ML071280622

Rev. 28 to Off-Site Dose Calculation Manual (ODCM)

ML071280622
Person / Time
Site:	Nine Mile Point
Issue date:	09/28/2006
From:	Fiorenza T, Hutton J A, Kurtz T M, Schimmel M A, Stinson G R Constellation Energy Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML071280622 (185)

Text

{{#Wiki_filter:Nire l,- IIEý Point Nu.clear Stationr NINE MILE POINT NUCLEAR STATION NINE MILE POINT UNIT 1 OFF-SITE DOSE CALCULATION MANUAL (ODCM)DATE APPROVALS Prepared by: SIGNATURES T. M. Khyrtzs Health Physicist REVISION 28 Z1 J'2~Reviewed by: Concurred by: G. R. tin~hson.. J. A. Hutton flai , 96 Ml~a~ ~ienral Manager M. A. Schimmnel 14 -4ift Man~ager Enginee~%ng'Services 2 S UMMAR Y OF REVISIONS Revision 28 (Effective September 29, 2006)PAGE DATE 1, 2,5,6, 8, 9, 11-13/15-18, 21, 24, 25, 36-44, 47-49, 52-81, 86-116 Februaryv1987 3, 4, 7, 10, 14, 19,.20, 22, 23, 26-35 December 1987 45, 46, 50, 51,.82-85 January 1988*29 May 1988 (Reissue)*64, 77, 78 May 27,1988 (Reissue)i, 19,:21, 22A, 22B, 124, 25,26, 112 February 1990 i, ii, iii, 12-16, 18, 28-40, 45-47 52, 55, 59-89, 92, 93, 97-129 June 1990 91-93,95 June 1992 3, 4, 21, 92, 95a-c February 1993 10, 16-20 March 1993 5, 13, 18, 20, 25-30, 65, 79 June 1993 66, 69 December 1993 16, 69 June 1994 10,12, February 1995 i10, 18,,67,69 December 1995 5, D-1 June 1996 5, D-1 June 1997 5, fD-i April 1999 D-1 December 1999 iv, 3, 6, 8, 9, 11, 13, 14, 27,29, 65, 66, 69, 69a December 2001 Added Part .& Revised Part i1 -!112-16, II 20-23 I1 25,11 26, II 29, Ii 30 November 2002 iv, v, vii, viii, 1.10-1 and 2,1I 3.1-1, 7 to 9, 11, 14, 18 to 24, 26 and 27, I B 3.1-1, 3 to 7,1 6.0-2, 4, and 5, 11 2,11 3,11 4,116, 1,9 to 11, 1113 to 22,1142, Figure D-8, Deleted Figures D-7, D-9, D-10 November 2002 Unit I ODCM Revision 28 September 2006 i

SUMMARY

OF REVISIONS (continued) Revision 28 (Effective Septemnber 29 2006)PAGE x, 1 1.0-1, 1 3.1-22, 1 3.1-38 and 39, I B 3.1-1, 16.1-0 and 3, Il 11, 12,17,18,24 and 25 DATE July 2003 1 3.1-7,1 3.1-8, 13.1-9,1 3.1-10,1 3.1-11, 1 3.1-12 and I B 3.1-1 1170, 1171, and II 73 1 3.1-5,13.1-10, 116, and I 24: February 2004 December 2005 May 2006 September 2006 1 1.0-1, 13.1-1, 1 3.1-2, 1 3.1-3, 1 3.1-4,13.1-5,1.3.1-7, 13.1-8, 13.1-9,1 3.1-10 1 3.1-11, 1 3.1-12, 13.1-23, 13.1-27, 13.1-28, 1 B 3.1-1, 1 B 3.1-8, 1 6.0-4, II 2, 115. II 11,11 17, and 1125 Unit I ODCM Revision 28 September 2006 ii ODCM -NINE MILE POINT UNIT 1 TABLE OF CONTENTS PAGE List ofTables............... ! ,.,......... ............................................................................ ................................ viii List of Figures .................................................................................................................................................. ix INTRODUCTION ................................................ e ............................................................................................. x PART I -Radiological Effluent Controls SECTION 1.0: SECTION 2.0: SECTIONS 3.0/4.0: 0 3/4.6.14 D 3/4.6.14.a. D 3/4.6.14.b D 3/4.6.15 D 3/4.6.15.a.(2) D 3/4.6.15.b.(1) D 3/4.6.15.bd(2) D 3/4.6.15.b.(3) D 3/4.6.15.d D 3/4.616.6 D 3/4.6.16.a D) 3/4.6.16.b D 3/4.6.17 D 3/4.6.18 D 3/4.6.19 D 3/4.6.20 D 3/4.6.2 1 D 3/4. 6.22 Definitions., ..... ...... .................. ...................................................................... i 1.0-0 Not Used Applicability ................................................... ............................................... I 3.0-0 RADIOACTIVE EFFLUENT INSTRUMENTATION ................................ 13.1-1 Liquid Effluent ............................................................................................ 13.1. 1 Gaseous Process and Effluent..... ............................... 13.1-7 RADIOACTIVE EFFLUENTS ..................................................................... 13.1-14 Liquid Concentration .. ................. ......... I.......... 1-14 LiquidDose .................................. 13.............. .......... ....... 3.1-15 Gaseous Dose Rate .......................................................................................... 13.1-19 Gaseous Air Dose ....................................................................................... 13.1-20 Gaseous Tritium, lodines and Particulates .................................................. 13.1-21 Uranium Fuel Cycle ..................... ....... ............................................................ 13.1-24 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS ........................ I 3.1-27 Liquid ..... ............. ................................. 13.1-27 Gaseous .................................................................................................... 13.1-27 Not Used MARK I CONTAINMENT ............................................................................ 13.1-29 LIQUID WASTE HOLDUP TANKS ........................................................... 3.1-30 RADIOI ,OGICAL ENVIRONMENTAL MONITORING PROGRAM.. 1 3.1-31 INTERLABORATORY COMPARISON PROGRAM ........................... 13.1-41 LAND USE :CENSUS ....................................... 13.1-42 Unit I ODCM Revision 28 iii September 2006 ODCM -NINE MILE POINT UNIT 1 TABLE OF CONTENTS (Cont)PA GE B 3/4.6.14 B 3/4.6.15 B 3/446.16 B 314.6,18 B 3/4.6.19 B 3/4.6.20 B 3/4.6,21 B 3/4.6.22 SECTION 5.0 SECTION 6.0 BASES.... ....... ý.a ........ ,............ .. ............. .I B 3.1-0 BASES FOR RADIOACTIVE EFFLUENT INSTRUMENTATION ............... I B 3.1-1 BASES FOR RADIOACTIVE EFFLUENTS .................................................... I B 3.1-2 Liquid Concentration ....................................................................................... I B 3.1-2 Liquid Dose .................................... ........... B3.13 Gase.......Ra.e................. .... I B 3.1-3 Gaseous'D ose Rate i'" ........ I ..............-.-... ..... I...... ......... ...... .e ,, ..... ...,I B 3.14 Dose-Noble Gases ............. ... ...... ... ............................ ............. ......-Dose-lodine-131, Iodine-133,Tritium, and Radionuclides in Particulate Form ......................................................................................... I B 3.1-6 Total Dose-Uranium Fuel Cycle ....................... .............. .................. 1B 3.1-7 BASES FOR RADIOACTIVE EFFLUENT TREATMENT SYSTEMS ......... I B 3.1-8 Liquid Radwaste Treatment System.. ... ! ......... .... ......................... I B 3.1-8 Gaseous Effluent Treatment Systems ............. ........ .......... I B 3.1-8 BASES FOR MARK I CONTAINMENT....... ....... ............. I B 3.1-9 BASES FOR LIQUID WASTE HOLDUP TANKS .............................................. I B 3.1-9 BASES FOR RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM .................... , ............... o ................................. I B 3.1-10 BASES FOR INTERLABORATORY COMPARISON PROGRAM.............. I B 3.1-11 BASES FOR LAND USE CENSUS........................ .... e .....I B 3.14-2 Not Used ADMINISTRATIVE CONTROLS .................................................... I 6.0-1 Reporting Requirements ......... ....... ........................ 16.0-2 Special Reports ....,............. ... ... ........... .............. ,.......1 6.0-4 Unit I ODCM Revision 28 September 2006 iv ODCM -NINE MILE POINT UNIT 1 TABLE OF CONTENTS (Cont).PAGE PART II -Calculational Methodologies .................. .............. ........... I...1.0 LIQUID EFFLUENTS ................................................................................................................. 112 1.1 *Setpoint Determinations .......... 112................................................................................ 112 1.1.1 Basis.. ............................ .......................................... 1..................................................... 112 1.1.2 Service Water System Effluent Line Alarm Setpoint ............................................ I 2 1.1.3 Liquid Radwaste Effluent Line Alarm Setpoint .................................................... II 3 1.1.4 Discussion ........................................... .......... .......... ........ ...... .. .... ..1.1.4.1 Control of Liquid Effluent BatchDischarges .............. ............................................ 5 1.1.4.2 Simultaneous Discharges of Radioactive Liquids ................................................. 11 5 1.1.4.3 Sam ple Representativeness .......................... ................. .................. ............... .5 1.1.4.4 Liquid Radwaste System Operation ................................ .6 1.1.4.5 Service Water System Contamination ......................................................................... 117 1.2 Liquid Effluent Concentration Calculation .......................................................... 11 7 1.3 Dose Determinations ............................................ 11 8 1.3.1 Maximum Dose Equivalent Pathway ..................................................................... 8 1.3.2 Dose Projections -Determination of Need to Operate the Liquid Radwaste Treatm ent System .................................................................................................... .11 2.0 GASEOUS EFFLUENTS .................................................. 1112 2.1- Setpoint Determinations, ...... e .................................................................................. 11 12 2.1.1 Basis ........................................................................................................................... 2 I 2 2.1.2 Stack Monitor Setpoints ....................................................................................... 11 12 2.1.3 Recombiner Discharge (Off Gas) Monitor Setpoints .......................................... 1114.2.1.4. Emergency Condenser Vent Monitor Setpoint ...... .......................................... II 15 Unit I ODCM Revision 28 V September 2006 ODCM -NINE MILE POiNT UNIT 1 TABLE OF CONTENTS (ConO PAGE 2.1.5 Discussion... ...................................................................................................... 115 2.1.5.1 Stack Effluent Monitoring System Description ................................................. Il 115 2.1.5.2 Stack.Sample Flow Path -RAGEMS Auxiliary Sample Point ..1........... ............... 1115 2.1.5.3 Stack Sample Flow Path -OGESMS... .......... ... ....... ....U 16 2.1.5.4 Sam ple Frequency/San ple Analysis ..................................................................... 1'16 2.1.5.5 1-133 and 1-135 Estimates ................. o ................... i .................... 116 2.1.5.6 Gaseous Radwaste Treatment System Operation ......................................... l.117 2.2 Dose and Dose Rate Determinations,..... ....... ... .................. II 17 2.2.1. DoseRate ........... ..................................................... .............................................. 11.18 2.2.1.1 Noble Gases ..... ...... ........................................... II 19 2.2.1.2 Tritium, Iudines and Particulates ..................... ............................. ............... 1120 2.2.2 Dose ................................................................ 2..... 112 2.2.2.1 Noble GasA ir Dose ................................................. .............................................. 1122 2.2.2.2 Tritium, Iodines and Particulates:...... ........ .................. ...... 123 2.2.2.3 Accumulating Doses .................................................................................. .124 2.2.3 Dose Projections -Determination of Need tolOperate Gaseous Radwaste Treatment System and Ventilation Exha ust Treatment System ............. 11..... .24 2.3 Critical Receptors...,.......... ....................... ............................... r.i .......R......1.........,..I125

2.4 Refinement

of Offsite'Doses Resulting From Emergency Condenser Vent Releasesý ...... ............................ ...... ..... ........................... it 26 Unit 1 ODCM Revision 28 vi September 2006 ODCM -NINE MILE POINT UNIT 1 TABLE OF CONTENTS (Con0 PAGE 3.0 40 CFR 190 REQUIREM ENTS. .................................................................................................. 11 27 3.1 Evaluation of Doses From Liquid Effluents ....................................................... 128 3.2 Evaluation of Doses From Gaseous Efflue.ts. .. .................... ,... ...... H29:3.3 Evaluation of Doses From Direct:Radiation ......................................................... 1130 3.4 Doses to Members ofthe Public Within the Site Boundary......... ................ 4 .... 1130 4.0 ENVIRONMENTAL MONITORING PROGRAM ................................................................... 1133 4.1 Sampling Stations ....... ; ...... .. .............. ....... .................

