osmium-lglm 76 Osmium-191 76 Osmium-193 76 Osmtum-194.

77 lridiun-182 2 77 Iridium-184 0D see 1 7 7 Re 3E-3 5E+3 W, see 1 7 Re -4E+3 0, all compounds except those given for W and Y 1E-5 4E+5 W, halides and nitrates -5E+5 Y, oxides and hydroxides

-5E-5 D, see oS 1E-4 4E-4 W, see 1 8 0 0s 5E-4 Y. see 0s 4E-4 0,on 8'0s 2E+3 61.3 0, see 1 8 0 Os 2 4E+3 W .see 1 8 0 0s4 Y, see 04E+3.se 1800- 2E13 5E+2 W, see 1 8 0 0s 1, see oUS 8E+2 Y, see 0 8E÷2 D, see 1800s 8E+4 2E15 W. see as 2E+5 Y, see 1 8 0 0s 2E13 0, see 1 8 0 0s IEa4 3E+4 U, see ~ ~Os -E31.3 Y, see 1 8 0 O s 2E+4 D,- see 10Oos- 2E+3 ZE+3 LLI wall 180- (3E+3) -0, see 1u2E+3.Y, see s 181 E+3, 0, see 18005s 2E-3 8E3 LLt wall Se o (4E.3) -V, se 180 an r 3E+3 Ysee 3E83 0, see 1800s 4E+2 4E+1 LLI w 3ll W ~,180 M6-2). -11, se as 6E+1 Y, se 18-0as 8E+O 0. all compounds except ,those given for W and Y 4E+4 1E+5 St. wall (4E+4)W, halides, nitrates, and metallic fridiuml 2E+5 Y. oxides and hydroxides 1E+5 0 .. 182t 8E-3 2E44 w. : see 2!'=rxr *3E-4 Y , see 18 ..3E-4 2E-6 7E-9 4E-5 4E-4 2E-6 6E-9 2E-4 5E-7 1E-3 1E-2 2E-4 7E-7 -2E-4 6E-7 -.2E-5 6E-8 2E-4 2E-3 2E-5 6E-8 2E-5 6E-8 2E-6 8E-9 3E-5 3E-4 2E-6 6E-9 -2E-6 6E-9 -2E-7 7E-10 3E-5 3E-4 3E-7 1E-9 -3E-7 1E-9 -1E-4 3E-7 1E-3 1E-2 9E-5 3E-7 -7E-5 2E-7 1E-5 4E-8 2E-4 2E-3 8E-6 3E-8 7E-6 2E-8 " " 9E-7 3E-9 -" 3E-5 3E-4 7E-7 2E-9 -6E-7 2E-9 -2E-6 6E-9 -E- 5 24-4 1E-6 4E-9 --2E-8 6E-11 ---8E-6 8E-5 2E-8 8E-11 -3E-9 1E-11 -6E-5< 2E-7 -6E-4 .. 6E-3 6E-5 2E-7 -5E-5 2E-7 1E-5 3E-8 1E-4 IE-3 1E-5 5E-8 -1E-5 4E-8 -B-36 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. 1. Col. 2 0ral Monthly Ingestion Inhalation

.Average Atomic Radionuclide Class AL I AL[ i C Air Water Concentration No. (pON) (pCi) (pCi/ml) (pCI/ml) (pCI/el) (pCi/ml)U1, 77 Iridium-185 w, see 1 8 2 Er W , s ee 182 Jirr Y, see r 77 Iridium-186 0 see 182.2r W, see 182ir Y, see 182 77 Iridium-187 D see 12Ir W, see 1 8 2: see 2r 182.77 lridium-188

, see 1821?W see 182frr OS1821?Y, see Sr 77 Iridium-189 0, see 182Ir Y. see 1821r-Y,u ee S 77 Iridium-l90m 2 0 see 182r hi, see 1821?Y. see Ir 182 i' see 1821?r Y, see 77 Iridium-192 D see 1821r W e: 182!?W, see1 8 21r 1821,¥,see 77 Iridium-192m 0 see 1821?2 182 r Y.ue 2!?r 77 lridium-194 0, see 182.1r hi; se 1821?r Y,usee W 182 I 77 Iridlum-194m D, see 1!2.r W see 1821r Y, see Ir 77 Iridlum-195 0 see 182 1821r V. see Ir 7 ln 182 77 Iridium-195 D, ee18Ir Y, s.e + Ir 78 Platinum-186 D, all. compounds 78 Platinum-188 all 1 compounds 78 Platinum-189 D, all compounds 78 Platinum-191 D, all compounds 51-3 1E+4 1E+4 1E+4 2E-3 8E+3 6E-3 1E+4 3E+4 3E+4 3E+4 2E+3 5E+3 4E-3 3E+3 SE83 5E43 LLI wall (5E+3) -4E+3 4E+3 2E+5 2E+5 2E-5 2E-5 1E13 9E12 9E+E 3E3S 9E+1 2E+2-2E+1 9E12 3E+2 4E+2 2E+2 6E+2 9E-1 2E+2 1E+2 1E+3 3SE3 2E+3 8E-3 2E+4--3E+4 2E+4 11+4 4E+4* 5E+4 4E+4 1E+4 4E'4 2E'3 2E+3 1E+4 3E+4 4E+3 8E+3 5E-6 2E-8 7E-5 7E-4 5E-6 2E-8 -4E-6 1E-8 -3E-6 1E-8 3E-5 3E-4 35-8 91-9 --2E- 8E-9 --15-5 5E-8 1E-4 1E'3 11-5 4E-8 5 4E-8 -2E-6 6E-9 3E-5 3E-4 1E-6 5E-9 -1E-6 5E-9 6 7E-9 -7E-5 7E-4 2E-6 5E-9 --11-6 5E-9 8E-5 3E-7 2E-3, 2E-2 9E-5 3E-7 -8E-5 3E-7 -4E-7 1E-9 1E-S 1E-4 4E-7 1E-9 --4E-7 1E-9 --4E-8 1E-10 4E-5 4E-4 9E-8 3E-10 -6E-9 2E-11 -1E-7 4E-10 1E-5. 1E-4 2E-7 6E-10 --9E-8 31-10 --4E-8 IE-10 9E-6 9E-5 7E-B 2-10 -E 4E-8 5-10 -E 1E-6 41-9 1E-5 IE-4 9E-7 3E-9 8E-7 3E-9 1E-S '3-8 1E-4 1E-3 18-5 45-8 -" 9E-6 3E-8 2E-S 65-8. 2E-4 2E-3 2E:5 7E-8 --2E-S 6E-8 --25-S 5E-8 2E-4 2E-3 71-7 2E-9 2E-5 21-4 11-S 45-8 1E-4 11-3 4E-6 1E-8 S-S 5E-4 B-37 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average AtLI AI UAZ Air Water Concentration (pCi) (pCi) (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml)Atomic Radionuclide Class No.78 Plati.um-193 D. al COpouds 78 Platinum-193 D, all compounds 78 .. Platinum-195m D, all compounds LLI wall (3E+4)4E84 2E+4 LLI wall (5E+4) -2E÷3 4E-3 LLI wall (2E+3) -it-b uft-1E-5 3E-8 2E-6 6E-9 4E-5 4E-4 6E-4 6E-3 U-U, 78 Platinum-197m 2 0, all compounds 2E+4 78 Platinum-197 D, all compounds 3E-3 78 Platinum-199 2 D, all compounds 5E+4 78 Platinum-200 0, all compounds 19+3 79 Gold-193 0, all compounds except-those given for W and Y 9E+3 W, halides and nitrates Y, oxides and hydroxides 79 Gold-194 , see 1933 W :193'u 3E+3 W, see 1 9 3 Au.Y, see Au 79 Gold-195 D see 1 9 3 Au 5E+3 Sw se: --1 g 3 Au Y, see Au 79 Gold-198m 0 see 3 W, se 1 9 3 Au 1E+3 se 1 9 3.u 79 Gold-198 0 see 19 3 Au 1E3 W, se 193u ,see *-P g, see 193Au 3+79 Gold-199 0. see 1 9 3 AU 3E+3 LLI wa W, see 1 3 Au3 V see 79 Gold-200m 0 see 1 9 3 3 193Au, 1+3 y, s. 193ku 79 Gold-200 2 0 see 1 9 3 Au W 193. 3E+4 70 see 1 9 3.Au7 79 Gold-2012 D, see 193 Au 7E÷4" 3E-4 2E+4 2E+4 8E+3 SE+3 SE+3 19÷4 1E43 4E+2 3E-3 1E+3* 4E+3 2E+3 2E+3 9E+3 Ill 4E83 4E+3 4E÷3 3E+3 2E+4-6E+4 8E+4 7E+4 2E+5 4E+4 1E+4 3E+3--3E-5 3E-4 28-5 6E-8 2E-4 2E-3 4E-6 1E-8 4E-5 4E-4 6E-5 2E-7 7E-4 7E-3 1E-6 5E-9 2E-5 2E-4 1E-5 4E-8 1E-4 1E-3 9k-6 3E-8 -.BE-6 3E-8 -3E-6 1E-8 40-5 4E-4 2E-6 8E-9 .-2E-6 7E-9 -5E-6 2E-8 7E-S 7E-4 6E-7 2E-9 --2E-7 6E-10 -.1E-6 4E-9 1E-5 1E-4 5E-7 2E-9 -5E-7 2E-9 -2E-6 5E-9 2E-5 2E-4 8E-7 3E-9 -7E-7 2E-9 -4E-6 1E-8 ---48-S 48-4 2E-6 6E-9 --2E-6 5E-9 1E-6 5E-9 28-5 2E-4 1E-6 4E-9 -1E-6 3E-9 -3E-5 9E-8 4E-4 4E-3 38-5 1E-7 3E-5 1E-7 9E-s 3E-7-1E-3 1E-2 1E-4 3E-7 -9E-5 3E-7 -193. AYesee St. wa (9E84)B-38 ODCM Rev. 25 App. B 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B* Table I Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. I Col. 2 Oral Monthly Ingestion Inhalation Average ALI ALl DAC Air Water Concentration (pCi) (pCi) (pCl/ml) (pCi/ml) (PCi/ml) (pCi/ml)-8E-3 AE-6 1E-8 -.4E*3 1E-4 5E-6 2E08 6E'5 6F-4 3E÷3 9E-3 4E-6 1E-8 4E-5 4E-4 L IC Atomic Radionuclide Class No.80 Mercury-193m Vapor Organic D D, sulfates W, oxides, hydroxides, halides, nitrates, and sulfides 80 Mercury-193 Vapor.Organic 0 0: e 193 H...Os*193. lg V, see Hg 80 Mercury-194 Vapor Organic D o. see V, see'9 "ig 80 .ercury-195 Vapor Organic O0 Os:e 13 g 80 Mercury-195 Vapor Organic O.. see 193N W:see 1 93 Ng 80 Mercury-197m Vapor Organic 0 O, se 1993~W, see 80 Mercury-197 Vapor Organic .o see 193m._W: see 1 9 3S-g 80 Mercury- 199m 2 Vapor Organic 0 0see 193m W: se;e13Hg 80 Mercury-203 Vapor Organic D D, se.ILg W, see -'g" 81 Thallium-194u 0, all coIpoands 2E+4 2E+4 2E÷1 BE+2 3E+3 2E+3 2E-4 1E÷4 4E+3 3E+3 7E+3 6E-3 6E÷4 St. wall (1EH5)6E-4 5E+2 2E-3 5E+4 St. wall (7E+4)-8E+3 3E+4 6E+4 4E+4 4E+4 3E+1 3E+1 4E+1 1E+2 4E+3 6E+3 5E-3 4E-3 3E-4 5E+4 4E+4 3E+4 5E-3 9E+3 7E.3 5E-3 8E*3 1E+4 1E+4 9E+3 8E4 2E+5.1E.S 2E+5 8E-2 8E-2 1E-3 1E-3 2E÷5 3E-6 1E-8 * -1E-S 4E-8 --3E-5 9E-8 3E-4 3E-3 2E-5 6E-8 2E-4 2E-3 2E-5 6E-8 -1A-8 4E-11 -1E-8 4E-11 2E-7 2E-6 2E-8 6E-11 1E-5 3E-4 5E-8. .2E-10 ..2E-6 6E-9 3E-6 8E-9 4E-5 4E-4 2E-6 7E-9' 3E-5 3E-4 ZE-6 5E-9 1E-5 4E-8 2E05" 6E-8 2E-4 2E-3 1E-5 5E-8 2E-4 2E-3 1E-5 5E-8 2E-6 7E-9 -" 4E-6 1E-8 SE-5 5E-4 3E-6 1E-8 4E-5 4E-4 2E-6 7E-9 " 4E-6 1E-8 -6E-6 2E-8 9E-5 9E-4.SE-6 2E-8 .8E-5 8E-4 4E-6 1E-8 30-5 1E-7 7E-5 2E-7--1E-3 1E-2 6E-S 2E-7 8E-4 8E-3 7E-5 2E-7 --4E-7 1E-9 -3E-7 1E-9 7E-6 7E-5 5E-7 2E-9 3E-S .3E-4 5E-7 2E-9 6E-5 ZE-7.--1E-3 1E-2 B-39 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Table I Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average Atomic Radionuclide Class ALl ALI T AC Air Water Concentration No. (pCi) (pCi) (pCi/ml) (pCI/ml) (pCi/ml) (tCi/ml)82 Thallium-194 2 0, .all compounds 81 Thallium-15 2 0, all compounds 81 Thallium-197 D, all compounds 81 Thallium-198m 2 0D, all compounds 81 Thallium-198 81 Thallium-199 81 Thallium-200 81 Thalliom-201 81 Thallium-Z02 8i Thalliwm-204 82 Lead-195mz 82 Lead-198 82 Lead-199 2 82 Lead-200 82 Lead-201 82 Lead-202m 82 Lead-202 82 Lead-203 82 Lead-205 82 Lead-209 82 Lead-210 82 Lead-211 2 82 Lead-212 82 Lead-214 2 83 Bismuth-200 2 83 Bismuth-2012 83 Bismuth-202 2 D, all compounds 0, all compounds D, all compounds 0, all compounds 0, all compounds 0, all compounds D, all compounds D, all compounds 0, all compounds 0, all compounds 0, all compounds D. all compounds 0, all compounds D, all compounds 0, all compounds D, all compounds D, all compounds 0, all compounds D, all compounds 0. all compounds 0, nitrates W: all other compounds 0, see 20oBi W, see 200 Bi 0,s 200 , see B 3E-5 6E+S 2E-4 8E-7 St. wall (3E.5) -4E-3 4E-2 6E+4 11+5 5E-5 2E-7 9E-4 9E-3 7E+4 10+5 5E-5 2E-7 1E-3 1E-2 3E04 5E+4 2E-5 8E-8 4E-4 4E-3 2E04 3E04 1E-5 5E-8 3E-4 3E-3 6E+4 8E+4 4E-5 1E-7 9E-4 9E-3 8E+3 1E+4 5E-6 2E-8 1E-4 1E-3 2E+4 ZE+4 9E-6 3E-8 2E-4 2E-3 4E+3 5E+3 ZE-6 7E-9 5E-5 5E-4 2E+3 2E+3 9E-7 3E-9 2E-5 2E-4 6E+4 2E+5 8E-5 3E-7 8E-4 8E-3 3E+4 6E+4 3E-5 9E-8 4E-4 4E-3 2E04 7E+4. 3H-5 11-7 3E-4 3E-3 3E+3 6E.3. 3E-6 9E-9 4E-5 4E-4 7E-3 2E.4 8E-6 3E-8 11-4 11-3 9E+3 3E+4 .1E-5 4E-8 2E-4 1E-3 1E+2 5E+1 2E-8 7E-11 2E-6 2E-5 5E-3 9E+3 4E-6 1E-8 7E-5 7E-4 4E+3. 11.3 6E-7 2E-9 5E-5 5E-4, 2E+4 6E÷4 2E-5 8E-8 3E-4 3E-3 6E-1 2E-1 1E-10 -Bone surf Bone surf (10E+O) (4E-1) -6E-13 11-8 1E-7 1E+4 6E÷2 3E-7 9E-10 2E-4 2E-3 8E+1 3E.1 1E-8 kE-11" Bone surf (1E+2) -2E-6 2E-5 9E÷3 BE2 3E-7 11-9 1E-4 1E-3 3E04 8E04 4E-5 .1E-7 4E-4 4E-3-1E+5 .4E-5 1E-7 -.1E+4 3E÷4 1E-5 4E-8 2E-4 2E-3 4E÷4 2E-5 5E-8 10E4 4E÷4 2E-5 6E-8 2E-4 2E-3 BE4 3E-S 1E-7 --B-40 ODCM Rev. 25 App. B App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 .Col. 3 Col. 1 Col..2 Oral Monthly Ingestion Inhalation Average Atomic Radionuclide Class ALI ALI DAC ..Air Water Concentration No. (pCi) (pCO ) (pCi/al) (pCi/ml) (pCi/ml) (pei/el)r L*[*83 Bismoth-203 D see 200"" W: see 2 83 Bismuth-205 D: see 200* 20081.W, see 83 Bismuth-206 0 see 20081 se 200 8 83 Bismuth-207 0 see 20081 W, see20 83 Bismuth-210m