11. 33 4.2 Interlaboratory Comparison Program .....................

.............. .. ...................... 1133 4.3 Capabilities for Thermoluminescent Dosimeters Used for Environmental M easurements ..................... ........ ............................................... 1134 Appendix A Appendix 1H Appendix C Appendix D Liquid Dose Factor Derivation (Ait) ....................................................................... 11 75 Plnme Shine Dose Factor Derivation (Bi and Vi) ............ ......................... 1178 Organ Dose Parameters for Iodine -131 & 133,:Particulates and Tritium (R 1) ...................... .... .................. 1182 Diagrams of Radioactive Liquid and Gaseous Effluent Treatment Systems and Monitoring Systems ............. ............................. 11 92 Unit I ODCM Revision 28 vii September 2006 ODCM- NINE MILE POINT UNIT I LIST OF TABLES PAR TI -Radiological EFfluent Controls PAGE D 3.6.14-1 Radioactive Liquid Effluent Monitoring Instrumentation. ............. 13.1-3 D 4.6.14-1 Radioactive Liquid Effluent Monitoring Instrumentation -SR.... ... 13.1-5 D 3.6.14-2 Radioactive Gaseous Process and Effluent Monitoring Instrumentation ....... .......................... ............ 1 3.1-8 D 4.6.14-2 Radioactive Gaseous Process and Effluent Monitoring Instrumentation -SR ............ ............................... 1 3.1-12 D 4.6.15-1 Radioactive Liquid Waste Sampling and Analysis Program -,SR ......... 3.1-16 D 4.6.15-2 Radioactive Gaseous Waste Sampling and Analysis Program -SR....... 13.1,22 D 3.6.20-1 Operational Radiological Environmental Monitoring Program ............. 13.1-34" 4.6.20-1 DetectionCapabilitieS rfor Environmental Sample Analysis Lower Limit of Detection LLD -SR ........................................ ....... ......... 13.1-38" 6.9.3-1 Reporting Level for Radioactivity-Concentration in'Environmental Samples .. ..................................... 1:6.0-5 PARTII -Calculational.Methodologies Table 1-1 Average Energy Per Disintegration ............ ....................... 3......6....... I,........... 1 6 Tables 2-1. Aiat Values for the NMP-1 Facility........................... ............ a 1..........

4137 to 2-8 Table 3-1 CriticaL

Receptor Dispersion Parameters for Ground Level and Elevated Releases ......................................................

11 45 Table 3-2 Gamma Air and Whole Body Plume Shine Dose Factors for Noble Gases (Bi and Vi) ............., ..... ....................... 1146 Table 3-3 Immersion Dose Factors for Noble Gases. ....................................................... 1147 Tables 3-4 to 3-22 Dose and Dose Rate Factors (Ri) .................................. ..... a 48 Table 3-23 Parameters for the Evaluation of Doses to Real Members of the Public from Gaseous and Liquid Effluents ................................................. 1167 Table 5.1 Nine Mile Point Nuclear Station Radiological Environmental Monitoring Program Sampling Locations ........................................................... ............................................ II 68 Unit I ODCM Revision 28 Viii September 2006 ODCM -NINE MILE POINT UNIT I LIST OF FIGURES Figure 5.1-1 Figure 5.1-2 Figure:5.1.3-1 Figure D-0 Figure DA Figure D'2 Figure D-3 Figure :D-4 Figure D-5 Figure D-6 Figure D-7 Figure D-8 Figures D-9, D-10 Figure D-11 PAGE Nine M ile Point On-Sife M ap ................................ ....................................... I 72 Nine M ile Point Offsite M ap ......................................................... .................. 1173 Site Boundaries.; ............................................. 4 .................... Piping Instrument and Equipment Symbols ..................... ........................ D-0 Radioactive Waste Disposal ..................................... D-1 Stm Packing, Exhauster, and. Reconmbincr ........................ .................. D-2 Reactor Building Vent System .................................. D-3 Waste Disposal Building Vent System .................................. D-4 NMP-1 Stack............... .............. I ................... Dý5 Offgas Building Vent System ............................................................................ D-6 Th is Page/Figure Deleted Stack Sample and Sample Return ................................ D-8 These Pages/Figures Deleted OGESMS Schematic: ........................................ D-11 Unit 1 ODCM Revision 28 ix September 2006 INTRODUCTION The Offsite Dose Calculation Manual (ODCM) provides the methodology to be used for demonstrating compliance with 10 CFR 20, 10 CFR 50, and 40 CFR 190. The contents of the ODCM are based on Draft NUREG-0472, Revision 3, "Standard Radiological Effluent Technical Specifications. forPresSurized Water Reactors," September 1982; Draft NUREG-0473, Revision 2, "Radiological Effluent Technical Specifications for BWR's", July 1979; NUREG 0133,"Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," October 1978; the several Regulatory Guides referenced in these documents; and, communication with the-NRC staff.Should it be necessary to revise the ODCM, these revisions will be made in accordance with Technical Specifications. The Offsite Dose Calculation Manual (ODCM) is a supporting document of the Technical Specifieatioiis Section 6.5.1, "Offsite Dose Calculation Manual." The previous Limiting Conditions for Operation that were contained in the: Radiological EffluentTechnical Specifications are now transferred to the QDCM as Radiological Effluent Controls. The ODCM contains two parts, Radiological EffluentControls Part I; and Calculational Methodologies, Part 1I. Radiological Effluent Controls, Part I, includes the following: (I) The Radioactive Effluent, Controls and Radiological Environmental Monitoring Programs required by Technical Specifications 6.5.3, "Radioactive Effluent Controls Program" and 6.5.1, "Offsite Dose Calculation Manual", respectively, and (2) descriptions of the infornation that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by Technical Specifications 6.6.2, "Annual Radiological Environmental Oper-ating Report" and 6.6.3 "Radioactive Effluent Release Report". Calculational Methodologies, Part II, describes methodology and parameters to be used in the calculation of liquid and gaseous effluent monitoring instrumentation alarm/trip setpoints and the calculation of offsite doses due to radioactive liquid and gaseous effluents& The ODCM also contains a list and graphical description of the specificsample locations for the radiological environmental monitoring program, and liquid and gaseous radwaste treatment system configurations. Unit I ODCM Revision 28 x September 2006 PART I -,RADIOLOGICAL EFFLUENT CONTROLS Unit 1 ODCM Revision 28 September 2006 I PART I-- RADIOLOGICAL EFFLUENT CONTROLS Section 1.0 Definitions Unit I ODCM Revision 28 September. 2006 11.0-0

1.0 DEFINITIONS

DEFINITIONS 1.0 NOTE: Technical Specifications, defined terms and the following additional defined terms are applicable throughout these controls and bases.Functional (Functionality) Functionality is an attribute of Structures, Systems, or Components (SSCs) that is not controlled by Technical Specifications. An SSC shall be functional or have functionality when it is capable. of performing its-specified function as set forth in the Current Licensing Basis (CLB).Functionality does not apply to specified safety functions, but does apply to the ability of non-Technical Specifications SSCs to perform specified support functions. Gaseous Radwaste Treatment System A gaseous radwaste treatment system is any system designed and installed to reduce radioactive gaseous effluents by collecting main condenser offgas and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment. Member(s) of the Public Member(s) of the public shall include persons who are not occupationally associated with the NineMile Point.Nuclear Station. This category does not include employees of owners and operators of Nine Mile Point NuclearStation and James A. Fitzpatrick Nuclear Power Plant, their contractors or vendors who are occupationally associated with Nine Mile Point Unit 1.. Also excluded from this category are persons who enter the site. to service equipment orto make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with Nine Mile Point Unit 1.Milk Sampling Location A milk sampling location is that location where 10 or more head of milk animals are available for the collection of milk samples.Unit I ODCM Revision 28 I 1.0-i September 2006 DEFINITIONS

1.0 Offsite

Dose Calculation Manual (ODCM)The Offsite. Dose Calculation Manual shall contain thecurrent methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation .of gaseous and liquid effluent monitoring alarm/trip setpoints, and in the conduct of the environmental radiological monitoring program. The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by Technical Specifications 6.6,2, "Annual Radiological Environmental Operating Report" and 6.6,3,"Radioactive Effluent Release Report", and Controls D 6.9.11d and D 6.9:1.e.Purge -Purging Purge or purging is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in.such a manner that replacement.air or gas is required to purify the:confinement The purge is completed when the oxygen concentration exceeds 19.5 percent.Site Boundary The site boundary shall be that line around the. Nine Mile Point Nuclear Station beyond which the land is neither owned, leased, nor otherwise. controlled by the owners and operators of Nine Mile Point, Nuclear Station and James A.. Fitzpatrick Nuclear Power Plant.Source Check A source check shall be the:qualitative assessment of channel response when'the channel sensor is exposed to a source of increased radioactivity. Unrestricted Area The unrestricted area shall be any area at or beyond the site boundary access to which is not controlled by the owners and operators of Nine Mile Point Nuclear Station and James A. Fitzpatrick Nuclear Power Plant for purposes of protection of individuals.from exposure to radiation and radioactive materials, or any area within the site boundary used for residential quarters orfor industrial, commercial, institutional, and/or recreational purposes. That area, outside the restricted area (10 CFR 20.1003) but within the site boundary will be controlled by the owner as required.Unit. ODCM Revision 28[ 1.0-2. September 2006 DEFINTIONS

1.0 Ventilation

Exhaust Treatment System A ventilation exhaust treatment system is any system designed and.installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers andlor HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is. not considered to have any effect on noble gas effluents. EngineeredSafety Feature (ESF) atmospheric cleanup systems are not considered to be ventilation exhaust treatment system components. Venting Venting is the controlled process of discharging air or gas from. a confinement to maintain temperature, pressure, humidity, concentrationor other operating condition, in such a manner that: replacement air or gas is not. provided or required during venting. Vent, used in system names, does not imply a venting process.Unit 1 ODCM Revision 28 September 2006 1 1.0-3 PART I -RADIOLOGICAL EFFLUENT CONTROLS-......1 1 . ............Sections 3.0/4.0 Applicability Unit I ODCM Revision 28 September 2006 1 3.0-0

3.0 CONTROLS

APPLICABILITY 3.0/4.0 The Offsite Dose Calculation Manual (0DCM) Part I, Radiological Effluent Controls, is subject to Technical Specifications Section 3.0 requirements, as applicable.

4.0 SURVEILLANCE

REQUIREMENTS The ODCM Part 1, Radiological Effluent Controls, is subject to Technical Specifications Section 4.0 requirements, as applicable. Unit I ODCM Revision 28 September 2006 13.0-1 RADIOACTIVE EFFLUENT INSTRUMENTATION -LIQUID D 314.6.14 CONTROLS SURVEILLALNCE RFOLIIPMENT DLCO 3.6.14 RADIOACTIVE EFFLUENT INSTRUMENTATION Applicability: Applies to the functionality of plant instrumentation that monitors plant effluents. Obiective: To assure the functionality of instrumentation to monitor the release of radioactive plant effluents. Specification:

a. Liquid Effluent The radioactive liquid effluent monitoring instru-mentation channels shown in Table D 3.6.14-1 shall be functional with their alarm setpoints set to ensure that the limits of Control DLCO 3.6.15.a.1 are not exceeded.