0. see 200B1 W, see 20081 83 8ismuth*210

0. see 200BiW, see 200 B$w.~ 20081 83 Bismuth-212 2 D, see 200B U, see .8 W: 2100i 83 Bismuth-213 2 0 see 2 o 0 0i U, see ?OOI 83 Bismuth-214 2 D. see 200Bi W, see 20081 84 Polonium-203 2 0, all compounds except those given for W W, oxides, hydroxides, and nitrates 84 2 , ee 203 203-P 84 Polonium-207 D, see 203 W, see e Ws 203.o 84 Poloniun-210 D, see 2 0 3 Po W see Po 85 Astatine-207 2 0, halides W 85 Astatine-211 D, halides W 86 Radon-220 With daughters removed With daughters present 2E-3 7E-3 6E-3 1E+3 3E83 1E+3 6E+2 1E+3 9E+2 1E+3 2E+3 4E+2 4E+1 5E+O Kidneys Kidneys (6E.1) (6E80)7E-1 8E+2 2E+2 Kidneys-(4E+2)3E+1 5E+3 2E+2 3E+2 7E.3 38.2 4E+Z 2E+4 8E+2 St. wall (2E+4) -9E-2 3E+4 6E+4 9E-4 2E+4 4E84 7E÷4 8E+3 3E+4 3E+4 3E8O 6E-1-6E-1 6E-3 3E-3-2E+3 1E+2 8E+1 5E+1-. 2E+4 1E-6 5E-7 6E-7 4E-7 7E-7 1E-7 2E-9 3E-10 1E-7 1E-8 1E-7 18-7 1E-7 1E-7 3E-7 4E-7 3E-5 4E-5 2E-5 3E-5 1E-5 1E-S 3E-10 3E-10 1E-6 9E-7 3E-8 2E-8 3E-9 2E-9 2E-9 11-9 2E-9 5E-10 9E-12 9E-13 5E-10 4E-11 3E-10 4E-10 4E-10 5E-10 1E-9 1E-9 9E-8 I1-7 5E-8 1E-7 3E-8 4E-8 9E-13 9E-13 4E79 3E-9 18-11 8E-11 2E-S 9E-6 8E-7 11-5 7E-5 1E74 3E-4 3E-4 3E-4 1E-4 4E-8 SE-5 2£-6 2E-4 9E-5 18-4 8E-6 1E-4.7E-4 1E-3 3E-3 3E-3 3E-3 1E-3 4E-7 8E-4 2E-5 3E-6 9E-9 3E-5 3E-4 3E-6 9E-9 --7E-6 2E-8.2E+1 9E-9 3E-11 (or. 12 working (or 1.0 level months) working level)B-41 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 Col. 3 Col. I Col. 2 Oral Monthly Ingestion Inhalation Average Atomic Radionuclide Class AL I ALI DAL Air Water Concentration No. (PCi) (pCi) (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml)86 Radon-222 With daughters removed -1E+4 4E-6 1E9 -With daughters present 1E+2 3E-8 1E10 -App. B E£D E g 87 Francium-2222 0, all compounds 87 Francium-223 2 0, all compounds 88 Radiuo-223 W, all compounds 88 Radium-224 W, all compounds 88 Radium-225 W, all compounds 88 Radium-226 W, all compounds 88 Radium-227 2 W, all compounds 88 Radium-228 W, all compounds 89 Actiniumr-224 D, all compounds except those given for W and Y W, halides and nitrates Y, oxides and hydroxides 89 Actinium-225 D, see 224A, a 224A ,see 224 Ac y, see 89 Actinium-226 0, see 224Ac 224.W, see 224A Y:. see 22 89 Actinium-227 0, see 2 2 4 Ac W, .see 224Ac Y, see 2 2 4 Ac (or 4 working (or 0.33 level months) working level)2E-3 5E-2 2E-7 6E-2 8E-2 3E-7 5E+6 7E-1 3E-10 Bone surf (9E+O)8E+O 2E+O 7E-10 Bone surf (26E+)8E6O 7E-1 3E-10 gone surf (2E+1)2E+O 6E-1 3E-10 Bone surf (5E+O) -2E+4 1E-4 6E-6 Bone surf Bone surf ZE÷4) (2E+4) -2E+O 1E+O 5E-10 Bone surf (4E+O)2E÷3 LLI wall (2E+3)5E-I LLI wall (5E+1)1E+2 LII wall (16E2)ZE-1 Bone surf (4E-1)3E+1 Bone surf (4E61)5E-1 3E-1 Bone surf (SE-1)6E-1 6E-1 3E+O Bone surf (4E*O)5E+O 5E+O 4E-4 Bone surf (BE-4)2E-3 Bone surf (3E-3)4E-3 6E-10 1E-9 9E-13 2E- 12 96E-13 9E-13 3E-8 2E-12 5E-il 7E-11 SE-11 7E-13 9E-13 9E-13 5E-12 7E-12 6E-12 1E-15 4E-15 6E-15 3E-5 BE-6 1E-7 2E-7 2E-7 6E-8 3E-4 6E-B 3E-5 7E-7 2E-6 5E-9 3E-4 8E-5 1E-6 2E-6 2E-6 6E-7 3E-3 66E-7 3E-4 7E-6 2E-5 5E-8 1E-B 2E-B 2E-8 1E-10 3E-10 3E-10 1E-9 2E-9 2E-9 2E-13 7E-13 2E-12 B-42 ODCM Rev. 25 App. B App.B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION A B Table 1 Table 2 Table 3.Occupational Values Effluent Releases to Concentrations Sewers Col. I Col. 2 Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average Atomic Radionuclide Class ALI ALI OAD Air Water Concentration NO: (pCi) (pC ) (pCi/ml) (PCI/mi) (pCi/ml) (pCi/ml)89 Actinium-228 0, see 224Ac 2E+3 9E+O 4E-9 -3E-S 3E-4 Bone surf W, see 224 (2E-1) 2E-11 W se Ac4E+1 2E-8 Bone surf Y, see 224- (E+l) -BE-lI Ac 4E+1 2E-8 6E-11 -9D Thorlum-226 2 W, all compounds except those given for Y 5E+3. 2E-2 6E-8 2E-10 St. wall (5E-3) -7E-5 7E-4 Y, oxides and hydroxides

-iE*2 6E-B 2E-10 -90 Thorium-227 W. see225 1E+2 3E-I IE-10 5E-13" 2E-6 2E-5 Y, see 2 h 3E-1 IE-10 SE-13 90 Thorium-228 W, see 2 2 6 Th 6E+O lE-2 4E-12 -Bone surf Bone surf Y, see 226 (iEf1) (2E-2) -3E-14 2E-7 2E-6 Th 2E-2 7E-12 2E-14 90 Thoriua-229 W, see 2 2 6 Th 6E-1 9E-4 4E-13 -a, Bone surf Bone surf (+E'O) (2E-3) -X-i5 2E-B 2E-7 Y, see 226 ThE-3 1EU12 UL Bone surf cc (3E-3) -4E-15 90 Thorismo-230 W, see 2 2 GTh 4E+O 6E-3 3E-12 --Bone surf Bone surf 226 (9E-0) (2E-2) -2E-14 l1E-7 iE-G Ysee 6Th 2E-2 G 6E-12 Bone surf (2E-2) 3E-14 -90 Thoritm-231 SW, see 2 2 6 Thi Y, see 2 2 Th 4E+3 6E+3 3E-6 9E-9 5E-5 BE-4 6E-3 '3E-6 9E-g 90 Thorism-232 W, see 2 26Th 7E-i 1E-3 5E-13 -Bone surf Bone surf Y 2 2 6 Th (2E+O) (3E-3) 4E-15 3E-8 3E-7 Yse- 3E-3 1E-12 Bone surf (4E-3) -6E-15 --90 Thoriy-234 W, see 2 2 6 Th 3E-2 -2E+2 BE-8 3E-1O --LLI wall Y, see 2 2 6 Th (4E+2) E " -O 5E-6 -E-5 se26h2E+2 6E-8 2E-10 91 Protactlnium-2272 W, all compounds except those given for Y 4E3i 1E÷2 SE-B 2E-10 SE-5 SE-4 Y, oxides and hydroxidei -E+2 &E-B E-ig -O 91 Protactinium-22B W, see 2 2 7 Pa IE+3 1E+1 5E-9 2E-5 2E-4 Bone surf-p(2E+1) 3E-11 --Yse 2 2 6 Pa 1E+1 SE-9 2E11- --B-43 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B Table I Table 2 Table 3 Occupational Values AEffluent Releases to Concentrations Sewers Ca1. 1 Col. 2 Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average Atomic Radionuclide Class ALl ALI OAC Air Water Concentration No. (pjCi} (PCi) (SiCi/ml) (pci/ml) (pCi/al) (liCi/ml)91 Protactinium-230 W, see 227Pa 6E-2 5E+O 21-9 7E-12 Bone surf (9m) 1E-5 1E-4 YO, see 2 2 7 Pa 4E+O 1E-9 5E-12 -91 Protactinium-231 W,.see 2 2 7 pa 2E-1 2E-3 6E-13 Bone surf Bone surf 2 2 6 pa ( 1E-1) (4E-3) 6E-15 6E-9 6E-8 Ysee Pa- 4E-3 2E-12 --Bone surf (6E-3) -8E-15 91 Protactinium-232 W, see 2 2 7 Pa 1E+3 2E+1 9E-9 2E-5 2E-4 Bone surf Y, see 227- (6.1) -8E-11 -Pa- 6E.1 2E-8 " -Bone surf-(7E+1) 11-10 U1 91 Protactinium-233 W, see 2 2 7 Pa 1E+3 7E+2 3E-7 1E-9 LLI wall (- Yse(2E+3)

-2E-5 2E-4 Y6see 2 2 7 Ps 6E+2 2E-7 8E-10 91 Protactinium-234 W, see 2 2 7-Pa 2E+3 BE+3 3E6 1-B 3E-5 3E4 Ssee Pa 7E+3 3E-6 9E-. -92 Uranium-230

0. UFs, U 0 2F2, U02(NO 3)2 4E+O 4E-1 2E-10 Bone surf.. Bone surf (6E.O) (6E-1) -BE-13 BE-8 8E-7 W.UdO., UT.,, UCl., -4E-1 IC-10 SE-13 -Y, UO0, Us08 -3E-1 1E-10 4E-13 92 Uranium-Z31 0, see 230 5 E+3wl BE+3 3E-6 1E-8 LLI wall (4E+3) --6 BE-5 6E-4 W: :e: -10 6E+3 .2E-6 BE- -Y ee 2 U

5E+3 2E-6 6E-9 -92 Uranium-Z32 9, see 230U 2E*O 2E-I 9E-11 -Bone surf Bone surf (4E÷O) (4E-1) -6E-13 6E-8 6E-7 W, se: 230 U 4E-1 2E-10 5E-13 Y, see 2U BE-3 3E-12 1E-14 92 Uranium-233

0. see 23% 1E-1 1E.O 5E-l0 Bone suit Bone surf 2(2E1) (2E1O) 3E-12 3E-7 3E-6 W: e 230up 7E-1 3E-10 1E-12 Y, see 23 -4E-2 2E-11 SE-14 -92 Urani um-234 3 0, see 230U 1E+1 11EO 5E-10 Bone surf Bone surf (2E+I) : (ZE£0) -3E-12 3E-7 3E-6 see 2 3 0 U -7E-i 3-E-10 1E-12 -Y: see U 4E-2 2E-11 * .5E-14 -B-44 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION..