The alarm setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in Part H.With a radioactive liquid effluent monitoring instrumentation channel alarm setpoint less conservative than a value which will ensure that the limits of DLCO 3.6. 15.a.1 are met, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel nonfunctional, or change the setpoint so it is acceptably conservative. DSR 4.6.14 RADIOACTIVE EFFLUENT INSTRUMENTATION Applicability: Applies to the surveillance of instrumentation that monitors plant effluents. Obiective. To verify operation of monitoring instrumentation. I I Specification.

a. Liquid Effluent Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated functional by performance of the sensor check, source check, instrument channel calibration and channel test operations at the frequencies shown in TableD 4.6.14-1.I1 Records -Auditable records shall be maintained, in accordance with procedures in Part II, of all radioactive liquid effluent monitoring instrumentation alarm setpoints, Setpoints and setpoint calculations shall be available for review to ensure that the limits of Control DLCO 3.6.15.a.1 are met.Unit 1 ODCM Revision .28 September2006 1 3.1-1 RADIOACTIVE EFFLUENT INSTRUMENTATION

-LIQUID D 3/4,614 CONTROLS.SURVEILLANCE REQUIREMENT i-.With less'than the minimum number of radioactive liquid effluent monitoring instrumentation channels functional, take.the. action shown in Table D 3.6.14-1. Restore the instruments to functional status within 30 days, or outline, in the next Radioactive Effluent Release Report the cause of the nonfunctionality and how the instruments were or will be restored to functional status.I Unit 1 ODCM Revision 28 September 2006 13.1-2 RADIOACTIVE EFFLUENT INSTRUMENTATION LIQUID D 3/4.6.14 TABLE D 3.6.14-4 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Instrument

1. Gross Radioactivity Monitors(a)

A, Liquid Radwaste Effluent Line B. Service Water System Effluent Line, 2. Flow Rate Measurement Devices A. Liquid Radwaste Effluent Line B. Discharge Canal 3. Tank Level Indicating Devices(g) A. Outside Liquid Radwaste Storage Tanks Minimum Channels Functional Annlicahilitv .A licaIli 1i (c)At all times_(b)At all times()1 !e At all times**1 (At all times**Pumps curves or rated capacity will be utilized to estimate flow.Unit I ODCM Revision 28 September 2006 1 3.103 RADIOACTIVE EFFLUENT INSTRUMENTATION -LIQUID D 314,6.14 NOTES FOR TABLE D 3.6.14-1 (a) Provide alarm, but do not, provide automatic termination of release.(b) An operator shall be present in the Radwaste Control Room at all times during a release.(c) With the number of channels functional less than required by the minimum channels functional requirement, effluent releases may-continue provided that prior to initiating a release: 1 At least two independent samples are. analyzed in accordance with Specification DSR 4.6.15.a, and.2. At least two technically qualified members of the Facility Staff independently verify the release rate calculationS. and discharge line valving.Otherwise suspend release of radioactive effluents via this pathway.(d) With the number of channels functional less than required by the minimum channels functional requirement, effluent: releases via this pathway may continue provided that, at least once per 12 hours, grab samples. are collected and analyzed for gamma radioactivity at a lower limit of detection of at least 5xW0e microcurie/ml.(e) During discharge, with the number of channels functional less than required by the minimum channels functional requirement, effluent. releases via this pathway may continue provided the flow rate is estimated at least once pet 4 hours during actual releases.(f) With the number of channels furnctional lessthan required by the minimum channels functional requirement, liquid additions to this tank may Continue provided the tank liquid level is estimated during liquid additions to the. tank.(g) Tanks included in this specification are those outdoor tanks. that are not surrounded by liners, dikes or walls Capable of holding the tank contents.(h) Deleted.(i) Monitoring will be Conducted continuously by alternately sampling the reactor building and turbine building service water return lines for approximately 15-minute intervals. Unit I ODCM Revision 28 1 3.1-4 September 2006 RADIOACTIVE EFFLUENT INSTRUMENTATION -LIQUID DZ34,6.14, RADIOA Instrument

1. ýGross Beta or Gamma Radioactivity Monitors a. Liquid Radwaste Effluent Line b. Service Water Effluent Line 2. Flow Rate Measurement Devices a. Liquid Radwaste Effluent Line b. Discharge Canal(")3. Tank Level Indicating Devices(e)

a. OutsideLiquid Radwaste Storage Tanks TABLE D 4.6.14-1&CTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Surveillance Requirement Sensor Check Source Check(Q Channel Test Test Chasnne iI (C2krlih-1 Once/day*Once/day Once/day(c)

None Once/day** Once/discharge* Once/92 days None None None Once/3 months(aW* Once/1 84 daystal None:,None Once/3 months Once/year~h)* Once/24 months(b)Once/24 months*Once/year Once/i 8 months I* Required prior to removal of blank flange in discharge line and until blank flange is replaced.During liquid addition. to the tank.Unit.l ODCM Revision 28 September 2006 1 3.1r5 RADIOACTIVE EFFLUENT INSTRUMENTATION -LIQUID D 3(4.6.14 NOTES FOR TABLE D 4.6.14-1 (a) The channel test shall also demonstrate that control, room alarm annunciation occurs if any of the following conditions exist: 1 Instrumentation indicates measured levels above the alarm setpoint.2. Instrument indicates a downscale failure.3. Instrument controls not set in operate mode.(b) The channel calibration shall be performed using one:or more reference standards certified by the National Institute of.Standards and Technology .(NIST), or using standards that are traceable to the NIST or using actual samples of liquid waste that have been analyzed on a system that has been calibrated with NIST traceable sources. These standards shall permit calibrating the system over its intended range of energy and measurement.(c) Sensor check shall consist of verifying indication of flow during periods of release. Sensor check shall be made at least once per 24 hours on days on which continuous, periodic or batch releases are made.(d) Pump performance curves or rated data may be used to estimate flow.(e) Tanks included in this specification are those outdoor tanks that are not surrounded by liners, dikes or walls capable of holding the tank contents.(f) Source check may consist of an installed check source, response to an external source, or (for liquid radwaste monitors) verification-within 30 minutes of commencing discharge of monitor response to effluent.Unit I ODCM Revision 28ý13.1-6 September 2006 RADIOACTIVE EFFLUENT INSTRUMENTATION -GASEOUS D 3/4.6.14 CONTROLS.SURVEILLANCE REQUIREMENT 1" b. Gaseous Process and Effluent The radioactive gaseous process. and effluent monitoring instrumentation channels shown in Table D 3.6.14-2,shall be functional, The Offgas process monitor alarm setpoint shall be set to ensure that the limits of Technical Specification 3.6.15 are not exceeded. The Effluent monitor alarm setpoints shall be set to ensure that the limits of Control DLCO 3.6.15.b.1 are not exceeded. The alarm setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in Part IL, With a radioactive gaseous process and effluent monitoring instrumentation channel alarm 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 nonfunctional, ,or change the setpoint so it is acceptably conservative. With less than the minimum number of radioactive gaseous process and effluent monitoring instrumentation channels functional, take the action shown in Table D 3.6.14-2. Restore the instruments to'functional status within 30 days or outline in the next Radioactive Effluent Release Report the cause of the nonfunctionality and how the instruments were or will be restored to functional status.b. Gaseous Process and Effluent Each radioactive gaseous process and effluent monitoring instrumentation channel shall be demonstrated functional by performance of the sensor check, source check, instrument channel calibration and instrument channel test operations at the frequencies shown in Table D 4.6.14-2.I Auditable records shall be maintained of the calculations made, in accordance with procedures in Part II, of radioactive.gaseous process and effluent monitoring instrumentation alarm setpoints. Setpoints and setpoint calculations shall be available for review to ensure that the limits of Control DLCO 3.6.15.b. 1are met.Unit 1 ODCM Revision28 September 2006 13.1.7 RADIOACTIVE EFFLUENT INSTRUMENTATION -GASEOUS D 314.6.14 TABLE D 3.6.14-2 RADIOACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION Minimum Instrument Channels Functional Applicability Action.Stack Effluent Monitoring

a. Noble Gas Activity Monitors (1) High Range 2 *(2) Low Range 1 *(a)b. Iodine Sampler Cartridge 1 *(i)c. Particulate Sampler Filter 1 *~(b)d. Sampler Flow Rate Measuring Device * (b)e. Stack Gas Flow Rate Measuring Device 1(c)2. Deleted (c), (ci)* At all times.*

• Note Deleted.Unit 1 ODCM Revision 28 13. 1-8 September 2006 RADIOACTIVE EFFLUENT INSTRUMENTATION.-

GASEOUS D 3X4.6.14 TABLE D 3.6.14-2 (cont'd)RADIOACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION Minimum Channels Functional Instrument Applicability Action 3. ,Condenser Air Ejector Process Monitor (Offgas System Recombiner Discharge)

a. Noble Gas Activity Monitor b. Offgas System Flow Rate Measuring Device c. Sampler Flow Rate Measuring Device 4. Emergency Condenser System Effluernt 2.1 (g)(c)(c)1 a. Noble Gas Activity Monitor 1 per vent: (h).During operation of the main condenser air ejector During power operating conditions and whenever the reactor coolant temperature is greater than 212°F except for hydrostatic testing with :the reactor notcritical.

Unit. 1 ODCM Revision 28 1 3.1-9 September 2006 RADIOACTIVE EFFLUENT INSTRUMENTATION -GASEOUS D 3/4.6.14 NOTES FOR TABLE D 3.6.14-2 (a) (1) With the number of channels functional 1 less than required by the minimum channels functional requirement, effluent releases via this pathway may continue provided: (a) The nonfunctional channel is placed in the tripped condition, OR (b) Vent and Purge valves are closed and administratively controlled, OR (c) Primary containment integrity is not irequired. (2) With the number of channels functional 2 less than required by the minimum channels functional requirement, effluent releases via this pathway may .continue. provided grab samples are taken once per 12 hours .and these samples are analyzed for gross activity within 24 hours.(b) With the number of channels functional less than required by the minimum channels functional requirements, effluent releases via this pathway may continue provided that samples are continuously collected with auxiliary sampling equipment starting within 8 hours of discovery in accordance with the requirements of Table D 4.6.15-2.(c) With the number of channels functional less than required by the minimum channels functional requirements, effluent releases via this pathway may continue provided the flow rate is estimated once per 8 hours.(d) Stack, gas flow rate may be estimated by exhaust fan..cperating configuration.(e) Deleted (f) Deleted (g) (1) With the number of channels functional 1 less than required by the minimum channels functional requirement, gases from the main condenser offgas treatment system maybe released provided: (a) The nonfunctional channel is placed in the tripped condition, OR (b) At least one Stack monitor is functional, Otherwise be in at least hot shutdown within 12 hours.Unit 1 ODCM.Revision 28 1 3.1-10 September 2006 RADIOACTIVE EFFLUENT INSTRUMENTATION -GASEOUS D 3/4.6.14 NOTES FOR TABLE D 3.6.14-2 (cont'd)(2) With the number of channels functional 2 less than required-by the minimum channels functional requirement, gases from the main condenser offgas treatment system may be released provided: (a) Offgas grab samples are collected and analyzed once per 1.2 hours, AND (b) At least one Stack monitor is functional, I*Otherwise be in at least hot shutdown within 12 hours.(h) With the number of channels functional less than required by. the minimum channels functional requirements, steam release via this pathway may commence or continue provided vent pipe radiation dose rates are monitored once per four hours.(i) With the number of channels functional less than required by the minimum channels functional requirement, effluent releases via this pathway may.continue provided grab samples are taken once per 12 hours and these samples are analyzed for gross activity within. 24 hours.I!Unit I ODCM Revision 28 September 2006 1.3.1-11 RADIOACTIVE EFFLUENT INSTRUMENTATION -GASEOUS D 3/4.6.14 TABLE D 4,6 14-2 RADIOACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION Surveillance Requirements Instrument

1.

• Stack Effluent Monitoring System a. Noble Gas Activity Monitors (High Range and Low Range)b. Iodine Sampler Cartridge c. Particulate Sampler Filter d. SamplerFlow Rate Measuring Device e. Stack Gas Flow Rate Measuring Device.2. Deleted 3. Condenser Air Ejector Process Monitor (Offgas System Recombiner Discharge):

a. Noble Gas Activity Monitor b. Offgas System Flow Rate Measuring Device c. Sampler Flow Rate Measuring' Device 4. Emergency Condenser System Effluent a. Noble Gas Activity Monitor Sensor Check Source Check Channel Test Channel Test Channel Calih-tion Once/day(a)

None None Once/day(ý) Once/day Once/day(t 9 Once/day(o Once/day(rO Once/day(h) Once/92 days None None None None Once/92 days None None Once/92 days Oncel 84 days(9)None None None None On.ce/24 months(c)None None Once184 days(g)Once/24 months(b)None None, Once/24 months I Once/24 monthsI one/24 monthsJb)Once/24 months I Oncef24 months Once/24 months(b)Unit 1 ODCM Revision 28 September 2006 13.1-12 RADIOACTIVE EFFLUENT INSTRUMENTATION -GASEOUS D 3/4,614 NOTES FOR TABLE D 4.6.14-2 (a) At all times.(b) The channel calibration shall be performed using one or more of the reference standards certified by the National Institute of Standards and Technology (NIST), standards .that are traceable: to the NIST or using actual samples of gaseous effluent that have; been analyzed on a system that has been calibrated with NIST traceable sources, These standards shall permit calibrating the system over its intended range of energy and measurement.(c) The channel function test.shall demonstrate that control room alarm annunciation occurs if either of the following conditions exist: 1) Instrument indicates. measured levels above the Hi or Hi Hi alarm setpoint.2) Instrument indicates a downscale failure.The channel function test shall also demonstrate that automatic, isolation of this pathway occurs if-either of the following conditions exist: 1) Instruments indicate two channels above Hi Hi alarm setpoint..

2) Instruments indicate one channel above Hi Hi alarm setpoint and one channel downscale.(d) Deleted (e) Deleted (f) During operation of the main condenser air ejector.(g) The channel testshall produce upscaleand downscale, annunciation., (h) During power operating conditions and whenever the reactor coolant temperature is greater than 212°F except for hydrostatic testing with the reactor not critical.Unit 1 ODCM Revision 28 I~3.-13 September 2006 RADIOACTIVE EFFLUENTS

-LIQUID CONCENTRATION D 3/4.6.15 CONTROLS SURVEILLANCE REQUIREMENT I.DLCO 3.6.15 RADIOACTIVE EFFLUENTS Applicability: Applies to the radioactive effluents from.the station.Objective: To assure that radioactive material is not released to the environment in any uncontrolled manner and is within the limits of 1OCFR20 and 1OCFR50 Appendix I.Specification:

a. Liquid (1) Concentration The concentration of radioactive material.

released in liquid effluents to unrestricted areas shall be limited to ten. times the concentrations specified in 1 OCFR Part 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 104 microcuries/ml total activity., Should the concentration of radioactive material released in liquid effluents to unrestricted areas exceed the above limits, restore the concentration to within the above limits immediately. DSR 4.6.15 RADIOACTIVE EFFLUENTS Applicability: Applies to the periodic test and recording requirements of the station process effluents. Oblective: To ascertain that radioactive effluents from the station are.within the allowable values of IOCFR20, Appendix B and 1OCFR50, Appendix I.Specification:

a. Liquid (1) Concentration Radioactive.

liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table D 4.6.15-1.The results of the radioactivity analyses shall be used in accordance with the methodology and parameters in Part II to assure that the concentrations at the point of release are maintained within the limits of Control DLCO 3.6.15.a.(1). Unit 1 ODCM Revision 29 September 2006 13.1-14 RADIOACTIVE EFFLUENTS. -LIQUID DOSE D 314.6.15 CONTROLS SURVEILLANCE REQUIREMENT 6 (2) Dose The dose or dose commitment to a member of the public from radioactive materialsin liquid effluents released, from each reactor unit, to unrestricted areas (see Figures 5.1-1) shall be lirrited: (a) During any calendar quarter to less than or equal to 1.5 mrems to the total 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 total body and to less than or equal to 10 mrems to any organ.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 Control. D 6.9.3, 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.(2) Dose 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 II monthly.Unit 1 ODCM Revision 28 September 2006 13.1715

RADIOACTIVE EFFLUENTS

-LIQUID D 3/4.6;15 TABLE D 4.6.15-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Surveillance Requirement Minimum Sampling Frequency Lower Limit(a) of Detection (LLD) (IfXirmI)Liquid Release Type Analysis Frequency Type of Activity Analysis A. Batch Waste(b) Tanks Each Batch Each Batch Principal Gamma(c)Emitters 5 x 161 1-131 Each Batch(d) Each BatchWd) Dissolved and Entrained 1 x 10-5 Gases (Gamma Emitters)* Monthly H-3 1 x.10"5 Each Batch Composite(e) Gross Alpha 1 x 10`7* Quarterly Sr-89, Sr-90 5 x 10.8 Each Batch Composite(8) Fe-55 1 x 10.6 Once/month(') Once/month(O Principal Gamma(c) Emitters 5 x 10.7 1-131 1 X.10-6 Dissolved and Entrained 1 x 10s Gases H-3 1 x 10;Gross Alpha 1 x 10-7 B., Service Water System Effluent Once/quarter(O) Once/quarter0 Sr-89, Sr-90 Fe-55 5X 10.8 1 x 10 6 k Completed prior to each release.Unit I ODCM Revision 28 September 2006 1 3.1-1.6 RADIOACTIVE EFFLUENTS -LIQUID D 314.6.15 NOTES FOR TABLE D 4.6.15-1 (a) 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 EiVe2.22 x 10 6 oY.exp (-XAt)Where: LLD is the "a priori" lower limit of detection as defined: above, as microcuries per unit mass or volume, Sb.is the standard deviation of the background counting rate. or of the counting rate of a blank sample as appropriatej 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 x 10 6 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield, when applicable, X is the radioactive decay constant for the particular radionuclide, and.At for plant effluents isthe elapsed time between the midpoint of sample collection and time of counting.Typical values of E, V, Y and At should be used in the calculation. It should be recognized that the LLD is defined as. a before the fact limit representing the capability of a measurement system and not as an after the fact for a particular measurement. Unit I ODCM Revision 28 S3.1-1i 7 September 2006 RADIOACTIVE EFFLUENTS -LIQUID D 314.6.15 NOTES FOR TABLE D 4.6.15-1 (bj A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated and then thoroughly mixed to assure representativesampling.(c) The principal gamma emitters for which the LLD Specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-1 34, 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 Radioactive Effluent Release Report.(d) If more than one batch is released in a calendar month, only one batch need be sampled and analyzed during that month.(e) 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.(0) If the alarm setpoint of the service water effluent monitor, as determined by the method presented in Part-II, is exceeded, the frequency of sampling shall be increased to daily until the condition no longer exists, Frequency of analysis shall be increased to daily for principal gamma emitters (including dissolved and entrained gases) and an incident composite for H-3, gross alpha, Sr-89, Sr-90 and Fe-55.Unit l ODCM Revision 28 1 3.14 18 September 2006 RADIOACTIVE EFFLUENTS -GASEOUS DOSE RATE D 3/4.6.15 CONTROLS CONTROLS SURVEII I AMM RlIIQ1rM=FMI

b. Gaseous (1) Dose Rate The dose rate due to radioactive materials released in gaseous effluents .from the site. to areas at or beyond the site boundary shall be limited to the following: (a) For noble gases: Less than or equal to 500 mrems/year to the whole body and less than or equal to 3000 mrems/year to the: skin, and.(b) For iodine-131, iodine-133, tritium and all radionuclides in particulate form with half lives greater than 8 days: Less than or equal to 1500 mrems/year to any organ.With the dose rate(s) exceeding the above limits, without delay restore the release rate to within the above limits(s).

b. Gaseous (1) Dose Rate The dose rate due to noble gases in gaseous effluents shall be determined to be within the limits of Control DLCO 3.6.15 in accordance with the: methodology and parameters in Part I1.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 limits.of Control DLCO 3.6.15 in accordance with methodology and parameters in Part II by obtaining representative.samples and performing analyses in accordance with the sampling and analysis program specified in Table D 4.6.15-2.Unit I ODCM Revision 28 September

.2006 13.1-19 RADIOACTIVE EFFLUENTS -GASEOUS DOSE D 3/4.6.15 CONTROLS SURVEILLANCE REQUIREMENT SURVEILLANCE REQUIREMENT (2) Air Dose The air dose due to noble gases released in gaseous effluents, from each reactor unit, to areas beyond the site boundary shall be limited to thefollowing: (a) During any calendar quarter: Less than or equal to 5 milliroentgen 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 milliroentgen for gamma radiation and less than or equal to 20 mrads for beta radiation. 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 Control D 6.9.3, 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 bein compliance with the above limits.(2) Air Dose Cumulative dose contributions for the current calendar quarter and current calendar year.for noble gases shall be determined monthly in accordance with the methodology and parameters in Part II.Unit I ODCM Rcvision 28 September.2006 1 3.1-20 RADIOACTIVE EFFLUENTS -GASEOUS DOSE D 3/4.6;.15 CONTROLS SURVEILLANCE REOUIRFMFNT SUVILNE EURMN (3) Tritium, lodines and Particulates 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, from each reactor unit, to areas beyond the site boundary 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.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 Control D 6.9.3, a Special Report that identifies the cause(s) for exceeding the limit: 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.(3) Tritium, lodines and Particulates 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 monthly in accordance with the methodology and parameters in Part I1.Unit I ODCM Revision 28 September 2006 1. 3. 1-2n1 RADIOACTIVE EFFLUENTS -GASEOUS D 314.6.15 TABLE D.4.6.15-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Surveillance Requirements Gaseous Release Type A, Containment Purge(b)Sampling Frequency Each Purge Grab Sample Minimum Analysis Frequency Prior to each release Each Purge Type of Activity Analysis Principal Gamma Emitters(c) Principal Gamma Emitters(c) B. Stack C. Stack Once/Month~ 8)Once./Month(h) Continuous ie)Continuous~e) CofltinuUOUS() Once/Month(d) Once/Month Once/WeekV 0 Charcoal Sample Once/WeekV 0 Particulate Sample OncelMonth Composite Particulate Sample H-3 Principal Gamma Emitters(C) H-3 1-1131 Principal Gamma Emitters(c) Gross alpha, Sr-89, Sr-90 Noble Gases, Gross Gamma or Principal Gamma :Emitters(c) Lower Limit(a).of Detection (LLD) (gCi/ml)i x ll 1 x ti0f 4 1 x 10.4*1 x 1 0-I ~X 10:12 1 x 10 1 1 I x:10-1 1 Continuous(e) Noble Gas Monitor 1 x 10-5()Unit 1 0DCM Revision 28 September 2006 i 3.1-22 RADIOACTIVE EFFLUENTS -GASEOUS 0 314.6.15 NOTES FOR TABLE D 4.8.16-2 (a) The LLD is defined in notation (a) of.Table D 4.6.15-1.(b) Purge is defined in Section 1.0.(c) The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-1 33, Xe-135 and Xe-138 for gaseous emissions and Mn-54, Fe-59, Go-58, Go-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, 1-131 and Ce-144 for particulate emissions. 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 Radioactive Effluent Release Report pursuant to Technical Specification 6.6.3, "Radioactive Effluent Release Report", and Control D 6.9.1.(d) Sampling and analysis shall also be performed following shutdown, startup or an increase on the recombiner discharge monitor of greater than 50 percent, factoring out increases due 'to changes. in thermal power level or dilution flow; or when the stack release rate is in excess of 1000 jiCi/second and steady-state gaseous release rate increases by 50 percent.(e) The sample flow rate and the stack flow:rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Controls DLCO 3.6,15.b.(1).(b) and DLCO 3.6,15,b,(3), (f) When the release rate is in excess of 1000 P Ci/sec and steady state gaseous release rate increases by 50 percent, the iodine and particulate collection device shall be removed and analyzed to determine the changes in iodine-131 ,and particulate release rate. The analysis shall be done daily following each change until it is shown that a pattern exists which can be used to predict the release rate; after which it may revert to weekly sampling frequency. When samples collected for 24 hours are analyzed, the corresponding LLD's may be increased by.a factor of 10.(g) When the continuous Noble Gas:Monitor is nonfunctional the.LLD. for noble gas gamma analysis shall be i x 104 OCi/cc.(h) Tritium grab samples shall be taken weekly from the station ventilation exhaust (stack) when fuel is offloaded until stable tritium release levels. can be demonstrated. Unit I ODCM Revision 28 13.1-23 September 2006 RADIOACTIVE EFFLUENTS -MAIN CONDENSER, URANIUM FUEL CYCLE D 314.6.15 SURVEILLANCE REQUIREMENT CONTROLS c. Deleted d. Uranium Fuel Cycle 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. Main Condenser The radioactivity rate of noble gases at the recombiner discharge.shall be continuously monitored in accordance. with Table D. 3.6.14-2.d. Uranium Fuel Cycle Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with. Controls DSR 4.6.15.a.(2), DSR 4.6.15.b.(2) and DSR 4.6.15.b;(3) and in accordance with the methodology and parameters in Part II.Unit I ODCM Revision 28 September 2006 13.1-24 RADIOACTIVE EFFLUENTS -URANIUM FUEL CYCLE D 314.6.15 CONTROLS SURVEILLANCE REQUIREMENT With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Controls DLCO 3.6.15.a(2), DLCO 3.6.15.b(2) and DLCO 136.15.b(3), calculations shall be made including direct radiation contributions from the reactor units and from outside storage tanks to determine whether the above listed 4OCFR190 limits have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to Control D 6.9.3, 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 Part 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.Cumulative dose contributions from direct radiation from the reactor units, and from radwaste storage tanks shall be determined in accordance with the methodology and parameters in Part I1. This requirement is applicable only under conditions set forth in Control DLCO 3.6.15.d.Unit 1 ODCM Revision 28 September 2006 13.1 -;25 RADIOACTIVE EFFLUENTS -.URANIUM FUEL CYLE D 314.6.15 CONTROLS SURVEILLANCE REQUIREMENT It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels orconcentrations. If the estimated dose(s) exceeds the above limits, and if the release condition resulting in Violation of 40CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40CFR 190. Submittal of the report is considered a timely request and a variance is granted until staff action on the request is complete.Unit 1 ODCM Revision 28 September 2006 13.1-26 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS 0 3/4.6.16 CONTROLS SURVEILLANCE REQUIREMENT 4.DLCO 3.6.16 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS Applicability: Applies to the operating status of the liquid and gaseous effluent, treatment systems.Objective: To assure functionality of the liquid and gaseous effluent treatment system.Specification:

a. Liquid The. liquid radwaste treatment system shall be .used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected dose due to the liquid effluent, from each unit, to the Unrestricted Areas would exceed 0.06 mrem to the total body or 0.2 mrem to any organ for any batch.b. Gaseous (1) The Gaseous Radwaste Treatment System shall be functional.