App. B Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col.. 2

Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average Atomic Radionuclide Class ALI ALtDAC Air Water Concentration

..No. (pli) (PCQ) (VCi/ml) (uCi/ml) (Pci/nl) (pCi/m])1, to CO IL 92 Uranuam-235 3 D, see 230 230.V, see 230U 92: Uranium-236 0, see 230, 230 Y, see 230 Y. see U 92 Uranium-237

0. see 230U 230.W, see 230U Y. see U 92 Uranium-238 3 D., see 230 230U W, see 23O0 Y see U 230.92 Uranium-2403 D, see 230 W see 230e YD see* " 230 Y, see 230U SY, see U 2 30 93 Neptunium-2322 W, all compounds.

93 Neptunium-233 2 W. all compounds 93 Neptuniua-234 W, all compounds 93 *eptunitu-235 W, all compounds 93 Neptunium-236 W, all compounds (1.15E+5 y), 93 Neptunium-236m W, all compounds (22.5 h)93 Neptunium-237 W, all compounds 1E+1 IE+O 6E-10 Bone surf Bone surf (2E+1) (2E+O) -BE-1 3E-10 4E-2 2E-)l 1E1 1E+O SE-10 Bone surf Bone surf (2E-1) (2E+O) -BE-1 3E-10 4E-2 2E-11 2E,3 3E-3 1E-6 LLI wall (2E+3) --2E-3 7E-7 2E-3 SE-7 1E÷1 1E+O 6E-10 Bone surf Bone surf (2E-1) (2E÷O) -8E-1 3E-10 4E-2 2E-11 7E+4 2E+5 BE-5 2E+5 7E-5 2E+5 6E-5 1E-3 4E+3 2E-6-3E-3 1E-6 2E÷3 1E-6 IEe1 lE+O 5E-10 Bone surf Bone surf (2E-1) (2E+0) --8E-1 3E-10-5E-2 ZE-11 1ES5 2E÷3 7E-7 Bone surf-(5E+2)BE+5 3E+6 HE-3 2E+3 3E+3 IE-6 2E+4 BE-2 3E-7 LLI wall Bone surf (2E+4) (1E+3)3E+6 2E-2 9E-12 Bone surf Bone surf (6E+O) (5E-2)3E+3 3E+1 1E-8 Bone surf Bone surf (4E+3) (7E+1) -5E-1. 4E-3 2E-12 Bone surf Bone surf (1E.O) (1E-2)3E-12 IE-12 6E-14 3E-12 1E_-12 6E-14 4E-9 2E-9 2E-9 3E-12 IE-12 6E-14 3E-7 2E-7 2E-7 SE-9 4E-9.3E-9 3E-12 9E-13 9E-14 6E-9 4E-6 4E-9 2E-9 BE- 14 3E-7 3E-7 3E-5 3E-7 9E-4 2E-5 XE-7 211-3 IE-2 3E- 5 3E-4 9E-8 3E-6 3E-6 3E-4 3E-6 9E-3 2E-4 3E-B 2E-2 IE-1 3E-4 3E-3 9E-7 IE-10 5E-5 5E-4 1E-14 2E-8 ZE-7 B-45 ODCM Rev. 25 App. B App. 13 PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION able I Table 2 Table 3 ptalesI Effluent Releases to Occupational Values Concentrations Sewers Col. Col. 2 Col. 3 Col. I Cal. 2 Monthly Oral Average Ingestionl Inhalation Ai atr Avrg ALi b AC n Water Concentration Atomic Radionuclide Class (p i) ( (pi) (PCt/Nl) (PC,/l pi/ml) .piml No. C 3- -93 Neptunium-239 W, all compounds IE-3 6E41 3E-8 Z2-5 j14 Bone surf.(21.Z) 2E-10 93 tNeptunium-2 3 9 W, all compounds ZE+3 2Ew3 9E-7 3E-9 LLI wall 214E-5 5E-4 (2E-3)82 3,nd ZE 4 8E+4 3E- B 1E-7 0 3E-4 3E-3 oc, EE 00 U, 93 Nleptunium-240-w, Al, ,-om...94 PlutoniUm-2 3 4 W, all compounds except PuO2 Y, Pu0o 94 Plutonium.235 W, see 234Pu Y: Ie PU 94 Plutoniua-2 3 6 W, see 2 4Pu Y, see 2APu 94 plutoniuLr 2 37 W, see see 234 U y, see r 94 Plutonlum-2 3 B W, see Pu y, see g34 pu.94 Plutoniumr239 W, "9+ 234 pu'Y, see 2 3 4Pu 94 Plutoniuir-2 4 0 W, see 2 3 4 Pu y, see 234pu 94 plutonium-241 W, see 2 3 4 Pu y, see 234Pu 8E+3 2E+ 2 9E .. .3E-10 1E-4* 2E+2 BE-B 3E-10 9E+5 3E-6 11-3 4E-6 1E-2 3E+6 1E-3. 3E-6 -2EO ZE-2 8E-12 Bone surf Bone surf (4E+O) (40-2) -5E-14 61-8 4E-2 2E-11 6E-14 iE+4 3E+3 1E-6 5E-9 2E-4 3E+3 1E-6 4E-9 -9E-1 7E-3 3E-12 -Bone surf Bone surf 2E-14 2E-8 (2E+O) (1E-2) " zE-2 BE-2 2E-14 -aE-1 6U-3. 3E-12 Bone surf Bone-surf

_ 1-14 ZE-8 EO (1E-21 7E-12 ZE-Z 7E1 Bone surf 2E-14 (2E-2) -BE-1 6E-3 3E-12 .Bone surf Bone surf 2E-14 2E-8 (1EO) (1E-2) "--21-2. 7E-12 -Bone surf (2E-2) 2E-14 4E+1 .3E-1 1E-10 -Bone surf Bone surf 81-13 11-6 (71E÷) (6E-1) -_ BE-1 3E-10 Bone surf.- (1E+O) -E-112 1A-1 6E-7 6E-7 ZE-3 2E-7 2E-7 1E-5 ODCM Rev. 25 B-46 App. B App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION Table I Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2 -Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average Atoic Radionuclide .Class ALI ALI OAC Air Water Concentration no. (PCi) (PCi) (Pct/ml) (lpCi/m~l) (PC i/ml) (Pci/mi)94 Plutonium-242 W, see 2 34Pu BE-1 7E-3 3E-12 -Bone surf Bone surf (IE-O) (1E-2) 2E-14 2E-8 ZE-7 Y, see 2E-2 7E-12 Bone surf (2E-2) -2E-14 " -94 Plutonium-243 W, see 234.234 u 2E÷4 4E+4 2E-5 5E-8 2E-4 ZE-3 Y, see Pu 4E+4 2E-5 5E-B --94 Plutonium-244 W, see 234pu BE-1 7E-3 3E-12 Bone surf Bone surf 234 (2E+O) (1E-2) -2E-14 2E-8 2E-7 Y, --. 2E-2 7E-12 -Bone surf (2E-2) -2E-14 94 Plutonium-245 W see 2 3 4 Pu 2E+3 5E43 2E-6 6E-9 3E-5 3E-4 Y, see Pu 4E+3 2E-6 6E-9 -94 Plutonium-246 W, see 2 3 4 pu 4E.2 3E+2 1E-7 4E-1 -LLI wall: 2 3 4 (4E+2) --BE-6 BE-S V. see 3 3E+2 1E-7 4E-10 95 Americium-237 2 W, all compounds BE+4 3E+5 1E-4 4E-7 1E"3 1E-2 95 Aaericiua-238 2 W, all compounds 4E+4 3E+3 1E-6 5E-4 5E-3 Bone surf S (6.E3) -9E-9 l -95 Amerilfue-239 W, all compounds SE+3 1E+4. 5E-6 2E-8 7E-5 7E-4 95 Aaericium-240 W, all.compounds 2E+3 3E+3 1E-6 4E-9 3E-5 3E-4 95 AmericiLmr241 W, all conpounds BE-1 6F-3 3E-12 Bone surf. Bone surf (11+0) (1E-2) 2E-14 2E-8 2E-7 95 Americian-242m W, all compounds BE-1 6E-3 3E-12 Bone surf Bone surf (IE+O) (1E-2) -2E-14 2E-8 2E-7 95 Americium-242 W, all compounds 4E+3 8E+1 4E-B 5E-S 5E-4 Bone surf* -. (9E+1) 1E-lO -95 Americium-243 W, all compounds 8E-1 6E-3 3E*12 -Bone surf Bone surf (1EmO) (1E-2) -2E-14. 2E-8 2E-7 95 Aaericitu-244%

2 W, all compounds BE+4 4E+3 2E-6 St. wall Bone surf (BE.4) (7E.3) IE-8 1E-3 1E-2 95 Americium-244 W, all compounds 3E+3 2E+Z BE-8 -4E-5 4E-4 Bone surf-(3E÷2) -4E-10 95 Aaericium-245 W. all compounds 3E+4 BE+4 3E-5 1E-7 4E-4 4t-3 B -47 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION.

Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Coil 2 Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average Atomic Radionuclide Class ALI ALI DAC Air Water Concentration I", (VCi) (C i) (lici/ml) (Vci/ml) VpI/M) (PUi/01)95 Americium-246m 2 W, all compounds 5E.4 2E-5 BE-S 3E-7 St. wall (6E+4) 7 8E-4 8E-3 95 Americium-246 2 W, all compounds 3E+4 1E-5 4E-5 IE-7 4E-4 4E-3*96 Curium-238 W, all compounds 2E+4 1E+3 5E-7 2E-9 2E-4 2E-3 96 Curium-240 W, all compounds 6E+1 6E-1 R2E-10 Bone surf Bone surf (BE+1) (6E-1) -9E-13 1E-6 1E-S 96 Curium-241 W, all compounds 1E+3 3E+1 1E-8 -. 2E-S 2E-4 Bone surf (4E.1) 5E-11 96 Curium-242 W, all compounds 3E+1 3E-1 1E-1O Bone surf Bone surf (5E+1) (3E-1) 4E-13 7E-7 7E-6 96 Curiur-243 W, all compounds 1E+O 9E-3 4E-12 -Bone surf Bone surf (2E÷O) (ZE-2) -2E-14 3E-B 3E-7 96 Curium-244 W, all compounds 1E+O 1E-2 SE-12, ---cc Bone surf Bone surf.(3E+O) (2E-2) 3E-14 3E-B 3E-7 96 Curium-245 W, all compounds 7E-1 6E-3 3E-12 --Bone surf Bone surf (1E+O) (1E-2) 2E-14 2E-8 2E-7 96 CuriLum-246 W, all compounds 7E-1. 6E-3 3E-12 Bone surf Bone surf (1E+O) (1E-2) -2E-14 2E-B 2E-7 96 Curium-247 W, all compounds BE-i, 6E-3 3E-12 -Bone surf Bone surf (1E+O) (1E-2) E2-14 2E-8 2E-7 96 Curium-248 W, all compounds 2E-1 2E-3 7E-13 -.Bone surf Bone surf (4E-1) (3E-3) -4E-15 5E-9 5E-B 96" Curiuum-249 2 W. all compounds 5E+4 2E+4 7E-6 7E;4 7E-3 Bone surf (3E+4) -4E-8 96 Curium-250 W. all compounds 4E-2 3E-4. 1E-13 Bone surf Bone surf (6E-2) (5E-4) BE-16 9E-10 9E-9 97 Berkelium-245 W, all compounds 2E+3 1E+3 5E-7 2E-9 *.3E-5 3E-4 97 Berkellum-246 W, all compounds 3E+3 3E+3 1E-6 4E-9 4E-5 4E-4 97 Berkelium-247 W, all compounds 5E-1 4E-3 2E-12 -Bone surf Bone surf (IE+O) (9E-3) -1E-14 2E-B 2E-7 97 Berkelium-249 W, all compounds 2E+2 2E-O. 7E-10 Bone surf Bone surf (5E+2) .(4E1O) 5E-12 6E-6 6E-5 App. B B-48 ODCM Rev. 25 App. B App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION Table 1 Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Severs Col. 1 Col. 2 Col. 3 Col. I Col. 2 Oral Monthly Ingestion Inhalation Average Atomic Radionuclide Class ALl Air Water Concentration N0. (pl)c (pici) (pCi/ml) (Pcitmt) (peti/mll (pet/ml)N£0 CD N Co U-Co U, 97 Berkellum-250 W, all compounds 98

Californlum-244 2 W, all compounds except those given for Y Y, oxides and hydroxides 98 Californitum-246 W,:see 244Cf Y, see 244 Cf 98 Callfornlium-248 W, see 244Cf Y, see 244Cf 98 Californium-249 W, see 244Cf Y, see 2 4 4 Cf 98 Callfornium-250 W, see 2 4 4 Cf Y, see 244Cf 98 Californium-251 W; see 244Cf Y. see 244Cf 98 Californium-252 W, see 2 44Cf Y, see 244 Cf 98 Californium-253 W, see 2 4 4 Cf Yse244(: 244f 98 Californium-254 W, see 244CF.Y, see Cf 99 Einsteiniu.-250 W, all compounds 99 Einsteinlum-251 W, all compounds 99 Einsteinium-253 W, all compounds 3E+4 St. wal l (3E+4)4E+2 8E+O Bone surf (2E+1)SE-1 Bone surf (1E+0)1E+0 Bone surf (2E+O)SE-1 Bone surf (1E÷O)2E.O Bone surf (SE+O)2E*2 Bone surf (4E+.2)2E+O 4E+4 7E+3 2E+2 6E+2 2E-7 6E+2 2E-7 9E+O 4E-9 9E+O 4E-9 6E-2 3E-11 Bone surf (1E-1)1E-1 4E-11 4E-3 2E-12 Bone surf (9E-3)IE-2 4E-12 Bone surf (lE-2) )9E-3 4E-12 Bone surf (RE-2) -3E-2 1E-11 4E-3 2E-i2 Bone surf (9E-3)IE-2 4E-12 Bone surf (1E-2). -2E-2 8E-12 Bone surf .(4E-2)3E-2 IE-11 2EfO 8E-10 2E+O 7E-1O 2E-2 9E-12 2E-2 7E-12 5E+2 2E-7 Bone surf (1E+3) -9E+2 4E-7 Bone surf (1E+3)1E+O 6E-i0 BE-lO BE-10 1E-11 IE-11 2E-13 1E-13 IE-14 2E-14 3E-14 4E-14 IE-14 2E-14 5E-14 SE-14 3E-12 2E-12 3E-14 2E-14 2E-9 2E-9 ZE-12 4E-4 5E-6 2E-7B 4E- 3 5E-5 2E-6 2E-7 9E+3 3E+2 1-7 1 11-4 1E-3 Bone surf 3E-8 3E-7 2E-8 2E-7 7E-B 7E-7 5E-6 5E-5 3E-B 3E-7 6E-4 6E-3 lE-4 IE-3 2E-6 2E-5 B-49 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION Table I Table 2 Table 3 Occupational Values Effluent Releases to Concentrations Sewers Col. 1 Col. 2. Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average Atomic Radionuclide Class AlI ALI DAC Air Water Concentration No. (pCi) (pCi) (pCi/ml) (pCi/ml) (pCi/nl) (pCi/ml)99 Einsteinium-254m W, all compounds 3E+2 1E+1 4E-9 1E-11 ELI wail (3E+2) 4E-6 .4E-S App. B 99 Einsteinium-254 W, all compounds 100 Fermium-252 W, all compounds 100 Fermium-253 W. all compounds 100 Fereium-254 W, all compounds 100 Fermium-255 W, all compounds 100 Fermium-257 W. all compounds 8E+O 7E-2 Bone surf Bone surf (2E+1) (1E-1)50E2 10E1 1E+3 1E+1 3E03 9E+1 5E÷2 2E+1 2E01 2E-1 Bone surf Bone surf (4E+1) (2E-1)7E+3 8E+1 Bone surf (gE÷1)3E+1 2E-1 Bone surf Bone surf (5E01) (3E-1)3E-11 5E-9 4E-9 4E-B 9E-9 7E-11 4E-8 1E-10 2E-13 2E-11 1E-11 1E-10 3E-11 3E-13 1E-10 SE-13 2E-7 6E-6 1E-5 4E-5 7E-6 5E-7 1E-4 ZE-6 6E-5 1E-4 4E-4 7E-5 SE-6 1E-3 CD 101 Mendelevium-257 W, all compounds 101 Mendelevium-258 W, all compounds Any single radionuclide not listed above with decay mode other than alpha emission or spontaneous fis-sion and with radioactive half-life less than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months --. .Submersion t-Any. single radionuclide not listed above with decay mode other than alpha emission or spontaneous fis-sion and with radioactive half-life greater than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .Any single radionuclide not listed above that decays by alpha emission or spontaneous fission, or any. mix-ture for which either the identity or the concentration of any radio-nuclide in the mixture is not known ....6E-7 6E-6 2E+2 1E-7 1E-9 2E-1 1E-lO 1E-12 1E-8 1E-7 4E-4 2E-13 1E-15 2E-9 2E-8 B-50 ODCM Rev. 25 App. B PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION App. B FOOTNOTES:

1"Submersion" means that values given are for submersion in a hemispherical semi-infinite cloud of airborne material.2These radionuclides have radiological half-lives of less than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The total effectivedose equivalent received during operations with these rsdionuclides might include a significant contribution from external expo-sure. The DAt values for all radionuclides, other than those designated Class "Submersion," are based upon the committed effective dose equivalent due to the intake of the radionuclide into the body and do NOT include poten-tially significant contributions to dose equivalent from external exposures.

The licensee may Tuh-stitute IE-7 pCi/mI for the listed DAC to account for the submersion dose prospectively, but should use individual monitoring devices or other radiation measuring instruments that measure external exposure to demonstrate compliance with the limits. (See § 20.1203.)3 For soluble mixtures of U[-238, U-234, and U-235 in air, chemical toxicity may be the limiting factor (see§ 20.1201(e)).

If the percent by weight (enrichment) of U-235 is not greater than 5, the concentration value for a 40-hour workweek is 0.2 milligrams 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 8E-3 (SA) pCi-hr/mi, where SA is the specific activity of the uranium inhaled. The specific activity for natural uranium is 6.77E-7 curies per gram U. The specific activity for other mixtures of U-23B, U-235, and U-234, if not known, shall be: SA = 3.6E-7 curies/gram U U-depleted 5A = [0.4 + 0.38 (enrichment)

+ 0.0034 (enrichment) 2] E-6 , enrichment

> 0.72 where enrichment is the percentage by weight of U-235, expressed as percent.NOTE: 1. If the identity of each radionuclide

'in a mixture is known but the concentration of one or more of the radionuclides in the mixture is not known, the DAC for the mixture shall be the most restrictive DAC of any radionuclide in the mixture.2. If the identity of each radionuclide in the mixture is not known, but it is known that certain radionuclides specified in this appendix are not present in the mixture, the inhalation ALI, DAC, and effluent and sewage concentrations for the mixture are the lowest values specified in this appendix for any radionuclide that is iD not-knownoto be absent from the mixture; or Table 1 Table 2 Tablea U. Occupational Values Effluent Releases to Concentrations Sewers Col. I Col. 2 Col. 3 Col. 1 Col. 2 Oral Monthly Ingestion Inhalation Average ALl AT OACR Air Water Concentration Radionuclide (pCi) (pCi) (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml)If it is known that Ac-227-D and Cm-250-W are not present 7E-4 3E-13. -If, in addition, it is known that Ac-227-W,Y, Th-229-WY, Th-230-W, Th-232-WY, Pa-231-W,Y, Np-237-W, Pu-239-W, Pu-240-W, Pu-242-W, Am-241-W, Am-242m-W, Am-243-W, Cm-245-W, Cm-246-W, Cm-247-W.Cm-248-W, Bk-247-W, Cf-249-W, and Cf-251W are not present 7E-3 3E-12 If, in addition, it is known that Sm-146-W, Sm-147-W, Gd-148-D,W, Gd-152-DW, Th-228-W,Y, Th-230-Y, U-232-Y, U-233-Y, U-234-Y, U-235-Y, U-236-Y' U-238-Y, Np-236-W, Pu-236-W,Y, Pu-238-W,Y, Pu-239-Y, Pu-240-Y, Pu-242-Y, Pu-244-WY, Cm-243-W, Cm-244-W, Cf-248-W, Cf-249-Y, Cf-250-W,Y, Cf-251-Y; Cf-252-W,Y, and Cf-254-W,Y are not present 7E-2 3E-11 If, in addition, it is known that Pb-210-0, Oi-210m-W, Po-210-D,W, Ra-223-W, Ra-225-W.Ra-226-W, Ac-225-D,W,Y, Th-227-W,Y, U-230-OW,Y, U-232-OW, Pu-241-W, Cm-240-W, Cm-242-W, Cf-248-Y, Es-254-W, Fm-257-W,.and Md-258-W are not present 7E-1 3E-10 B-51 ODCM Rev. 25 M App. B PART20 STANDARDS FOR PROTECTION AGAINST RADIATION ApO.B Table I Table 2 Table 3 Occupational Values Effluent Releases to.Concentrations Sewers Col. I Col. 2 Col. 3 Cal. 1 Cal. 2 Oral Monthly Ingestion Inhalation

.Average ALI. ALI DAC Air Water Concentration Radionuclide (pCi) (pCi) (pCi/ml) (pCi/ml) (pCi/nf) (pCi/ml)If, in addition, it is known that Si-32-Y, Ti-44-Y, Fe-60-0, Sr-90-Y, Zr-93-O, Cd-113m-O, Cd-l13-f, Inrhl5-O,W' La'138-f, tu-176-W, Hf178mn,W, flf-182-flW, Bi-21I.-O, Ra-224-W, Ra-228-W, Ac-226-0,W,Y, Pa-230-W,Y, tI-233-D,W.

U-234-f.W, U-235-f,W, U-236-O,W, U-238-D,W, Pu-241-Y, Bk-249-W, Cf-253-W,Y*

and ES-253-W are not present 7E-O 3E-9 If it is known that Ac-227-D,W,Y, Th-229-W,Y, Th-232-W,Y, Pa-231-W,Y, Cm-248-W, and Cm-250-W are not present -1E-14 If, in addition, it is known that Sm-146-W, Gd-148-D,W, Gd-152-0, Th-22B-W,Y, Th-230-W,Y, U-232-Y, U-233-Y, U-234-Y. U-235-Y, U-236-Y,-U-238-Y, U-Nat-Y, Np-236-W.

Np-237-W, Pu-236-W,Y, Pu-238-W,Y, Pu-239-W,Y, Pu-240-W,Y, Pu-242-WY, Pu-244-W,Y, Am-241-W, Am-242m-W, Am-243-W, Cm-243-W, Cm-244-W Cm-245-W, Cm-246-W, Cm-247-W, 8k-247-W, Cf-249-W,Y, Cf-250-W,Y, Cf-251-W,Y, Cf-252-W,Y, and Cf-254-W,Y are not present 1E-13 If, in addition, it is known that Sm-141-W, I" Gd-152-W, Pb-n1-fo, Bi-2l0m-W, Po-210-OW, cc Ra-223-W, Ra-225-W, Ra-226-W, Ac-225-fW,Y,Th-227-W,Y, U-23O-0,W,Y, U-232-O,W, U-Nat-W, t Pu-241-W, Cm-240-W, Cm-242-W, Cf-248-W,Y, Ul) Es-254-W, Fm-257-W, and Md-258-W are not present .I-E-12 If, in addition it is known that Fe-GO, Sr-90, Cd-l13m, Cd-113, In-15, 1-129, Cs-134, Sm-145, Sm-147, Gd-148, Gd-152, Hg-ig4 (organic), Bi-210m, Ra-223, Ra-224, Ra-225, Ac-225, Th-228, Th-230, U-233, U-234.U-235, U-236, U-238, U-Nat, Cm-242, Cf-248, Es-2G4, Fm-257, and Md-258 are not present 1E-6 iE-G 3. If a mixture of radionuclides consists of uranium and its daughters in ore dust (10 pm AMAD particle distribution assumed) prior to chemical separation of the uranium from the ore, the following values may be used for the DAC of the mixture: 6E-11 pCi of gross alpha activity from uranlum-238, uranium-234, thorium-230, and radium-226 per milliliter of air; 3E-11 pCI of natural uranium per milliliter of air; or 45 micrograms of natural uranium per cubic meter of-air.4. If the identity and concentration of each radlonuclide in a mixture are known, the limiting values should be derived as follows: determine, for each radionuclide in the mixture, the ratio between the concentration present in the mixture and the concentration otherwise established in Appendix 8 for the specific radionuclide when not in a mixture. The sum of such ratios for all of the radianuclides in the mixture-may not exceed "1" (i.e., "unity").Example: If radionuclides "A," "B," and "C" are present in concentrations CA, C%, and CC, and if the applicable ACs are DACA' DACE' , and B ACC. respectively, then the concentrations shall be limited so that the following relationship exists: CA C8 CC UA -C I- C -C B-52 ODCM Rev. 25 APPENDIX C EMS SOFTWARE DOCUMENTATION C-1 ODCM Rev. 28 APPENDIX C EMS SOFTWARE DOCUMENTATION TABLE OF CONTENTS CONTENTS Attachment I Effluent Management System Software Test Report for Seabrook Station, May 1994 PAGES/REV.

C-3 Cover ii 1-11 Attachment 2 Attachment 3 Attachment 4 Resolutions of EMS Software Test Report Discrepancies Software Requirements Specification for North Atlantic Energy Service Corporation, Seabrook Station, Effluent Management Systems, Revision 04, FP 75486 Technical Reference Manual, Effluent Management System NAESCO Seabrook Station, July 1994, FP 75486 C-4 1-2 C-5 1-35 C-6 36 37 38 thru 93 R28 R28 R22 C-2 ODCM Rev. 28 APPENDIX C: EMS SOFTWARE DOCUMENTATION ATTACHM4ENT

.1: EFFLUENT MANAGEMENT SYSTEM SOFTWARE TEST REPORT FOR SEABROOK STATION, MAY 1994 0 c-3 ODCM Rev. 16 S EFFLUENT MANAGEMENT SYSTEM: SOFTWARE TEST REPORT FOR SEABROOK STATION MAY 1994 Prepared by Reviewed by, Approved-by D* ate.flat.Date Yankee Atomic Electric Company Nuclear Services Division 580 Main Street Bolton, Massachusetts 01740 I Table of Contents

1.0 INTRODUCTION

.1.. ................... ..1 1.1 Background

..1.... ......... ... .1.2 Acceptance Criteria ..................................

1 2.0

SUMMARY

OF FINDINGS.............

.......... 2 2.1 EMS 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

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9 References

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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 ODCM assumptions and methods. All executions of the EKS 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 Mi) information contained in the EMS data files is consistent with the ODCH, (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 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 ODCM. 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 ODCH 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-i 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 Pp 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 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 Caseous 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 ODCH methods and the EMS program is excellent, indicating that the EMS calculation method is consistent with the ODCM.There is also excellent agreement between inhalation doses from the EMS program and ODCK 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 ODCX 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 ODCM methods. This difference increases to about a factor of 4 when the current values for Fp and SF assumed in the EMS program and ODCK 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, Co60 and Cs137 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, Co60, 1131, and*5 H3, the missing dose rate information was unexpected.