The Gaseous Radwaste Treatment System shall be.used to reduce radioactive materials in :gaseous waste prior to their discharge as necessary to meet the requirements of Control DLCO 3.6.151 DSR 4.6.16 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS Applicability: Applies to the surveillance requirements for the liquid and gaseous effluent treatment systems.Obiective: To verify functionality of the liquid and. gaseous effluent treatment system.Specification:

a. Liquid Doses due to liquid releases to unrestricted areas shall be projected prior to the release of each batch of liquid radioactive waste in accordance With the methodology and parameters in Part 11.b. Gaseous (1) Doses due to gaseous releases to areas at or beyond the site boundary shall be calculated in accordance with the methodology and parameters in Part II.I Unit I ODCM Revision 28 September 200-6 13.1-27 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS D 3/4.6.16 CONTROLS With gaseous radwaste from the main condenser air ejector system being discharged without treatment~for more than 7 days, prepare and submit to the Commission within 30 days, pursuant to Control D 6.9.3, a Special Report that identifies the nonfunctional equipment and the reason for its nonfunctionatity, actions taken to restore. the nonfunctional equipment to.functional status, and a summary description of those actions taken to prevent a recurrence.

(2) The Ventilation Exhaust Treatment System shall be functional and appropriate portions of this system shall be used to reduce releases of radioactivity when the projected doses in 31 :days due to gaseous effluent releases, from each unit, to areas beyond the site boundary would exceed 0.3 mrem to any organ of a member of the public.With radioactive gaseous waste being discharged without treatment and in excess of the above limit, complete a CR evaluation of the degraded condition within 30 days that identifies the nonfunctional equipment,. the reason for the nonfunctionality, and plans and.schedule to restore the equipment to functional status.(2)---- ----- ------NOTE: Only required to be met-when the Ventilation Exhaust Treatment System is not being fully utilized.Project the doses from the iodine and. particulate releases from each. unit.to areas beyond the Site Boundary at least every 31 days.Unit, 1 ODCM Revision 28 September 2006 13.1-28 MARK I CONTAINMENT D 3/4.6.18 CONTROLS SURVEILLANCE P1=Q1 IIP;:MPtdT SIJR'JFII I AMCF PFAI IIPFM~tdT* --. ---.--- -.~ ~ I DLCO 3.6.18 MARK I CONTAINMENT Applicability: Applies to the~venting/purgjng of the Marki Containment. Objective: To assure that the Mark I Containment is vented/purged so that the limits of Controls DLCO 3.6.15.b(1) and DLCO 3.6,15.b(3) are met.Specification: The Mark I Containment drywell shall be vented/ purged through the Emergency Ventilation System unless Controls DLCO 3.6.15.b.(1) and DLCO 3.6.15b.(3) can be met without use of the Emergency Ventilation System, If these requirements are not satisfied, suspend all venting/purging ofihe drywell.DSR 4.6.18 MARK I CONTAINMENT Applicability: Applies to the surveillance requirement for venting and purging of the Mark I Containment when required to be vented/purged through the Emergency Ventilation System.Obiective: To verify that the Mark I Containment is vented through the Emergency Ventilation System when required.Specification: The containment drywell shall be determined to be aligned for venting/pUrging through the Emergency Ventilation System within four hours prior to start of and at least once per 12 hours during ventinglpurging of the.drywell. Unit 1 ODCM Revision 28 September 2006 13.1-29 LIQUID WASTE HOLDUP TANKS D 3/4.6,19 SURVEILLANCE REQUIREMENT CONTROLS DLCO 3.6.19 LIQUID WASTE HOLDUP'TANKS* DSR 4.6.19 LIQUID WASTE HOLDUP TANKS;Applicability: Applies to the quantity of radioactive material that may be stored in an outdoor liquid waste holdup tank.Obiective: To assure that the quantity of radioactive material stored in outdoor holdup tanks does not exceed a specified level.Specification:.. The quantity of radioactive material contained in an outdoor liquid waste tank shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases.With the quantity of radioactive material in any such tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank. Within 48 hours reduce the~tank contents to Within the limit and describe the events leading to this condition in the next Radioactive Effluent Release Report.*Tanks included in this:Control are those outdoor tanks that are notsurrounded by liners, dikes, orwalls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system, Applicability: Applies to the surveillance requirements for outdoor liquid waste holdup tanks.Obiective: To verify the quantity of radioactive material stored in an outdoor liquid Waste holdup tank.Specification: The quantity.of radioactive material contained in each of the tanks listed in Control DLCO 3.6.19 shall be determined to be within the limit of Control OLCO 3.6.19 by analyzing a representative sample of the tank's contents at least weekly when radioactive materias are being added to the tank.Unit 1 ODCM Revision 28 September 2006 13.1-30 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 SURVEILLANCE REQUIREMENT CONTROLS*DLCO 3.6.20 RADIOLOGICAL ENVIRONMENTAL MONITORING DSR 4.6.20 RADIOLOGICAL ENVIRONMENTAL PROGRAM MONITORING PROGRAM Applicability: Applies to radiological samples of station environs.Objective: To evaluate the effects of station-operations and radioactive. effluent releases on the environs and to verify the effectiveness of the controls on radioactive material sources.Specification: The radiological environmental monitoring program shall be conducted as specified in Table D 3.6.20.1.With the radiological environmental monitoring program not being conducted as specified in Table D 3.6.20-1, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence. Deviations are permitted from the required sample schedule if samples are unobtainable due to hazardous conditions, seasonal unavailability, theft, uncooperative residents or to malfunction of automatic sampling equipment. In the event of the latter, every effort shall be made to complete corrective action prior to the end of the next sampling period.Applicability: Applies to the periodic sampling and monitoring requirements.of the radiological environmental monitoring program.Obiective: To ascertain what effect station operations and radioactive effluent releases have had upon the environment. Specification: The radiological environmental monitoring samples shall be collected pursuant to Table D.3.6.20-1 from the specific locations given in the table and figure(s) in Part II and shall be analyzed pursuant to the requirementsof Table D 3.6.20-1 and the detection capabilities required by Table D 4.6.20-~1.Unit 1. ODCM Revision 28 September 2006 13.1-31 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D.3/4.6.20 CONTROLS SURVEILLANCE REQUIREMENT With the level of radioactivity (as the result of plant effluents), in an environmental sampling medium.exceeding the reporting levels of Table D 6.9.3-1 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days from the end of the affected calendar quarter a Special Report pursuant'to Control D 6.9.3. The Special Report shall identify the cause(s) for exceeding the limit(s) and define the corrective action(s) 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 Controls DLCO 3.6.15.a.(2),. DLCO 3.6.15;b.(2) and DLCO 3.6.15.b.(3). When more than one of the radionuclides in Table D 6.9.3-1 are detected in the sampling medium, this report shall be submitted if: concentration (1) + concentration (2) +......... limit, level. (1) limit level (2)>1.0 When radionuclides other than. those in Table D 6.9.3-1 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of Controls DLCO 3.6.15.a.(2), DLCO 3.6.15.b,(2).and DLCO 3.6.15.b.(3). Unit I ODCM Revision 28 i 3.1-32 September 2006 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6,20 CONTROLS SURVEILLANCE REQUIREMENT 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.With milk or fruit and/or vegetables no longeravailable at one or more of the Sample locations specified in Table D 3.6.20-, identify locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days. The specific locations from which samples were unavailable may then be deleted from the monitoring program. Identify the cause of the.unavailability of samples and identify the new location(s) for obtaining replacement samples in the next Radioactive Effluent Release Report and also include in the report a revised figure(s) and table for Part II reflecting the new location(s). Unit I ODCM Revision 28 1 3.1-33 September 2006 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 TABLE D 3.6.2041 OPERATIONAL RADIOLOGICAL ENVIRONMENTAL MONITORING. PROGRAM Exposure Pathway andlor Samole Number.of Samplesla) and Locations Sampling and Collection Preauenev (a)Type of Analysis and ,F reauenev ('ýFrequnc Radioiodine &Particulates Samples from 5 locations:

1) 3.Samples from off-site locations in different sectors of the highest calculated site. average D/Q (based on all site licensed reactors)2) 1 sample from the vicinity of an established year round community having the highest. calculated site average D/Q (based on all site licensed reactors)3) 1 sample from a control location 10-17 miles distant and in .a least prevalent wind direction(d Continuous sampler operation with sample collection.

weekly or as. required by dust loading, whichever is more frequent Radioiodine Canisters analyze once/week for 1-131.Particulate Samplers Gross beta radioactivity following filter change, )composite (by location) for gamma isotopic analysis(d) once per 3 months,. (as a minimum)Gamma dose once. per 3 months Direct. Radiation(e) 32 stations with two or more: dosimeters to be placed as Once per 3 months follows; an inner ring of stations in the general area of the site boundary and an outer ring in the 4 to 5. mile range from the site with a station in each land based sector.* The balance of the stations should be placed in special interest areas such as population centers, nearby residences, schools and in 2 or 3 areas to serve as control stations.At this distance, 8 wind rose sectors are over Lake Ontario.Unit 1 DCM Revision 28 September 2006 1[3.1-34 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 TABLE D 3.6.20-1 (Cont)OPERATIONAL RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway and/or Sample Number of Samples(al and Locations Sampling and Collection Frequency (a)Type of Analysis and Freauncv ~Frenue .nrcv --.WATERBORNE Surface(0 1) 1 sample upstream Composite sample over I month period 1 P)2) 1 sample from the site's downstream cooling water intake I sample from a downstream area-with existing or potential recreational value Gamma isotopic analysis(c) once/month. Composite for once per 3 months tritium analysis.Gamma isotopic analysistc) Sediment from Shoreline Twice per year INGESTION Milk 1) Samples from milk sampling locations in 3 locations within 3.5 miles distance having the highest calculated site average D/Q. If there are none, then 1sample from milking animals in each of 3 areas 3.5-5.0 miles distant having the highest calculated site average D/Q (based on allsite licensed reactors)Twice per month, April-December (samples will be collected in January-March if 1-131 is detected in November and Decemberof the preceding year)Gamma isotopic(c) and 1-131 analysis twice per month when animals are on pasture (April-December); once/month at other times (January-March) if required 2) 1 sample from a milk sampling location at a control location (9-20 miles distant and in a least prevalent wind direction)(d) Unit I ODCM Revision 28 September. 2006 13.1-35 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 TABLE D 3.6.20-1 (Cont)OPERATIONAL RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway and/or Sample Number of Samples(a) and Locations Sampling and Collection Type of Analysis and F uen (3)nhlý tn ..Fish Food Products 1) 1 sample each of two commercially or recreationally important species in the vicinity of a plant discharge area.t 1)2) 1 sample each of the same species from an area at least 5 miles distant from the site.(d)1) Samples of three different kinds of broad leaf vegetation (such as vegetables) grown nearest to each of two different off-site locations of highest calculated site average D/Q (based.on all licensed site reactors).