It is noted that organ dose rate information was provided on Page 4 of the E!1S printout during a demonstration of the EMS program prior to testing.0 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 ODCM 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 I1, 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 ODCH 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 0 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 NlUREG-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 radioiodine in gaseous effluent.4. The radiation monitor set point determination method for liquid releases produces a set point value that is consistent the ODCM 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, t-*, is used consistently to the ODCH.80 4.0 SU12ARY 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 expected to used in EMS discharges.

follow ODCM assumptions program. for Method I or Method II.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 Pp 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 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 ODCM methods, fraction for elemental iodine that is consistent with ODCM.S 9 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, 1133, H3, and particulates.

10 References.-

1. Yankee Atomic Electric Company, Effluent Monitorinp 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 Compan= Effluent Management System Technical Reference Manual (07-0545), January 1993.4. NRC Regulatory Guide 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purposes of Evaluating Compliance with 1OCFR Part 50. Appendix I, Revision I, October 1977.5. NRC NUREG-0133, Preparation of RadioloKical Effluent Technical Specifications for Nuclear Power Plants, October 1978.S 11 APPENDIX C: EMS SOFTWARE DOCUMENTATION 0 ATTACHMENT 2: RESOLUTIONS OF EMS SOFTWARE TEST REPORT DISCREPANCIES 0 C-4 ODCM Rev. 16 Attachment 2@ 2. Resolution of EMS Software Test Revort DiscreRancies 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.Discrepancv:

Mixing ratio for shoreline activity used in EMS Program not equal to the value used in the ODCM Method I (Mp = 1.0).I 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.EMS dose factors based on Fp (fraction of year animals are on pasture) value which is not consistent with ODCM.I 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 correspond to the growing season where the pasture fraction is assumed to be 1.0. The EMS software assumes an Fp value of 1.0 for animal grazing (meat and 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 dose limits. No changes to the EMS software are necessary.

Discrepancy:

I Shielding factors (SF) applied uniformly to dose rates and doses in the EMS program.0 I I Attachment 2 2. Resolution of EMS Software Test Report Discrepancies (Continued)

I 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 Discrepancy:

Incorrect receptor location identified on EMS printout for ground level release point.I Resolution:

Incorrect name is identified on report with no impact on dose or dose rate calculations which were verified to be correct.I Discrepancy:

Assumed fraction of elemental iodine used in EMS program differs from ODCM 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:

I Missing organ dose rate information on EMS printout for effluent discharges containing 1-131, 1-133, H-3, and particulates.

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 0 C-5 ODCM Rev. 16 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:

a,.- ./Approved: Engineering (CVNDS)A pproved: Approved: Pro~jq Arinager (Uabrook Station)Date: /Date: -Date:.Date: _O/_7__/Y Software Requirements Specification RS-8448-04 Revision History Initials DJH DJH DJH DJH DJH Revision 00 01 02 03 04 Date 2/26/93 3/22/93 4/30/93 8/3193 9114/93 Description initial version Updated incorrect dose equation Updated to include all dose and dose rate equations Updated based on modifications to software and customer's requested modification to the use of the default nuclide for gaseous permit processing.

Updated based on customers request to remove modification to the default nuclide for gaseous permit processing.

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Software Requirements Specification RS-8448-04

1. Scope 1 2. Applicable Documents 1 3. Interfaces 1 3.1 Hardware 1 3.2 Software 3.3 Human .2 3.4 Packaging 2 4. Definitions 2 5. Principal Changes from Existing Packages 2 6. EMS Functionality 4 6.1 Database Maintenance Transactions 4 6.2 Editing Values through INGRES OBF 7 6.3 Liquid 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 Liquid 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 SemI-Annual Reporting 28 6.11.1 User Interface and Functionality 28 6.11.2 EMS Trend Plots 30 6.12 End-of-the-Year Data Archiving 30 6.12.1 User Interface and Functionardy 30 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 Manuar (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 customized 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 "Seabrook Station Offslte 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 VAX/FORTRAN or VAX-DCL. Utility programs provided by INGRES that are installed on the hardware configuration may be used if applicable.

S-9~c~H S Software Requirements Specification RS-4M48-04 S 3.3 Human The user may be expected to have received operator training from the system manager, Canberra/NDS, 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 X/Q values.

Software Requirements Specification RS-8448-04 5.1.3 Modification of noble gas dose rate and dose calculation methods to multiply X/Q and D/Q 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 Liquid 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 RS-8448-04

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 nuclide 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 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: " 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 Setpolnt 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: " DEFNUC: 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 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: " DEFCONC: 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.

~47~7 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:

DEFTYPE: For a liquid or gaseous release, this parameter will contain the default nuclide type that will be used in setpoint calculations for low gamma concentration 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)1, 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 Setpolnts 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 "Y"', 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-DS], 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 Pump" or "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 list of MET DATA TABLES. These additional menu items-are as follows: XIQ -Noble Gases (Gamma)"a" Factor -D/O-Part/Iodines"a" Factor -Noble Gases"a" Factor -X1Q-Partflodines"a" Factor -Gamma Noble Gases Software Requirements Specification RS-8448-04 6.1.10 6.1.11 On the Meteorological Data transaction

[EM-DM-ME (Form 1)], the following menu items will be used to store short-term (1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ) D/Q and X/Q values.D/O -Partics/Radioiodines X/Q-- Partics/Radioiodines X/Q -Decayed Noble Gases X/Q -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.On the Meteorological Data transaction

[EM-DM-ME (Form 1)], the X/Q, D/0, and "a" 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/0, D/Q, and "a" 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. -0 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)dvdate (dilution volume date)tvol (total volume)0 [not used]The first day of the month to which the volume applies (time not required).

Dilution volume for the month, In user units.0 [not used]aflow (average flowrate)0 twe- ýo Software Requirements Specification RS-8448-04 6.3 Liquid Pre-Release Processing 6.3.1 User Interface and Functionality Liquid 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 Permit Definition Screen (Screen 3.04): Upon entering the permit definition screen, if the DILOOKUP parameter is set to "Y" for the release point associated with the current permit being processed, the Dilution Flow Rate parameter will default to zero.If a user uses the "Tab" or "Return" 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 Uquid Permit Definition Screen (Screen 3.04): When a "Fill" (F14) or a "Save" (F1O) 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.

IT Software Requirements Specification RS-8448-04 6.3.1.3 Prior to entering 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 DEF..NUC parameter.

The default concentration for this nuclide will be attained from the DEFCONC parameter.

The default type for this nuclide should be attained from the DEFTYPE 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 "Yq 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 Operator's 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 *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.6.3.2.2 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 Setpoint will appear below the Release Point and Discharge Point Setpoint values in the Radiation Monitor(s) portion of the report.

Software Requirements Specification RS-W48-04 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 Liquid 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.

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 Uquid Permit Definition Screen (Screen 3.13): If the DILOOKUP parameter is set to Or for the release point and a user uses the "Tab" or *Return" 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 Uquld Permit Definition Screen (Screen 3.13): When a "Fill* (F14) or a "Save" (F10) 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.)

Software Requirements Specification RS-4W48-04 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 appear following the Nuclide Concentration Screen (Screen 3.15). The monitor response values entered should include the monitor background values.6.4.2 Associated Reports.Liquid 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 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.6.4.3 Underlying Calculations The calculations performed by the EMS software for Liquid 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.

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.respectively.

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 fquid permits through the EMS software shall be described section 3 or the EMS operators Manual (Reference 2.1.1).

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 SCALNUC parameter Is set to "r, 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 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.6.7.1.3 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 setpolnt 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 DEFCONC parameter.-

The default type for the default nuclide should be attained from the DEFTYPE 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 "Y" 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): 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._14-Software Requirements Specification RS-8448-04 6.7.2 Associated Reports 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 "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.6.7.2.3 On the Pre-Release Permit Report (4.01), the Noble Gas Alert Alarm Setpoint will appear below the Max Monitor Setpoint values. The Alert Alarm Setpoint will be calculated by multiplying the noble-gas monitor setpoint value by a multiplier specified with the ALRT SET parameter mentioned above.6.7.2.4 On the Gaseous Special Report (4.02), the Noble Gas Alert Alarm Setpoint 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.6.7.2.5 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:

Software Requirements Specification RS-4A48-04 S 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 < 80 meters): Dt = shf -X/Og .8760-a -FO -X (Ki

QRiv)where Dt = the total body dose rate due to gamma emissions by noble gas releases from vent v (mrem/yr)sOf shielding factor (dimensionless)

QRiv = release rate of noble gas radionuclides, i, In gaseous effluents from vent or stack v ( pCi/sec).Fo = occupancy factor defined for the receptor at the given location (dimensionless)

Ki = total body dose factor due to gamma emissions for noble gas radionuclide I (mrem/yr per pCi/m 3)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/rn 3)8 7 6 0-a= adjustment factor used to convert the 1-hour X/Q value to an average 1 year X/Q value (dimensionless) where 8760 = number of hours in a year-,a = a" factor for gamma noble gas X1Q For Noble Gas Total Body Dose (for vents or stacks < 80 meters): sht- Fo

Z (Ki -QRiv) -XIOg .t-a Db =(5.256 -10 5 / dur)where Dtb .= total body dose from gaseous effluents (mrem)5.256

105 = number of minutes in a year dur = duration of the release (minutes)

Software Requirements Specification RS-8448-04 0 t-a = adjustment factor to convert the 1-hour X/Q value to the short term X/O 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 -[(Li .X/Q -8760"b) + (1.11M, -X/Qg .8760"a)]where Ds = skin dose rate from gaseous effluents (mrem/yr)XtQ = highest value of the noble gas 1-hour X/Q for vent or stack v at the site boundary (sec/rm 3)Mi = air dose factor due to gamma emissions for noble gas radionuclide I (mrad/yr per pC/rm 3)1.11 = conversion factor from mrad to mrem Ii = skin dose factor due to beta emissions for noble gas radionuclide i (mremtyr per pC/rm 3)b = "a factor for noble gas X/Q" For Noble Gas Skin Dose (for vents or stacks < 80 meters): shf. Fo Z QRiv° [(i* X/Q -t-b) + (1.11 Mi X/Og- t-a)l Dsk =(5.256 -105 / dur)where Dsk = total skin dose from gaseous effluents (mrem)

Software Requirements Specification RS-8448-04

For Noble Gas Air Dose due to gamma radiation (for vents or stacks < 80 meters): Dy = (3.17 8) -X/Qg *t a -Fo

I Mi a Qiv where D V total gamma air dose from gaseous effluents (mrad)3.17 8 = inverse of number of seconds in a year Qiv- release of noble gas radionuclides, i, in gaseous effluents from vent or stack v (pCQ)v = 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): SDp = (3.17 8)- yjQ _ t-b

Fo -Z: Ni -Qiv where Dp = total beta air dose from gaseous effluents (mrad)Ni = air dose factor due to beta emissions for noble gas radionuclide i (mrad/yr per piCVm 3)For Critical Organ Dose Rate-Inhalation Pathway and all Pathways for H-3, C-14 (for vents or stacks < 80 meters): DR-ra X/Qr -8760-C -X PIpra OQRiv 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)PipT = dose factor for each radionuclide 1, pathway p, organ T, and age group a (mremryr per pCi/mi)

Software Requirements Specification RS-8448-04 X/Qr highest value of the radioiodine/particulate 1 -hour XIQ for vent or stack v at the site boundary (sec/m 3)c "a" factor for Radiolodine/Particulate X/0 Note: It is assumed Pipra will not contain long term XIQ or D/Q values.For Critical Organ Dose Rate--Ground and Food Pathways (for vents or stacks < 80 meters): DR = DIQ -8760-d .Z Rip.a

QRiv where D3/ = highest value of the 1-hour deposition factor at the distance of the site boundary (1/m 2)d = "a" factor for D/Q RipTa = dose factor for each radionucllde i, pathway p, organ T, and age group a (m 2 .mremn/yr per pCVsec)Note: It is assumed RipTa will not contain long term X/Q or D/Q values.V For Critical Organ Dose-Inhalation Pathway and all Pathways for H-3, C-1 4 (for vents or stacks < 80 meters): D7a 7 (3.17 in0" 8) -X/Qr tc a Fo T' PipTa"lv where DTa = dose for age group a and organ T from iodines and particulates with half lives greater than 8 days in gaseous effluents (mrem)Note: It is assumed PipTa will not contain long term X/Q or D/Q values.-19-I Software Requirements Specification RS-8448-04 For Critical Organ Dose-Ground and Food Pathways (for vents or stacks < 80 meters): D~a = (3.17"10-8)-D/Q t-d-Fo.-ERipTa Qiv Note: It is assumed RipTa will not contain long term X/O or D/Q values.6.7.3.2 On the Nuclide Concentration Screen (Screen 4.06), nuclide concentrations will be "scaled" if the SCAL._NUC parameter is set properly for a Release Point. This "scaling" is described as follows: Cinew = (t / s), C!where Cinerw = concentration (after "scaling")

of nucrldei s = sum of all nuclide concentrations on the Nuclide Concentration Screen.t = total nuclide concentration entered by the user Ci = concentration (before "scaling")

of nuclidei 0 Software Requirements Specification RS-4M48-04 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 *Processe (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 can 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 VAXGSP (F172) 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 "-f 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:

Software Requirements Specification RS-8448-04 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 SCAL.NUC 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 reporL 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 clarifications:

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/QWg 8 7 6 0-a -Fo 0 X (Ki -QRiv)where Dt = the total body dose rate due to gamma emissions by noble gas releases from vent v (mrem/yr)shf = shielding factor (dimensionless)

QRiv = release rate of noble gas radionuclides, i, in gaseous effluents from vent or stack v ( /Ci/sec).Fo = occupancy factor defined for the receptor at the given location (dimensionless)

Ki = total body dose factor due to gamma emissions for noble gas radionuclide I (mrem/yr per pCi/m 3)

Software Requirements Specification RS-8448-04 XJOg = 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/Q value (dimensionless) where 8760 = number of hours in a year a = "a' factor for gamma noble gas X/Q For Noble Gas Total Body Dose (for vents or stacks < 80 meters): sht -Fo -Z (KI .QRIv), X/Qg. t-a (5.256 ° 105 / 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)tra = 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 I QR (Li- X/Q 8760-b) + (1.11MI .X/Q1 9 8760-a)]where Ds = skin dose rate from gaseous effluents (mrem/yr)X/Q = 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-8448-04 Mi = air dose factor due to gamma emissions for noble gas radionuclide i (mrad/yr per pCi/m 3)1.11 = conversion factor from mrad to mrem Li = skin dose factor due to beta emissions for noble gas radionuclide i (mremlyr per p Cl/m 3)b = "a" factor for noble gas X/Q For Noble Gas Skin Dose (for vents or stacks < 80 meters): shf -Fo 0 -QRiv* [(Li .X/Q t.rb) + (1.11,MI XMgC t'a)]Dk=(5.256 .105 / dur)where Dsk = total skin dose from gaseous effluents (torem)For Noble Gas Air Dose due to gamma radiation (for vents or stacks < 80 meters):-Dy = (3.17 ° 10-8) ° X/Qg

t-a Fo. -IMi

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 from vent or stack v (pCi)Qiv = QRiV -dur- 60 where 60= number of seconds in a minute Software Requirements Specification RS-8448-04" For Noble Gas Air Dose due to beta radiation (for vents or stacks <80 meters): Dp = (3.17*10"8)*XiQ.t-b.Fo*Z Ni*iv where DP = total beta air dose from gaseous effluents (mrad)N 1 = air dose factor due to beta emissions for noble gas radionuclide i (mrad/yr per pCI/m 3)" For Critical Organ Dose Rate-Inhalation Pathway and all Pathways for H-3, C-14 (for vents or stacks < 80 meters): DRTa War "8760-c Z PipTa "QRiv 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)PipTa= dose factor for each radionuclide i, pathway p, organ T, and age group a (mrem/yr per pCi/rm)Wa~r = 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 Radloiodine/Particulate XWO Note: It is assumed Pipra will not contain long term X10 or DIQ values.For Critical Organ Dose Rate--Ground and Food Pathways (for vents or stacks < 80 meters): DRTa D/Q -8760-d XT Ripra -QRiv where D/I = highest value of the 1-hour de 2 position factor at the distance of the site boundary (1/rn d = "a" factor for DIQ Software Requirements Specification RS-8448-04 Ripza = dose factor for each radlonuclide i, pathway p, organ r, and age group a (m 2 .mrem/yr per pCVsec)Note: It is assumed RipTa will not contain long term XIQ or D/Q values.For Critical Organ Dose-inhalation Pathway and all Pathways for H-3, C-14 (for vents or stacks < 80 meters): DTa = (3.17 8) -X/Qr

t-c -F0

1 PipTa

Qiv where DTa dose for age group a and organ T from lodines and particulates with half lives greater than 8 days in gaseous effluents (mreom)Note: It is assumed PipTa will not contain long term X/Q or DI0 values.For Critical Organ Dose-Ground and Food Pathways (for vents or stacks < 80 meters): DTa = (3.17.10-8 y°D/Q-t4-Fo.XRipTa.

Oiv Note: It Is assumed RipTa will not contain long term X/Q or D/Q 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 / s)

Ci where Cinew = concentration (after "scaling")

of nuclide, s = sum of all nuclide concentrations on the Nuclide Concentration Screen.t = total nuclide concentration entered by the user Ci = concentration (before "scaling")

of nuclidei Software Requirements Specification RS-8448-04 6.9 Gaseous Permit Editing 6.9.1 User Interface and Functionality 0'Functionality for editing gaseous permits through the EMS software shall be described in section 4 of the EMS Operator's 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 original 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)._fýrý3 O Software Requirements Specification RS-848-04 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 100 Dag /Lag 100. DabI/ Q-ab 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]QLab = 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 (over T) ofýj 100-(E DIT)/OLrp 3 Software Requirements Specification RS-4W48-04 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 QLrp = 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 quarter q in the report, calculate the reportable dilution volume (DVrq, 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: DVrq = 28.31685 -sdlvolf -X DVrm The values DVrm are from the column tvol of the QDVOL table. The value DVrq is Included In the report on line F, and Is used in the calculations below.' "sd Ivolf' should be the user unit conversion factor to convert from user units to ft 3.28.31685 is a unit conversion factor from ft 3 to liers.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 Clq = Z [Actiq / (1000 DVrq)]where Actiq = total activity of nuclide I released during the portion of the quarter q that is within the period (p Ci)DVrq = reportable dilution flow for the portion of quarter q that is within the report period (liters), as calculated above.Compute each value on line A.3 and B.3 of the report by taking the greater of j 100 -Dit / QLIt 100 -Dio I OL2o Software Requirements Specification RS-8448-04 where Dit = the liquid total body dose in the applicable quarter at the site boundary receptor (mrem)D1o = the liquid maximum organ dose in the applicable quarter at the site boundary (mrem)QLft = the quarterly limit on Dlt (mrem) [usually 1.5]QLIo = the quarterly limit on Dlo (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 line C.3 of the report as follows: pq = 100 -Cq Ldg where Cq = sum of noble gas concentrations Pq = Percentage applicable to a given quarter for dissolved and entrained gases Ldg = Uquid dissolved gas limit (pCL/mI) [usually 2.0E-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 section 6 of the EMS Operator's Manual (Reference 2.1.1) with no revisions.

Documentation Review Report Document Reviewed IS2A' S'!A~,?~ ~f1A'7hAJ

~)tS~ 2J4'4'~-o4' Does the document meet-the requirements?(a or No Is the document approved?r or No Ift not please state the exceptions:

0 AK6f Date: f/O-T Signature:

-)I Ij"O-5792 Documentation Review Report~: (x ~ I~ L)m a Document Reviewed Does the document meet the requirements?3e'or No Is the document approved?d">

or No if not please state the exceptions:

Signature:

---4-./Date: ___________

WO-67.2 APPENDIX C: EMS SOFTWARE DOCUMENTATION ATTACHMENT 4: TECHNICAL REFERENCE MANUAL, EFFLUENT MANAGEMENT SYSTEM NAESCO SEABROOK STATION, JULY 1994, FP 75486 C-6 ODCM Rev. 28 Canberra Industries, Inc.800 Research Parkway Meriden, CT 06450 July 1994 NAESCO Seabrook Station EMS Technical Reference Manual 07-0625 Copyright 1994, Canberra Industries, Inc.Printed in U.S.A.C C~\P~~3(

THIS PAGE IS INTENTIONALLY BLANK Page 37 ODCM Rev. 28 A A,.I -1 TABLE OF CONTENTS Page CHAPTER 1 INTRODUCTION

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1-1 1.1 SETPOINT CALCULATIONS

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1.2 RELEASE PROCESSING

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1.3 COM POSITE NUCLIDES ..........................................................................

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1-2 1-3 1-3 1- I CHAPTER 2 LIQUID RELEASE CALCULATIONS

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2-1*I*-2.1 LIQUID PRE-RELEASE PERMIT .............................................................................

2.2 10CFR20 COMPLIANCE

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Dissolved and Entrained Gases ....................................

2.3 MAXIMUM WASTE FLOW .......................................................................................

2.4 MINIMUM DILUTION FLOW RATE ............................................................................

2.5 SETPOINT CALCULATIONS

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Recommended Setpolnt ...................................................................................

Setpoint in pCVmI .............................................................................................

Recommended Setpoint in User Units (e.g. cpm) ........................

Setpolnt for Discharge Point .................................

Setpoints in pCi/sec ...........................................................................................

2.6 DOSE CALCULATIONS FOR LIQUID RELEASES ......................................

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2.7 31 DAY PROJECTED DOSE CALCULATIONS.................................................

2.8 POST-RELEASE PROCESSING

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2-1 2-1 2-4 2-4-2-5 2-5 2-8 2-8 2-10 2-11 2-13 2-13 2-17 2-18 CHAPTER 3 GASEOUS RELEASE CALCULATIONS

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3-1 3.1 GAS PRE-RELEASE PERMIT ....................................................................................

3.2 RADIONUCLIDE ACTIVITIES AND COMPOSITE VALUES .....................................

Activity Released .........................................

3.3 1OCFR20 COM PLIANCE ............................................................................................

3.4 SETPOINT DETERMINATION

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f ..........................................................

Noble Gases ..... ... '..............

... .....................-

"*.........

......Radioiodlnes and Particulates

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3.4a SETPO INTS ..............................................................................................................

3.4b REPORTED SETPOINTS

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Setpoints in p Ci/sec ..........................................................................................

3,5 MAXIMUM WASTE FLOW ............................................

3.6 DOSE RATE AND CUMULATIVE DOSE CALCULATIONS

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Noble Gas Dose and Dose Rate Calculations

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3-1 3-1 3-2 3-2 3-5 3-5 3-6 3-6 3-10 3-11 3-12 3-12 3-12 0@I P3 3 L ODCAk Yjb 2-2 Oroan Dose Calculations

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3-15 3.7 RESOLVING DOUBLE-COUNTING OF DOSE AND ACTIVITY ..............

3-16 3.8 31 DAY PROJECTED DOSE CALCULATIONS

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3-17:3.9 GAS POST-RELEASE PROCESSING

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3-18 CHAPTER 4 LIQUID DOSE FACTOR EQUATIONS

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4-1 4.1 POTABLE W ATER ...................................................

t .................................................

4-2.4.2 AQUATIC FOODS PATHWAYS ................................................................................

4-2 4.3 SHORELINE RECREATION PATHWAY ....................................................................

4-3 4.4 IRRIGATED VEGETABLE PATHWAY .................................

4 4.5 REDUCTION TO NUREG-0133 EQUATIONS

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4-5 CHAPTER 5 GAS DOSE FACTOR CALCULATIONS

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5-1 5.1 INHALATION PATHW AY ............................................................................................

5-1 5.2 GROUND PLANE PATHWAY ...............................

5-2 5.3 MILK PATHW AY ......................................................................................................

5-2 Carbon-14 in Milk .........................................

5-4 Tritium in Milk .... ....................

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5-4 5.4 MEAT PATHW AY ......................................................................................................

5-5 Carbon-14 in Meat ...........................

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5.................................................

5-5 Tritium in M eat ..................................................................................................

5-6 5.5 VEGETABLE PATHWAY ..........

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5-6 Carbon-14 In Vegetables

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5-7 Tritium In Vegetables

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5-7 5.6 REDUCTION TO NUREG-0133 EQUATIONS

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5-7 APPENDIX A REFERENCES

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A-1 igo 0 .0 9 CHAPTER 1 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 equations reduce to those described in NLTREG-0133 by proper selection 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 IOCFR20 (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 the effluent release portions of the technical specifications 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 lOCFR20, 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 10CFR20. The old IOCFR20 approach for gaseous effluents has been accepted by the NRC for use under the new 10CFR20.I*W 1-1 Q-CUi4( razz o For gaseous effluents, the maximum dose to any member of the public will be less than the limits given in l0CFR50, Appendix I.o The maximum dose to any member of the public will not exceed the limits given in 40CFR290.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 gas.eous 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 IOCFR20 compliance is met. For the'old 10CFR20 *compliance, liquid' calculati'ons use the maximum perinissible concentrations from IOCFR20 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 1OCFR20 requirements, the effluent concentration limits are used for liquid setpoint calculations.

For gaseous setpoint calculations under the new 10CER20, 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/secl.

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.

0 1-2 EMS has a "nuclide specific" option. In this option only the nuclides listed in the monitor slope table are used in the setpoint calculations.

I&2.2. 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 calculating the doses and setpoints, comparing to the 10CFR20 limits, and comparing to the IOCFR50 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 aonalogou's 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.

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 K-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 i For liquid releases, the composite nuclides are generally H-3, Fe-55, Sr-80, Sr-90, and gross alpha. For gaseous releases, Fe-55 is generally not included.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 NAESCO Seabrook Station EMS Operator's Manual 07-0589.code, see the So 1-4 P') q ODQ- &CeQ zZQ 00 CHAPTER 2 LIQUID RELEASE CALCULATIONS 2.1 LIQUID PRE-RELEASE PERMIT A liquid pre-release permit is generated with a program that uses the nucli'de activities to determine the radiation monitor setpoint (for 1OCFR20 compliance) and the potential doses for 10CFR50.compliance.

leo Continuous releases are treated similarly.

2.2 1OCFR20 COMPLIANCE IOCFR20 compliance calculations are broken down into two paths. The first path calculates compliance 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.20CFR20 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 IOCFR20 NEW lOCFR20 S = Z% C./ECL5 10 1 1L i S Z. C./HPC. E 1 1 1 1 ORI for concentrations Ci released from the site. MPCi is the maximum permissible concentration from the old lOCFR20, Appendix B, Table Ii, Column 2, for nuclide i and ECLi is the effluent concentration limit from the new 10CFR20, Appendix B, Table 2, Column 2, for nuclide i.2-1 tjf If the summation is greater than the limit, then dilution is required.

The required dilution factor is: If the IOCFR20 option is OLD: C.MPC.req f R max where Dreq = Total required dilution factor Ci = Concentration of nuclide .i in pCi/mL MPCi = Maximum permissible concentration of nuclide i in pCifrnL f IRelease point setpoint safety factor (usually equal to.0.5) from the release point definition.

Rmax A The maximum .PC ratio from the release point setpoint definition.

If the IOCFR20 option is NEW: 0@Ci ECL.i=g 1 req, g f 0 R max ci Z--ECL.D i=ng i reqng f m a max Dreq

req, g + Dreq, ng where Dreq,g =Dreq, ng =ECL -Required dilution factor for gamma-emitters Required dilution factor for non-gamma-emitters Effluent concentration limit of nuclide i in PCi/mL 2-2 PJ0 00ý ye_ý7_2__

and the sums extend over gamma-emitters (g) and non-gamima!-ernitters (ng), respectively.