2) Once sample of each of the similar broad leaf vegetation grown at least 9.3-20 miles distant in a least prevalent wind direction.

Twice per year Gamma isotopic analysis t C on edible portions twice per year.Once per year during harvest season Gamma isotopic(c) analysis of edible portions (isotopic to include 1-131 ora separate I-131 analysis may be performed) once during the harvest season Unit1 ODCM Revision 28 September 2006 13.1-36 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM*D 3/4.6.20 NOTES FOR TABLE D 3.6.20-1 (a) it is recognized that, at times, it may not be possible or practical 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 may be substituted. Actual locations (distance and directions) from the site shall be provided in the Annual Radiological Environmental Operating Report, Highest D/Q locations are based on historical meteorological data for all site licensed reactors.(b) Particulate sample filters should be analyzed for gross beta 24 hours or more after sampling to allow for radon and thoron daughter decay. If the gross beta activity in air is greater than 10 times a, historical yearly mean of control samples, gamma isotopic, analysis shall be performed on the individual samples.(c) Gamma isotopic analysis means the identification and quantification of gamma emitting radionuclides that may be attributable to the effluents from the facility.(d) The purpose of these samples is to obtain background information. If it is not. practical:to establish control locations in accordance with the distance and wind direction criteria, other sites, such as historical control locations which provide valid background data may be substituted.(e) One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purpose of this table, a thermoluminescent dosimeter may be considered to be one phosphor and two or more phosphors in a packet may be considered as two or more dosimeters. Film badges shall not 'be used for measuring direct radiation, (f) The "upstream sample" should betaken at a distance beyond significant influence of the discharge. The "downstream sample" should be taken. in an area beyond but near the mixing zone, if possible.(g) Composite samples should be collected with equipment.(or equivalent) which is capable of collecting an aliquot at time intervals which are very short (e.g. hourly) relativeto the compositing period (e.g. monthly) in order to assure obtaining a representative sample.(h) In the event commercialIor recreational important species are not available as a result of three attempts, then other species may be utilized as available. Unit I ODCM Revision 28 1.3.1-37 September 2006 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 TABLE D 4.6.20-1 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALySISRa,b) LOWER LIMIT OF DETECTION LLD'o)Surveillance Requirement Water(c) Airborne Particulate or Fish Milk Food Products Sediment Analysis (pCill) Gases (pCilm , (pCi/kg, wet) (pCi!/I) (pCilkg, wet),, (pCI/kg, dry)gross beta 4 0.01 H-3 2000*Mn-54 15 130 Fe-59 30 260 Co-58, Co-60 15 130 Zn-65 30 260 Zr-95, Nb-95 15 1-131 1* 0.07 1 60 Cs-134 15 0.05 130 15 60 150 Cs-137 18 0.06 150 18 80 180 Ba/La-140 15 15* If no drinking water pathway exists, a value. of 3000 pCi/litermay be used.*