Any nuclides with MPCi i 0 are excluded from the sum.Any nuclides with ECLi i 0 are 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.F avail =F (ff/100) (1 -E C./XXX.)aal ant f1 1 where Ci = 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-The anticipated dilution factor is then Dant F waste+ falloc Favail /Fwaste where Fwaste waste flow anticipated for this release Fa.i= available dilution flow avail falloc= fraction of available dilution stream flow allocated to this release 00 OWL~k R'V _ Z 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/nmL under OLD.10CFR20, or I 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 HPC 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 WELOW 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 R1wmax and Rcwmax where Wmax =Maimum permissible waste flow rate for this release Rwmax -Release point maximum waste flow rate, as'set in the release point definition If the SET OPT option.= NONE: Rcwmax 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 Dreq -1.0 If the SET OPT option -CALC Favail

falloc + Fwaste Rcwmax. -Dreq 2-P h 7 0 0 If the SETOPT option = NWAS Favail

falloc Rcwmax Dreq 2.4 MINIMUM DILUTION FLOW RATE If Dreq > I, the minimum dilution flow rate is determined as follows: If the SETOPT option is NWAS: Fwaste a Dreq 0.-LJ* 4 ' -A 4 -- =~C, falloc * (ff 1 00) *[1 -z xxx.where XXXi is MPCi under OLD 10CFR20, and is ECLi under NEW 1OCFR2o.If the SET OPT option is other than "NWAS: Fwaste (Dreq 1.0)falloc * (ff / 100) 1 x if o

i Otherwise:

rmin dflow -0.0 2.5 SETPOINT CALCULATIONS Setpoints are calculated for individual release pointsr and for the discharge point that may combine several release points.A setpoint adjustment factor, Sadj is determined from the value of Dreq" 010'2-5 1) C If D > I or the dilution factor option is N, and the setpoint Dreq ecuation 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 NODIL, then no credit is taken for dilution, and the setpoint adjustment factor is: Sadj = I/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 SETPEQN is set to DILUT, and Fwaste > 0, then: Fant + Fwaste Sadj=Fwaste If the SETP__EQN is set to STD, and the SET_OPT option is set to NWAS, and Fwaste > O then: falloc 0 Favaii Sadj Fwaste

Dreq Otherwise, if the SETPEQN option is set to STD, and the SETOPT option is set to other than NWAS, and Fwaste > 0 , then: (falloc *Favail) + Fwaste Sadj =Fwaste

Dreq Otherwise, if the SETP EQN option is set to LOWACT, and the SETOPT option is set to NWAS, and Fwaste > 0, then: (falloc

Favail)E~waste -Dreq, ng Fwaste Sadj =Dreq'g 2-6 49 ce-fv\ {

Otherwise, if the SETP_EQN option is set to LOWACT, and the SETOPT option is set to other than NWAS, and Fwaste > 0, then: (falloc Favail) + Fwaste__________________-

Dreq~ng Fwaste Sadj Dreqvg Otherwise, Sadj is unchanged.

The setpoint adjustment factbr is further tested against, a limiting value (Sadj lim which is set using the.Release Point transaction in Database MAntenance).

if Sadj > Sadj,lim'then Sadj = Sadj,lim All of this leads to the maximum setpoint value, Smaxt based on the ganmma-emitting radionuclide mix: S max (i/ml) Sadj I Ci 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: Smax (cpm) =S adj(RMon -B) + B where B = monitor background (cpm)Rmon = monitor response (cpm)= offset + slope 0 Z C. + quad * (X Ci)2 + B where offset, slope, and quad are the coefficients

'in a quadratic fit to the monitor response to nuclide activity.@5 2-7 so EMS provides an option to calculate nuclide specific responses zo 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, Z (offseti+

slope 1-Ci + quad. * (Ci) 2 + B 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 i'n 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 .Ci/m.Note: In this version of the software, the reported setpoint is the user units setpoint.

The setpoint calculations using the original concentrations

(#Ci/ml) 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 puCi/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. Cs 1 3 7 equivalent response)2-4 A candidate setpoint is calculated based on the expected response: Sexp= ftol z Ci where ftol= 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 SeXp < Smax and if Sexp< Sdef and Sdef _. SMW%then use Sdef Case 1 Otherwise use Sexp. Case 2,5 If Sexp > Smax use Smax. Case 3 If Sm~ = 0, use Sdef Case 4 Case 4 occurs if no activity is detectable in the sample (Sadj 0).*I*-, 06" 2-9 P3 5z CO&GU22 Case I Case 2 Case 3 Case 4 Case 5 V Sexp ---- Sdef-.--Smax---- Smax-... Smax-... Sma -8exp Sexp Sdef-... Sdef---- Sdef ... Sdef-...Sex Use Sdef Use Use SMa, Use Sdef Use Sexp 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) = f (Rn- B) + f B exp tol mon, Btcal where fBtol background tolerance factor (set using QBF on the releasept table)If the default setpoint value includes background:

Brp =0 2-10 ObcGWý&- QC4 Z 2 0.If the default setpoint value does not include bac)cground:

Brp B where B is the monitor background count rate and Brp is used below.If Sexp (cpm) < Smax (cpm)and if Sexp (cpm) < Sdef (cpm) + Brp and Sdef (cpm) + Brp s Sma (cpxm)then use Sdef (cpm) + Brp Case 1 Otherwise, use Sexp (cpm) Case 2, 5 If Sexp (cpm) > Smax (cpm)00 use SmX (cpm) Case 3 If Smax (cpm) = 0, use Sdef + Brp Case 4.NOTE: Smax is due to concentrat-ion only (i.e., excludes backgroond) for Case 4 Setpoint for Discharqe Point For the discharge point, the total MPC/ECL fraction is: E Ci/VT Cio Y Fa + ( z C. /3 C.)/ F F + F OR Z Ci/ECLi)

F + ( Z C./ECL.)

F 0 1 3 .F +F 2-1l P3SL{r~i L where 0 (Z Ci/MPCi )o=total MPC fraction for existing concurrent releases for this discharge point excluding this additional release.Z Ci/MPCi = total MPC fraction for the new release (Z Ci/ECLi )o 0 total ECL fraction for existing concurrent releases for this discharge point. excluding this additional release.X Ci/ECLi = total ECL fraction for the new release=o 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: 4p C. [F1/(F + Fo R E(offset.+

slope.. C. + qud 2C+4pmon ad (C 1) dpmon where Cd = the discharge point isotope concentration from this release point Rdpmon = the discharge monitor response in user units Rdpmono= the discharge monitor response before the current release is added including the background 2-12 otuvA. reV2-2 For non-isotope specific response: Rdpmon = offset + slope .CdP + quad -(CdP)2 + Rdpmono where Cdp = I Ci] [F/(F+Fo)]

Setpoints in MCi/sec Setpoints in units of pCi/sec can be obtained by setting the UNITS parameter for the monitor to "pCi/s" or "pCi/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: D =A. I At C F inr iTr s is sr where: The sum extends over all time periods.Di~r= 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@ 0 with the Dose Factor Processing (DFP) option. The 2-13 0bCW~k.?4z equations used are presented in Chapter 4 of this manual.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. from any liquid release, in PCi/mI.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.6p*P:)i If the STREAM-FLO option in the OPTIONS Table 'is set to Y, then Denom = Fstrm (Uf /60) (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/G0) (I/Rimx)Else if denom-typ is 2, (dilution flJow includes waste flow)then Denom =FdiI ' (Uf/60)- .{I/Rnd.)else (denomtyp is not 1 or 2)if the QV OPT option in the OPTIONS table is set to ON, (dilution flow is from the QDVOL table) then Denon = (tw + Fg1oi1) (Uf/60) *(/RmiL)else (the normal standard calculation)

Denom = (Fw + Fdil) * (Uf/60) I end of if on QVOPT 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 of if denom is greater than 0.0 1* 0 Denom Denom / (Uf/60)2-15 Y-c-vrW 7, z?

Where: Fstrm= 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 $TATIONDATA 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 valu?" 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 CFS.Fd~i = flow rate of-the dilution flow in user liquid flow rate units.Rnx= 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 Fl.ow rate from user entered quarterly dilution flow rate. These values are from 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 processedi 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 streamflow for 2-16 the month is used, then each liquid release point entry in the CUMDOSE table f or the month could be multiplied by the ratio of the&ctual 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 iiermit 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 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 10CFR20 limits, and the actual dilution flow and waste flow are used instead of the anticipated dilution flow and waste flow.2-17 L3 o For batch releases, t'he duration is determined from the start and end dates and times, and is used with the volume input to calculate the release rate.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.2-18 ~5L'2-:18 P-5 LO I CHAPTER 3 GASEOUS RELEASE CALCULATIONS The "annual average X/Q" method is used, in which fixed X/Q and D/Q values are used fox 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 age group specified in the Gas Rec-eptors 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 IOCFR20 compliance.

The noble gas or air dose's and the organ doses are checked against the corresponding limits for IOCFR50 compliance.

3.2 RADIONUCLIDE ACTIVI.TIES 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 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 j r-bcik w Z~Z.

1. Composite Records 2. Particulate File 3. Radioiodine File 4. Noble Gas File The activity (Qi) and the activity release rate (Q0) are calculated for each nuclide i.Activity Released For the plant stack and turbine building vent: Qi Ci *Vv *28316.85-UF l 1 e-6 (pCi) =(pCi/ml) (cubic feet) (ml/cubic feet)where: V = vent release volume in user units (usually FT 3)Cr = concentration' in pCi/ml 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 pCi/sec is;i = Ci Vf 0 28316.85 9 UF/60 For containment purge:;= Ci pump release rate (CFM) 4 28316..85 e UF/60 0 Qi Qi duration of release (min)

60 3.3 1CCFR20 COMPLIANCE The maximum dose rate during the release is determined by summing together the dose rates for this release, with all concurrent releases'in the database for the time of the release.0@00 3-2 H 1,3 C.'DCW,_ V-C-,/

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 IOCFR20 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;Z X shf Z- [ (X/Q) v Q 3v < 500 mRem/yr f Zii vfr valloc" fs OR Z shf X V. Q. Vi < 500 rnem/yr f f v oo.er iv alloc s 0@ IElevated Stack t 80m Z. shf (L. + l.iMi) Z1 [ (X/Q))vr Qiv] < 3000 mRem/yr f a. a. a v yr falloc fa OR X shf X i(L + l.IBr) Q iv <.3000 mRem/yr f alloc "fs Elevated Stack 2r 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 Ev [fp P ip Wmv Qyi < 1500 m~em/yr f alloc fa Igo where: 313 f i 9 c'cL4-~.

~W Si = index over all radionuclides v= index over all vents or stacks for the unit p = index over all pathways r index for receptor locations= the total body dose factor due to ganuna emissions for noble gas radionuclide i, in mrem/yr per pCi/m 3.Li.= the skin dose factor due to beta emissions for noble-gas radionuclide i, in mrem/yr per pCi/m3r 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/r 3.Bir the elevated plume gamma skin dose factor for'nuclide i at receptor location r, in inrad/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/m3) and for ground plane and food pathways (in units of m 2 x(mrem/yr'per pCi/sec)).

The most restrictive age group is used.fp factor to select which pathways are included in the calculation.

Factor m 1 to include a pathway, 0 to exclude.WMv Q(X/Q)mv for tritium and the inhalation pathway and -(D/1)nv for other nuclides and pathways.(X/Q)vr = the highest value of the annual average 'atmospheric dispersion factor at the site boundary, for. all sectors, in sec/m 3.(X/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 3.3-ý4 PCR I ,C'4DDC-V%,,-

Q",( -Z--L-00 (DIQ)mv = the highest value of the annual average deposition factor at the distance of the site boundary, for all sectors, in m-2.Q;i =the average release r'ate o.f nuclide i in-gaseous effluent from release point v, in PCi/sec. Noble gases may be 'averaged over a period of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , 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 mrrem/year shf noble gas dose shielding factor falloc- fraction of the dose limit allocated to this release point*0 fs safety factor for the release point 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 dose rate)= for a vent release, lesser of 500 mrem/yr shf 2K 1 Q0. i (X/Q) mv 3-5 and*3000 mrem/yr shf Z (L. + 1.lM.) (-Q).1 .1 v mv* for an Elevated Stack t 80m, lesser of 500 mrem/yr shf Z V.* 6*S ir i and 3000 mrem/yr shf Z [ L. * (X/Q) + 1. ] I .r ir Radioiodines and Particulates In these cases, the ratio is obtained by summing over the appropriate nuclide indices: 1500 mrem/yr rpratio -= maximum organ dose rate zP i Qiv a 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) !+ B= offset + [slope.

Ci] t (quad (ZCij)2] + B 3-6 f..%_

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 a Ci] + [quadi * (Ci)2 J) + 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.

dp o= R n+ Z [offseti+

slopei cp + quad.dpmon dpmon ( p)2 1 00 where.C1p C. -(F IF .)I xp dp Ci =concentration for the release point 3 -7 Of~r-.A~~

re~u F rp= flow rate for the release point Fdp = flow rate for the discharge point W 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.jPon -offset + slope Z (ZCP ) + quad * (C )2 ÷ dpmon 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 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 W falloc" ratio -SUM The limiting setpoint for the monitor (in user units, e.g., cpm) is given by: SJ~mx =f s " alloc

ratio * (Rmon- B) + B 3-.8 00 0)where offset = I.2.3.slope = 1.2.3.noble gas offset factor radioiodine offset factor particulate offset factor noble gas slope factor radioiodine slope factor particulate slope factor quad 1. noble gas quadratic factor 2. radioiodine quadratic factor 3. particulate quadratic factor fs =safety factor for the release point fa~lloc dose rate allocation factor for the release point 0 'j ratio 1.2.3.SUM 1.2.3.Rion 1.2.3.nratio for noble gases rpratio for radioiodines rpratio for particulates I noble gas concentrations, for noble gases I radioiodine concentrations, for radioiodines I particulate concentrations, for particulates noble gas monitor response radioiodine monitor response particulate monitor response H = 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.a I 3-9 I CO 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: SUmsx ='[mrtol * (Rmon B)] + (mrrtolb B)-where.mrtol = 1. monitor response tolerance factor (noble gas)2. monit-or 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)

R.on =as defined above 3.4b REPORTED SETPOINTS The setpo~nt 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 defaulIt 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 O~4 e). 2-2-.