• If no.drinking water pathway exists, a value of 15 pCi/liter may be used.Unit I ODCM Revision 28 1 3.1.-3 8 September 2006 RADIOLOGICAL.ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 NOTES FOR TABLE D 4.6.20.1 (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 pursuant to Technical Specification 6.6.2, "Annual Radiological Environmental Operating Report", and Control D 649.1.dA (b) Required detection capabilities for thermoluminescent dosimeters used for environmental measurements are given in ANSI N545 (1975), Section 4.3. Allowable exceptions to ANSI N.545 (1975), Section 4.3 are contained in Part il, Section (c) 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.Vý2,22eYvexp (-2At)Where: LLD is the "a priori" lower limit of detection as defined above, as picocuries per unit mass or volume, 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, where applicable, X is the radioactive decay constant for the particular radionuclide, and At for environmental. samples is the elapsed time between sample collection, or end of the sample collection period and time of counting.Typical values of E, V, Yand At should be used inthe calculation. Unit 10DCM Revision 28 13.1-39 September 2006 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 NOTES FOR TABLE D 4.6.20-1 It should be recognized that the LLD is defined as a before the fact limit representing the capability of a measurement system and not as an after the fact limit for the particular measurement. 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 LLDsunachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 6.6.2, "Annual Radiological Environmental Operating Report",,and Control D 6.9.1.d.Unit I ODCM Revision 28 1 3A,40 Septernber 2006 INTERLABORATORY COMPARISON PROGRAM D 3A4.6.21 CONTROLS SURVEILLANCE REQUIREMENT DLCO 3.6;21 INTERLABORATORY COMPARISON PROGRAM Applicability: Applies to participation in an interlaboratory comparison program on environmental sample analysis.Objective: To ensure the accuracy of measurements of radioactive material in environmental samples.Specification: Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which has been approved by the Commission. Participation in this program shall include media for Which environmental samples .are routinely collected and for which intercomparson samples are available. 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.DSR 4,6,21 INTERLABORATORY COMPARISON PROGRAM Applicability: Applies to testing the validity of measurements on environmental samples.Objective: To verify the accuracy of measurements on radioactive material in environmental samples.Specification: The Interlaboratory Comparison Program shall be described in Part 1!. 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. Participants in the EPA Cross Check Program may provide the EPA program.code designation. in. lieu of providing results, Unit I1.DCM Revision 28 Septcmber 2006 13.141 LAND USE CENSUS D 3/4.6.22 CONTROLS SURVEILLANCE REQUIREMENT SURVEILANCE.REU.R.MEN DLCO 3.6.22 LAND USE CENSUS*Avolicability: Applies to the performance of a land use census in the vicinity of the Nine Mile Point Nuclear Facility.Objective: To determine the utilization ofland within a distance of three miles from the Facility.Specification: A land use census shall be conducted and shall identify within a distance of three miles the location in each of the 16 meteorological sectors the nearest residence and within a distance of three miles the location in each of the 16 meteorological sectors of all milk'animals. In lieu of a.garden census, specifications for vegetation sampling in Table D.3.6.20-1 shall be followed, including analysis of appropriate controls.With a land use census identifying a milk animal location(s) that represents a calculated DIQ value greater than the D/Q value currently: being used in Control DSR.4.6.15.b.(3), identify the new location(s) in the next Radioactive Effluent Release Report.DSR 4.6,22 LAND USE CENSUS Applicability: Applies to assuring that current land use is known.Obiective: To verify the appropriateness of the environmental surveillance program.Specificatiorn: The land use census shall be conducted during the growing season at least once per 12 months using that information that will provide the best results, such as conducting a door-to-doorsurvey, aerial survey or consulting local agriculture authorities. The results of the land use census shall be included in the Annual Radiological Environmental Operating Report.Unit 10 DCM Revision 28 September 2006 13.1-42 LAND. USE CENSUS D 3/4.6.22 CONTROLS SURVEILLANCE REQUIREMENT I If the DIQ value at a new milk sampling location is significantly greater (50%) than the DIQ value at an existing milk.sampling location, add the new location to the radiological environmental monitoring program within 30 days. Thesampling location(s) excluding the control station location, having the.lowest calculated D/Q may be.deleted from this monitoring program after October 31 of the year in which this land use census was conducted. Pursuant to Control D 6.9.1 .e identify the new location(s) in the next Radioactive Effluent Release Report and also include in the report a revised figure(s)-and table for Part II reflecting the new location(s).,Unit I ODCM ReviSion 28 September.2006 1 3.1-43 PART I -RADIOLOGICAL EFFLUENT CONTROLS Bases Unit.1 ODCM Revision 28 B 3.1-0 September 2006 I RADIOACTIVE EFFLUENT INSTRUMENTATION, RADIOACTIVE EFFLUENTS -LIQUID CONCENTRATION B 314.6.14 BASES FOR DLCO 3.6.14 and DSR 4.6.14 RADIOACTIVE EFFLUENT INSTRUMENTATION The radioactive liquid and gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid and gaseous effluents during actual or potential releases of liquid and gaseous effluents. The alarm/trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in Part II to ensure that the alarm/trip will occur prior to exceeding the limits as described in Technical Specification 6.5.3, "Radioactive Effluent Controls Program".The alarm/trip setpoint for the Offgas process monitor is limited by Technical Specification 3.6.15. The Objective of that Specification is to assure radioactive material released is within the limits of 1OCFR20 and 10CFR50 Appendix I. By doing so, total body exposure to an individual atlthe exclusion area boundary will not exceed a very small fraction of the limits of 10 CFR 100 in the event this effluent is discharged directly without treatment. The Offgas Process Monitors and Stack Effluent Monitors are interdependent. The Stack Effluent Monitors provide Effluent Monitoring (which requires a. minimum of 1 Low Range and 1 High Range monitor) and Containment Purge and Vent Isolation (which requires 2.High Range monitors). When the Purge and Vent isolation capability is not required (Primary containment .not required OR Purge and Vent valves shut and clearance applied), only 1 High Range Monitor and 1 Low Range monitor are required to satisfy the monitoring function. When serving as back-up: to the Offgas Monitors (Table D 3.6.14-2 Note g), this function may be satisfied by a single Low Range or High Range monitor because all Stack monitors function in the region:of interest due to their design overlap.The functionality and. use of this instrumentation is consistent with the requirements of General Design Criteria 601 63 and 64 of Appendix A to 10CFR Part 50. The purpose oftanklevel indicat, ing 'devices is to assure the detection and control of leaks that if not controlled could potentially result in the transport of radioactive materials to unrestricted areas.Unit 1 ODCM Revision 28 I B ý3.11 September 2006 RADIOACTIVE EFFLUENT INSTRUMENTATION, RADIOACTIVE EFFLUENTS -LIQUID CONCENTRATION B 314.6.15 BASES FOR DLCO 3.6.15 AND DSR 4.6.15 RADIOACTIVE EFFLUENTS Liquid Concentration This control is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to unrestricted areas will be less than ten times the concentration levels specified in 1OCFR Part 20, Appendix B, Table 2, Column 2. 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 Il.A design objectives of Appendix 1, IOCFR Part 50, to a member of the public and (2) the limits of 10 CFR 20.1301(e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its effluent concentration in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs).Detailed discussion of the LLD and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A.,"Limits for Qualitative Detection and Quantitative Determination -Application to Radiochemistry," Anal. Chem, 40. 586-93 (1968), and Hartwell, J.K., "Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).Unitl ODCM Revision 28 LB 3A-2 September 2006 RADIOACTIVE EFFLUENTS -LIQUID DOSE B 314.6.15 BASES FOR DLCO 3.6.15 AND DSR 4.6.15 RADIOACTIVE EFFLUENTS Liquid Dose This control is provided to implement the requirements of Section II.A, lIl.A and IV.A of Appendix I, 10CFR Part 50. The.controls expressed as, quarter and annual limits are set at those values found in Section II.A. of Appendix I, in accordance with Section iV.A. The action statements provide the required operating flexibility and at the same time implement the guides set forth in Section IVIA 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." There are no drinking water supplies that can be potentially affected by plant operations. The dose calculation methodology and parameters in .Part II implement the requirements in Section III.A of Appendix I that conformance with the guides of-Appendix I be shown by calculation procedures based on models and data, such that the actual exposure of a member of the public through appropriate pathwvays is unlikely to be substantially underestimated. The equations specified in Part 11 for calculating the doses due to the actual release rates of radioactive materials in liquideffluents 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 10CFR 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.Unit I ODCM Revision 28 1133.1-3: September 2006 RADIOACTIVE EFFLUENTS -GASEOUS DOSE RATE B 3M4.6.15 BASES FOR DLCO 3.6.15 AND DSR 4.6.15 RADIOACTIVE EFFLUENTS Gaseous: Dose Rate This control is provided to ensure that the dose at any time at or beyond the site boundary from gaseous effluents from all units on the site will be within the annual dose limits of 10CFR Part 20 to unrestricted areas. The annual dose limits are the doses associated with the concentrations of 1 OCFR 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'10CFR Part 20 or as governed by 10 CFR 20.1302(c). For members of the public who may at times be within the site boundary, the occupancy of that 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. 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 total 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 mremslyear. The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the'lower limits of detection (LLDs).Detailed discussion of the LLD and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Curte, L A.,"Limits for Qualitative Detection and Quantitative Determination -Application to Radiochemistry," Anal. Chenm. 40, 586-93 (1968), and Hartwell, J.K., "Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-21 5 (June 1975).Unit j 0DCM Revision 28 1B 3.1-4 September 2006 RADIOACTIVE EFFLUENTS -DOSE -NOBLE GASES B 3/4.6.15 BASES FOR DLCO 3.6.15 AND DSR 4.6.15 RADIOACTIVE EFFLUENTS Dose -Noble Gases This control is provided to implement the requirements of Sections 11.B, III.A and IV.A of Appendix I, 1OCFR Part 50. The controls expressed as quarter and annual limits are set at those values found in Section 11.B of Appendix I in accordance with the guidance of Section IN.A. 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 gaseous effluents to unrestricted areas will be kept "as low as is reasonably achievable." The Surveillance Requirement .implements the requirements in Section ltlA of Appendix I that conform with the guides ofAppendix 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 Part II 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 .109, "Calculation of Annual. Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I,"Revision 1, October 1977 and Regulatory Guide 1.111 "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977.The Offsite Dose Calculation Manual Part I! equations provided to determine the air doses beyond the site boundary are based upon the historical average atmospheric conditions. Unit I ODCM Revision 28 lB 3.1-5 Septeniber 2006 RADIOACTIVE EFFLUENTS -DOSE- IODINE -131, IODINE -133, TRITIUM.AND RADIONUCLIDES IN PARTICULATE FORM B 3/4,6.15 BASES FOR DLCO 3.6.15 AND DSR 4.6.15 RADIOACTIVE EFFLUENTS Dose -Iodine-131, Iodine-133, Tritium and Radionuclides in Particulate Form This control is provided to implement the requirements of Sections 11.C, tlI.A and IV.A of Appendix I, 10CFR Part 50. The controls expressed as quarter and annual limits are set at those values found in Section II.C of Appendix I in accordance with the guidance of Section IV.A. The action statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to unrestricted areas will be kept "as low as -is reasonably achievable."' The Part II calculational methods specified in the Surveillance Requirement implements 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 Part II 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 1 OCFR 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 average atmospheric conditions. The release rate specifications for iodine-131, 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 beyond the site boundary. The pathways that were examined in the development of-these calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides ontogreen leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meatproducing animals graze with consumption of the milk and meat by man and 4) deposition on the ground with subsequent exposure of man.Unit I ODCM Revision 28 I B 3.1-6 September 2006 RADIOACTIVE EFFLUENTS -TOTAL: DOSE -URANIUM FUEL CYCLE B.3/4.6.15 BASES FOR DLCO 3.6.15 AND DSR 4.6.15 RADIOACTIVE EFFLUENTS Total'Dose -Uranium Fuel Cycle This control is provided to meet the dose limitations of 40CFR Part 190 that.have been incorporated into 10CFR Part 20 by 46FR 18525. The control requires the preparation and submittal of a Special Report whenever the calculated doses from plant generated radioactive effluents and direct radiation exceed 25 mrenis 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 4 reactors, it is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I and if direct radiation doses from the reactor units and outside storage tanks'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 40CFR Part 190 limits. For the purposes of the Special Report, it may be assumed.that the dose commitment to a member of the public from other uranium fuel cycle sources is negligible, with the exception that dose contribution from other nuclear fuel cycle facilities at the same site or 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 40CFR Part 190, the Special Report with a request for variance (provided the release conditions resulting in violation of 40CFR Part 190 have not already been corrected), in accordance with the provisions of 40CFR Part 190.11 and 10 CFR Part 20.2203(a)(4) is considered to be a timely request and fulfills the requirements of 40CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40CFR Part 190 and does not apply in any wayto the other requirements for dose limitation of 100FR.Part 20, as addressed in Controls DLCO 3.6.15.a.(I) and DLCO 3.6.15.b.(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.Unit I ODCM Revision.28.J B 3.1-7 September 2006 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS -LIQUID AND GASEOUS B 314.6.16 BASES FOR DLCO 3t6.16 AND DSR 4.6.16 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS Liquid Radwaste Treatment System The requirement that the appropriate portions of this system be used provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This control implements the requirements of 10 CFR Part 50.36a, General Design Criterion'60 of Appendix A to 10 CFR 50 and the design objective given in Section 11.0 of Appendix .lto 10 CFR Part.50. Projected doses are calculated on a batch rather than every 31 days due to the low frequency of releases.Gaseous Effluent Treatment Systems The functionality 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 priorto release to the environment. The requirement that appropriate portionsof these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in .gaseous effluents will be kept t as low as is reasonably achievable.' This control implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60,of Appendix A to 10 CFR 50 and the..design objectives given in Section IL.D of Appendix I to 10 CFR Part 50. The control governing the use of appropriate portions of the Gaseous Radwaste Treatment System is based on time without treatment rather than dose, due to the wide variability in effluent with changing power conditions, :Since the capability exists to operate within specification without use of the Gaseous Radwaste Treatment System, it is conceivable that.due to unforeseen circumstances, limited operation without the system may be made sometime during the life of the plant. The control governing the use of appropriate portions of the Ventilation Exhaust Treatment System was specified as a suitable fraction of the dose design objectives set forth in IL.C of Appendix I, 10CFR Part 50, for gaseous effluents. Unit 1 ODCM Revision 28 LB 3.1-8 September 2006 MARK I CONTAINMENT, LIQUID HOLDUP TANKS B 3/4.6.18, B 3/4,6.19 BASES FOR DLCO 3.6.18 AND DSR 4.6.18 MARK I CONTAINMENT This control provides reasonable assurance that releases from drywell purging operations will not exceed the annual dose limits of 10CFR Part 20 for unrestricted areas.BASES FOR DLCO 3.6.19 AND DSR 4.6.19 LIQUID HOLDUP TANKS This control applies to any outdoor tank that is not surrounded by liners, dikes or walls capable of holding the tank contents and that does not have tank overflows and surrounding areas drains connected to the liquid radwaste treatment system.Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than ten times the concentrations of 1OCFR Part 20, Appendix B, Table 2,. Column 2, at the nearest potable water supply and the nearest surface water supply in. an unrestricted area.Unit 1 ODCM Revision 28 I r3 3.1-9 September'.2006 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM B 3/4.6.20.BASES FOR DLCO 3.6.20 AND DSR 4.6.20 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM The radiological environmental monitoring program required by this control provides representative measurements of radiation and of radioactive materialsin those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of members of the public resulting from the station operation. This monitoring program implements Section IV.B.2 of Appendix Ito 1 OCFR Part 50 and. thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials-and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring. The initially specified monitoring program will be effective forat least the first three years of commercial operation. Following this period, program changes may be initiated based on operational experience. The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs required by Table D 4.6.20-1 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as a before the fact limit representing the capability of a measurement system and not as an after the fact limit for a particular measurement. Detailed discussion of the LLD, and other, detection limits, can be found in HASIL Procedures Manual, HASL-300 .(revised annually), Currie, L.A.,"Limits for Qualitative Detection and Quantitative Determination -Application to Radiochemistry," Anal. Chem 40, 586-93 (1968) and Hartwell, J.K.,"Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-21 5 (June 1975)., Unit 1 ODCM Revision 28 I B 3.1-10 September 2006 INTERLABORATORY COMPARISON PROGRAM B 314.6.21 BASES FOR DLCO 3.6.21 AND DSR 4.6.21 INTERLABORATORY COMPARISON PROGRAM 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 for the purposes of Section IV.B.2 of Appendix I to 10OCFR Part 50.Unit 1 ODCM Revision 28 1B, 3.1-11 September 2006 LAND USE CENSUS B 3/4.6.22 BASES FOR DLCO 3.6.22 AND DSR 4.6.22 LAND USE CENSUS This control is provided to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the radiological environmental monitoring program are made if required by the results of this census, The best survey information such as from a door-to-door survey(s), from an aerial survey or from consulting with local agricultural authorities shallbe used. This census satisfies the requirements of Section IV.B.3 of Appendixl to 10CFR Part 50.In lieu of a garden census, the significance of the exposure via the garden pathway can be evaluated by the sampling of vegetation as specified in Table D 3.6,20-1.A milk sampling location, as defined in Section I, requires that at least 10 milking cows are present at a designated milk sample location. It has been found from past experience, and as a result of conferring with local farmers, that a minimum of 10 milking cows is necessary to guarantee an adequate supply of milk twice per month for analytical purposes. Locations with less than 10 milking cows are usually: utilized for breeding purposes which eliminates a stable supply of milk for samples as a result of suckling calves and periods when the adult animals are dry.Unit 1 ODCM Revision 28 I B 3.11-12 September2006 PART I -RADIOLOGICAL EFFLUENT CONTROLS Section 6.0 Administrative Controls Administrative COntrOls Unit 1 ODCM Revision 28 September 2006 I 6.0-0 Administrative Controls 6.0 6.0 ADMINISTRATIVE CONTROLS The ODCM Specificationslare subject to Technical Specification.Section 6.6.2, "Annual Radiological Environmental Operating Report," Section 6.6.3, "Radioactive Effluent Release Report," Section 6.5.1, "Offsite Dose Calculation Manual (ODCM)," and Section 6.5.3,"Radioactive Effluent Controls Program;" Unit I ODCM Revision 28 September 2006 ,16.0-1 REPORTING REQUIREMENTS D 6&9.1.d D 6.9 Reporting Requirements D 6.9.1.d Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report shall include a comparison with operational controls as appropriate, and with environmental surveillance reports from the previous 5 years, and an assessment of the observed i.mpactsof the plant operation on the environment. The report shall also include the results of land use censuses required by Control DLCO 3.6.22.The. report 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, required by Control DLCO 3.6.21; discussion of all deviations from the sampling schedule of Table D 3.6.20-1; and discussion of all analyses in which the LLD required in Table D 4.6.20-1 was not achievable;

• • One map shall cover stations near the site boundary; a second shall include the more distant stations.D 6.91.e Radioactive Effluent Release Report The Radioactive Effluent Release Report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste releases from the unit 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.The Radioactive Effluent Release Report shall include an annual summary of hourly meteorological data collected over the7 previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability.*

This same report shall include an assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary (Part It Figure 5.1.3-1) during the reporting period. All assumptions used in making these assessments, i.e., specific activity,. exposure time and location, shall be included in. these reports.The assessment of radiation doses shall be performed in accordance with'the methodology and parameters in Part I!L The 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 doses from liquid and gaseous effluents are given in Part II.* In lieu of submission with the 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.Unit I ODCM Revision.28 1 6.0-2 September 2006 REPORTING REQUIREMENTS D 6.9.1.e The Radioactive Effluent Release Reports: shall include the following information for each class of solid waste (as defined by 10 CFR Part 61)shipped offsite during the report period.a. Container volume, b. Total curie quantity (specify whether determined by measurement or estimate), c, Principal radionuclides (specify whether determined by measurement or estimate), d. Source of waste and processing employed (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms), e. Type of container (e.g., LSA, Type A, Type B, Large Quantity),, and,.f. Solidification agent or absorbent (e.g., cement)The Radioactive Effluent Release Reports shall include any changes made during the reporting period to the Process Control Program (PCP)and to the Offsite Dose Calculation Manual (ODCM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Control DLCO 3.6.22.Changes to the Process Control Program (PCP).shall be reported to the Commission in the Radioactive Effluent Release Report for the period in which the change(s) was made. This submittal shall contain: a. Sufficiently detailed information to totally support the rationale .for the change without benefit -of additional or supplemental information; b; A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes;and c. Documentation of the fact that'the change has been reviewed and found acceptable. Changes to the Offsite Dose Calculation Manual (ODCM) shall be in accordance with Technical Specification 6.5.1, "Offsite Dose Calculation Manual (ODCM)".Unit ODCM Revision 28 1 6.0-3 September 20.06 SPECIAL REPORTS D 6.9.3 D 6.9.3Special Reports Special reports shall be submitted in accordance with 10 CFR 50.4to the.Regional Office within the time:period specified for each report.These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable. reference specification:

a. 4 b;.d. Not. applicable to RETS e.f.h,. C h. Calculate Dose from Liquid Effluent in Excess of.Limits, Control DLCO 3.6,15.a(2)

(03 days from the end of the affected calendar-quarter).

i. Calculate Air.Dose from Noble Gases Effluent.in Excess of, Limits, Control DLCO 3.6..15,b1(2).