The reported setpoint is as follows: 1. Reported Sresponse.

if Sresponse

< Smax < Sdefault OR if Sdefault < Sresponse

< Snmx 2. Reported Smax if Sresponse

ý Sma_%3. Reported Sdefault if Sresponse

< Sdefault < Smax where Sax zi az defined .in the previous section r mrtol SUM [PCi/mi]Sresponse

-I L [mrtol (Rmon -B)] + (mrtolb

B) [User Units]Sdefault normal setpoint defined for the release point in units of [pCi/m)] 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 gCi/sec Setpoints in units of pCi/sec can be obtained by setting the UNITS, parameter for the monitor to "pCi/s" or "pCi/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 yCi/ml units, will belmultiplied by the corresponding effluent flow rate (release point or discharge point) for the monitor to get a reported setpoint in pCi/sec.3-11-572 OiOwý kv 'z 3.5 MAXIMUM WASTE FLOW The maximum waste flow calculation is -based on what the WFLOW M 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 Rwmax and Rcwmax where Rwmax -Release point maximum waste flow rate as stored in the release point definition If WFLOW M option = NONE Rcwmax= waste flow rate for the sample, Vf If WFLOW M option = DOSE fs nratio Vf Fwsfac 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:

0@3-12 f 3 f5~vA. q3A 4 Z 00 For Noble Gas Air Dose due to gamma radiation (for vents or stacks < 80 meters): (3.17 10-8)

X/Q t-a fo For Noble Gas Air Dose due to beta radiation (for vents or stacks < 80 meters): For Nob2 For Nob]Dp = (3.17 0 10-8) 0 X/Q t-b .fo a I Ni 0 Qiv Le Gas Total Body Dose Rate (for vents or stacks < 80 meters): Dt = shf

X/Qg 0 8 7 6 0-a

fo

1 (Ki e QRiv)le Gas Total Body Dose (for vents or stacks < 80 meters): shf f o " Z (Ki QRiv) , X/Qg

t-a.Dt=(5.256 -105 / dur).e Gas Skin Dose Rate (for vents or stacks < 80 meters): Ds = shf

f

Z QRiv S [(Ti -X/Q .8 7 6 0-b) + (1.11Mi

X/Qg S 8760-8) ]Le Gas Skin Dose (for vents or stacks < 80 meters): shf -fo 0 Z QRiv 0 (.(Li

X/Q

t-b) + (1.11Mi X/Qa t-avg*10 For Nobl For Nobl Dsk =(5.256

105 / dur)where Dp Dy Y Dt Dtb Ds.1 Dsk=total beta air dose from gaseous effluents (mrad)total gamma air dose from gaseous effluents (mrad)the total body dose rate due to gamma emissions by noble gas releases from vent v (mrem/yr)total body dose from gaseous effluents (mrem)skin dose rate from gaseous effluents (mrem/yr)skin dose from gaseous effluents (mrem)conversion factor from mrad to mr em 3-13 Pj.7L4 3.17

10-8 = .inverse of number of seconds in a year 5.256 l05 = number of minutes in a year 8 7 6 0-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 = "a" 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) total body dose factor due to gamma emissions for noble gas radionuclide i (mrem/yr 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 = inverse of number of seconds in a year Qiv = release of noble gas radionuclides, i, in gaseous effluents from vent or stack v (yCi)QRiv = release rate of noble gas radionuclides, i, in gaseous effluents from vent or stack v ( jCi/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/m 3)3-14 P5 15 ODLYAw (r-C' LZ-W-)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)OrQan Dose Calculations For Critical Organ Dose Rate--Inhalation Pathway and all Pathways for H-3, C-14 (for vents or stacks < 80 meters).DRza = X/Qr

8760-c v X PipTa

QRiv For Critical Organ Dose Rate--Ground and Food Pathways (for vents or stacks < 80 meters): DRra = D/Q 0 8 7 6 0-d X Z Ripra

QRiv For Critical Organ Dose-Inhalation Pathway and all Pathways for H-3, C-14 (for vents or stacks < 80 meters): Dra = (3.17

10-8)

X/Qr t-c fo

2 Pip. a

Qiv For Critical Organ Dose-Ground and Food Pathways (for vents or Stacks <80 meters): where Dra= (3.17 8) D D/Q -t-d .fo 0 2 RipTa

Qiv DRTa= dose rate for age group a and organ r.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/0 D/Q = highest Value of the 1-hour deposition factor at the distance of the site boundary (1/m2)PipTa = dose factor for each radionuclide i, pathway p, organ T, and age group a (mrem/yr per pCi/n 3)RipTa = dose factor for each radionuclide i, pathway p, organ T, and age group a (m2 .mrem/yr per pCi/sec)3-15 0i I Pn 7L, C;PC4k,_ rev e.;

X/Qr = highest value of the radioiodine/particulate 1-hour X/Q .W for vent or stack v at the site boundary (sec/m 3)Note: It is assumed PipTa will not contain long term XIQ 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 Aei 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 "EFFECT" release point activity for nuclide i Aci =the "CAUSE" release point activity for nuclide i Corrected concentrations are calculated as follows: Ccei (Acei / Ve) 6 35.315 3-16 where: Ccei =the corrected I"EFFECT" release point concentrations for nuclide i Ve =the waste volume for the "EFFECTi T release point 35.315 =conversion factor from Ci/ft 3 to FCi/ml (Ci/ft 3 ft 3/1728 in 3 d in 3/16.387 cm 3)3.8 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 I(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.3-17 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 IOCFR20 limits o Check 1OCFR50 limits o Add.the dose and- activity data into the cumulative totals.Compliance with IOCFR20 limits is checked in the same way as-described for the pre-release program..Dose rates are calculated and compared to 10CFR20 limits. Monitor sjetpoints are not calculated at the post release stage.0., 3-18 79 0.CHAPTER 4 LIQUID DOSE FACTOR EQUATIONS The DFP option is 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.

For a particular receptor, the total dose factor (AiT~r) is a sum over each pathway p with its specific mixing ratio: AiTr m ixr,p A r,p XR .mix, r where AiTr,p =the dose factor for nuclide i, organ T, receptor age group r, and pathway p mix, r,p mixing ratio for the pathway Rmix,r -mixing ratio for the receptor, which is the first non-zero value of "ix,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-zero Value. If the receptor mixing ratio for a given pathway is zero, that term is not included in the sum.0@4-:1 re,/ 12-The DFP option of EM-S 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 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: AiT.rjp= kO (U-rp /dw). Ni *.DiT,r *e(-ýitp where AiT,r,p =dose parameter for organ T, for the receptor age group r, for nuclide i, due to exposure pathway p, in mrem/hr per pCi/ml ko= units conversion factor, = 1.142E5 = IE6(pCi/VCi) 1000 (ml/Kg)/ 8760 hr/yr Urp -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.DFiT,r ingestion do-se 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 = decay constant for nuclide i t = average transit time in seconds 4.2 AQUATIC FOODS PATHWAYS The liquid dose factor is Ai.rxp kO

Urp BF .' Ni. DFiTzr.- exp(-"it p %4-2

)where B1 p = bioaccumulation factor for pathway p and n~clide i (from Reg. Guide 1.109, Table A-i). 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:-Il-e i tb _ t AT r~= k W NiT ei sd DG ir f p s s i f,rp e. DFGiT A.1 where Ws shoreline width factor ks conversion factor = k 0 *kc m tv/3600 kc =water to sediment transfer coefficient in L/kg hr rtv Mass density of'sediment in kg/m2, 40 kg/r 2 3600 = Seconds per hour units conversion factor tb length of time sediment is exposed to contaminated water, 4.716E8 sec tsd transit time tq deposit activity on shoreline DFGir = 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= 1.24 * -_10 AiT, r,p U fr CF i DFiT,r where: 1.14

10 5 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 irrigated vegetables, CFiv as used in the equation for AiT, calculated as follows for all radionuclides other than Tritium: in is CFiv =N a M 0 1 (1 -e ie yv Ei For Tritium, the equation is as follows: CFiv -Ni e M

Lv where fIBiv (I e-e P A.-A it e 2 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/m 2 h). " the fraction retained on vegetables.

available for of irrigation-deposited activity the edible portions of leafy There are separate values radioiodines and particulates.

Yv= the agricultural productivity of irrigated leafy vegetables (kg/m 2).4-4 AML.-Arý 8'V-3 5 C)_NoM Tew 2

= the fraction of the year that vegetables are irrigated.

Biv = the crop to soil concentration factor applicable to radionuclide

i. (p;i/kg vegetables)/(pCi/kg soil).P the effective surface density of soil (kg/m 2).Ai -the decay constant for radionuclide i (h-1).AEi -the effective removal rate for activity.deposited on crop leaves (h-1), calculated as AEi Ai + Aw Aw = the rate constant for removal of activity from plant leaves by weathering (h1l).ýte = 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).L9j = 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 1 (this can be set in the definition of Fraction of Activity Reaching Receptor in DFP)average transit time t p 0 (this can be set in the definition of Dose Calculation Parameters in DFP)4-5 0Ociýw rev~ 2

))0 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 (mrem/yr per pCi/m3 K' = 1E6 pCi/pCi)1 0 5-1 04ý-P3 4D___3 r"/ Z.2 (BR)a = breathing rate for age group a, in cubic m/yr (DFAij)a -inhalation dose factor for organ T, for age group a, for nuclide i, in mrem/pCi 5.2 GROUND PLANE PATHWAY RiTa K'K" (SF) DFGiT [(I -e-Ait)1Ai](m2 -mrem/yr per pCi/sec).where K' 1E6 pCi/lCi K" 8760 hr/yr Ai = decay constant for nuclide i, in sec-1 t = exposure time (sec) = 4.73E8 (15 years)DFGiT= ground plane conversion factor for. nuclide i, organ r (The sameDFGi, 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 RiTa -K' (DFLir)a.

e-Aitf QEF, p Uap, 0 0~I-(A +A. )t[r (1-e i w e)fp Y[ Y (A. +A)(1 e- itb) 1 p A. j a-1 5-2 O0Li%'-(es72-2

+ (1-f f )e- i h P s-r (1-e-(Ai+

A w e) ( -e- itb)+ B.Y (A. + ) iv s 1 I (I (m, 2-mrem/yr per pCi/sec)where K' = IE6 pCi/pCi Q= feed consumption rate by the milk animal (cow or goat)(Kg/day)Uap = 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=mi 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.

Lir)a = ingestion dose factor for organ T, for nuclide i, for receptor in age group a, in mrem/pCi Ai =decay constant for nuclide i w= decay constant for removal of activity on leaf and plant surfaces by weathering) in sec-I 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.f = fraction of year that animal is on pasture 5- 3 (DF 0 p38.(<)bcW"._ ve-x/ Z;ý fs = fraction of animal feed that is pasture grass while animal is on pasture Carbon-14 in Milk Rira K'K"' Fm .Q Uap (DFLi)a PC (0.11/0.16) e-Aitf (m2-mrem/yr per pCi/sec)where K"' IE3 gm/Kg P- fractional equilibrium ratio 0.11 = fraction of total plant mass that is- natural carbon 0.16 = concentration of natural carbon in the atmosphere (gfm3)and all other parameters as defined above Only Qjr and Uap depend on cow or goat.Tritium in Milk RiTa =K'K"' Fmi OF Uap (DFLiT)a (0.75) (0.5/H) e-Aitf (m 2-mrem/yr per pCi/sec)where Klr = 1E3 gm/Kg H = 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 4 Ii 5-4 P5 86.OW& fC-V 2--)

0 0 Only QF and Uap' depend on cow or goat.5.4 MEAT PATHWAY RiTa = KI (DFLiT~a eAitf F Ffi Uap ( -e-(A .+A )t -A Y -(-- (t -A +t )X t ( --e ( i wA w e) e-e-- ib b (1-f f )e ih + B. h A+SYs (Ai+ Aw) IV A I I 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 frpm pasture to receptor, in sec and all other'factors are as described for the cow-milk pathway Carbon-14 in Meat K'K" Ffi QF Uap (DFLiP)a Pc (0.11/0.16) e-itf (m 2-mrem/yr per pCi/sec)where all terms are as defined above.5-:5 0fev 22.0O Tritium in Meat Rira = K'K"' Efi QF Uap (DFLir)a (0.75) (0.5/H) e-ýitf (m2-mrem/yr per pCi/sec)where all terms are as defined above.5.5 VEGETABLE PATHWAY (L -At R K' (DFL U f e itL RiTa iT a L-(A.+A )t -A t r (1-e 3. w e) Bi (l-e itb)Y~ (A. + A) p A.V -L W i S -.t+ f e i s a g-(A.+A )t -A t r(-e 3. W e) B. (1-e i b)iv Y (A.+ A p A (m 2 mrem/yr per pCi/sec)where U La = consumption rate of fresh leafy, vegetation for age group a, in Kg/yr us= consumption rate of stored vegetation for age group a, in Kg/yr fL= fraction of annual intake of leafy vegetation grown locally fg fraction of annual intake of stored vegetation grown locally* @1 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/m2 p = effective soil surface density Biv =soil to vegetation transfer factor for nuclide i All other factors are as defined above.Carbon-14 in Vegetables Rira = KIK"' (T( + UL ) (DFLit)a Pc (0.11/0.16) e-Aitf (m 2-mrem/yr per JCi/sec)where all variables are as defined earlier.Tritium in Vegetables Rira "= 'K"' (UL+ U') (DFLir)a -(0.75) (0.5/H) e-.itf (m 2-mrem/yr per yCi/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: tbO 5-7 GYY4C(7e~vzz te = 9.999E19 tf = 0 (in tritium equations only)There are no C-14 equations in NUREG-0133, which can be obtained by setting Pc = 0.0 0 5-8 cPaw f(A 2-2ý

)0 APPENDIX A REFERENCE S 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 1October 1977).0)Vitý3

SBK-L-12097, Offsite Dose Calculation Manual, Revision 36, Program Manual

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SBK-L-12097, Offsite Dose Calculation Manual, Revision 36, Program Manual

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