(30 days from the end of the affected. calendar quarter).j. Calculate Dose from 10131, H-3 and Radioactive Particulates with half lives greater than eight days in Excess of Limits, Control DLCO 3.6.1 5.b(3)(b) (30 days from the end of the affected calendar quarter).k. Caculated Doses from Uranium Fuel Cycle Source in Excess of Limits, Control DLCO 3.6.15.d (30 days from the end of the affected calendar year).I. Nonfunctional Gaseous Radwaste Treatment System, Control DLCO :3.6.16.b (30. days from the end of the affected calendar year).m. Environmental Radioloqical Reports. With the level of radioactivity (as the result of plant effluents) in an environmental sampling media exceeding the reporting level of Table D 6.9.3-1, when averaged over any calendar quarter, in lieu of.a Licensee Event Report,. prepare and submit to the Commission within thirty (30) days from the end of the calendar quarter a special report identifying the cause(s) for exceeding the limits, and define the corrective action to be taken.ýUnit I ODCM Revision 28 16.0-4 September 2006 SPECIAL REPORTS D 6.9.3 Table D 6.9.3-1 REPORTING LEVEL FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES REPORTING LEVELS Water (pCi/I)Airborne Particulate. or Gases (oCi/m 3)Fish Milk Food Products Analysis ICi/ko. wetl I Cifik H-3 Mn-54 Fe-59 Co-58 Co-60 Zn-65 Zr-95, Nb-95 1-131 Cs-134 Cs-137 Ba/La-140 20i000" 1,000 400 1,000 300 300 400 2**30 50 200 30,000 10,000 30,000 10,000 20,000 0.9 10.0.20.0 1,000 2,0o0 3 60 70 itnn~100 1,0600 2,000 For drinking water samples, This is a 40.CFR 141 value. If no drinking water pathway exists, a value of 30,000 pCi/liter may be. used.** If no drinking water pathway exists, a value of 20 pCi/liter may be used.Unit 1 ODCM Revision 28 September 2006 1 6.0-5 PART 1I -CALCULATIONAL METHODOLOGIES Unit I ODCM Revision 28 September 2006 II 1

1.0 LIQUID

EFFLUENTS.1A Setpoint Determinations 11.1 Basis Monitor setpoints~will be established such:that. the concentration of radionuclides in the liquid effluent releases in the discharge.canal. shall be limited to ten times the concentrations specified in 10 CFR 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 2E-04 VtCilml total activity. S etpoints for the Service: Water System Effluent Line will be calculated quarterly based on the radionuclides identifiedduring the previous year's releases from the liquid radwaste system .or the.isotopes identified in the most.recent radwaste release or other identified probable source. Setpoints for the Liquid Radwaste Effluent Line will be based on ýthe radionuclides identified in each batch of liquid waste prior to its release.Afterrelease, the Liquid Radwaste -monitor. setpoint may remainas set, or revert back to a setpoint based on A previous Radioactive Effluent Release Report, or installblank flange in the discharge line and declare nonfunctional in accordance with the ODCM Part I.Since the ServiceWater System effluent monitor and Liquid Radwaste effluent monitor can only detect gamma radiation, the alarm setpoints arc calculated by using the concentration of gamma emitting isotopes only (or the corresponding Maximum Effluent Concentration (MEI) values for the same isotopes, whicheverare higher) in the j i(gCi/ml)iy expressionm(Section 1.1.2, 11.3).The RequiredDilution Factor (RDF) is calculated using concentrations of all isotopes present (or the corresponding MEC values for the same:isotqpes, whichever are higher) including tritium and other non-gamma emitters to ensure that all radionuclides in the discharge canal do not exceed Technical Specifications Radioactive Effluent Controls Program limits.11.2 Service Water System Effluent.Line Alarm Setpoint The detailed methodS for establishing setpoints for the Service Water System Effluent Line Monitor shall be contained in the Nine Mile Point.Station.Procedures. These methods shall'be in accordance with the following: The General Setpoint Equation is Setpoint < (Conservative Factor) (Concentration)(ADF)(CF) RDF From the above General Setpoint Equation the Hi and Alert alarms are calculated as follows: Setpint (Hi.alam)<0.9 (Ci / ml) E (CF)TEDF / Fn9 + background/ ml),irC/)MEDF F Setpoint (Alert. alarm) < 0.7 Ei.(Ci / m) (CF) TEDF F, + background ,f C n 1) / MEC .(j.tCi/mi)i= concentration of gamma emitting isotope i in the sample, or the corresponding MEC of gamma emittinig isotope i (MEC)i, whichever is higher (units = liCi/ml).Unit I ODCM Revision 28 11 2 September 2006

1.1.2 Service

Water System Effluent Line Alarm Setpoint (Cont'd)(GCi/ml),iT concentration of any radioactive isotope i in the sample including tritium and other non-gamma emitters or corresponding MEC of.isotope:i, MECj,. whichever is higher (units = utCi/m I).TF = Tempering Fraction TDF Total Dilution Flow (units = gallons/ninute). TEDF = Total Effective. Dilution. Flow = TDF (1-T.F) (units = gallons/minute) Ft, = Service Water Flow (units gallons/rninute). CF = Monitor calibration factor (units = net cpm!iCi/ml). MECi -Maximum Effluent Concentration, ten times the Effluent Concentration for radionuclide i as specified in 10 CFR 20, Appendix B, Table 2, Column 2 (unitsi).Sample = Those nuclides present in the previous batch release from the liquid radwasteeffluent system or thosenuclides present in the Iast Radioactive Effluent Release. Report (units.tCi/ml) or those nuclides present in the. service water system.*"= same as MEC, but for gamma emitting nuclides only.0.9 and 0.7= factors of conservatism to account for inaccuracies. RDF Required Dilution Factor, X i [(gtCi/ml) ,-/MECi]. If MEC values are used in the (QtCi/ml), 7 , they must also be usedin calculating RDF (numerator). RDF= FMEC (See Section! 14.2).ADF -Actual Dilution Factor, TEDF/FSW** For.periods with known reactor water to Reactor Building Closed Loop. Cooling (RBCLC) system leakage, RBCLC concentration may be prudently substituted for the above.1.1.3 Liquid Radwaste Effluent Line. Alarm Sctpoint.The detailed methods fbr establishing setpoints for the Liquid Radwaste Effluent Line Monitor shall be contained in the Nine Mile Point Station Procedures. These methods shall be in accordance with the following: The General Setpoint Equation in Section 11-1..1.2 is used to developthe Hi-Hi and Hi alarm setpoints, below:.zj ('uCi / ml~jý (CF)TEDF / ýF,, akgon Setpoint (Hi-Hi alarm) < 0.9 + background 7-i [(6Ci / 1ml) ir / MEC j ]Unit 1 ODCM.Revision 28 113 September 2006

1.1.3 Liquid

Radwaste Effluent Line.Alarm. Setpoint (Cont'd)Setpoint (Hi alarm) <.02 0. (YuCi / m)j(CF)TEDF /F -I-backgroundF .. fP(Ci / mO iT / MEC j](gCi/ml)= concentration of gammaemitting isotope iVin the sample or the corresponding MEC of gamma emitting isotope i, (MEC)1 whichever is higher.(htCi/ml)jT = concentration of any radioactive isotope i in thesample includingtritium and other non.gamma emitters or the corresponding MEC of isotope i, MECQ, whichever is higher. (units = giCi/ml).TF = Tempering. FractiOn TDF = Total Dilution Flow (units= gallons/minute). TEDF = Total Effective Dilution Flow = TDF (1-TF). (units = gallons/minute) FR; = Radwaste Effluent Flow (units.= gallons/minute). CF = Monitor calibration factor (units = net cps/RCi/mi). MECG Maximum'Effluent Concentration, ten times the Effluent Concentration for radionuclide i .as specified in .10 CFR 20, Appendix B, Table 2, Column 2, for those nuclides detected. by spectral analysis of the contents of the radwaste tanks to be released. (units = ýCi/ml)(MEC)j = same as.MECG but for gamma emitting nuclide only..0.9 and 0.7 = factors-of conservatism to-account for inaccuracies.. RDF = Required Dilution Factor, Z [(p.iCUml) iT/MECiG]. If MEC values are used in the they must also be used in calculating RDF (numerator). ADF Actual Dilution Factor Notes: (a) If TEDF/Fr = E [(j.Ci/ml)iT/MEGc] (if ADE= RDF)the discharge could not be made, sincethe monitor would be continuously in alarm, To avoid this situation, F,. will be reduced (normally by a factor of 2). to allow setting the alarm point at a concentration higher than rtank.concentration... This will also result in a discharge canal concentration at approximately 50%Maximum Effluent Concentration.(b) TF is tempering fraction (i.e., diversion of somc fraction of dischargc flow to the intake canal for the purpose of temperature control).Unit 1 ODCM Revision 28 H1 4 September 2006

1.1.4 Discussion

1L1.4.1 Control of Liquid Effluent Batch Discharges At Nine Mile Point Unit I Liquid RadwasteEffluents are released only on a batch mode. To prevent the inadvertent. release of any liquid radwaste effluents, radwaste discharge is mechanically isolated (hlank.flange installed or discharge valve chain-locked closed).following the completion of a batch release or series of batch, releases.This mechanical isolation remains in place and will only be removed prior to the next series of liquid radwaste discharges after all analyses required in station procedures and Table D 4.6.1 5-A of Part I are performed and monitor setpoints have been properly adjusted.1.1.4.2 Simultaneous Discharges of Radioactive Liquids If during the discharge of any liquid radwaste batch, there is an indication that the'service water canal has become contaminated (through a service water monitor alarm or through a grab sample analysis in the event that the service water monitor is nonfunctional) the discharge shall be~terminated immediately. The liquid radwaste discharge shall notbe continued until the cause of the servicewater alarm (or high grab sample analysis result) has been determined and the appropriate corrective measures taken to ensure ten times the effluent concentrations specified in IOCFR20, Appendix B, Table 2, Column 2 (Section D 3.6. :5.a(l) of Part .) are not exceeded. In accordance with Liquid Waste procedures,, controls are in place to preclude a simultaneous release.of liquid radwaste batch.taiks. In addition, an independent verification of the discharge valve line-up is performed

prior to discharge to .ensure that simultaneous dischargesare prevented.

1.1.4.3 Sampling Representativeness This sectioncovers Part I Table D 4.6.15-1 Note b concerning thoroughly mixing of each batch of liquid radwaste prior to sampling.Liquid Radwaste Tanks scheduled for discharge at Nine Mile Point Unit I are isolated (ixe.inlet valves marked up) and at leasttwo tankvolumes-of entrained fluids are recirculated prior to sampling. Minimum recirculation time is calculated as follows: Minimum Recirculation Time = 2.O(T/R)Where: 2.0 = Plant established.mixing factor, unitless T= Tank volume, gal R = Recirculation flow rate, gpm Additionally, the Hi Alarm setpoint of the Liquid Radwaste Effluent Radiation Monitor is set at a value corresponding to not more than 70% of its calculated response to the. grab sample or corresponding.MEC values. Thus, this radiation monitor will alarm if'the grab sample, or correspondingMEC value, is significantly lower in activity than any part of the tank contents being discharged. Unit I ODCM Revision 28]I 5 September,.2006 1.1..4.4 Liquid Radwaste System Operation Part I Section DLCO 316.16.a requires thatfthe liquid radwaste system shalibe used to reduce the:radioactive materials in liquid wastes prior to their discharge, as necessary, to meet the concentration and dose requirements of Section DLCO 3.6;i 5.Utilization of the radwaste system will be based ýon the capability of the indicated components of each process system to process contents of the respective low conductivity and high conductivity collection tanks: 1) Low Conductivity (Equipment Drains): Radwaste Filter andlRadwaste Demin.(See Fig. D!) or modulir waste water technology ("T-ERMEX")

2) High Conductivity

'(Floor Drains):. Waste Evaporator (See Fig. D-1) or modular waste water technology. (*THERMEX") directly to the Waste Collector Tank or the Waste Sample Tanks.Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be deternilined as described in Section 11-1.3 of this manual prior to the release of each batch of liquid waste. This same.dose projection of-Section 11-1.3 will also be performed in the event that untreated liquid waste is discharged, to ensure that the dose limits of Part I Dl ,CO 3.6.15.a(2) are not exceeded. (Therehy:implementing the.requirements of IOCFRS0.36a, General Design Criteria 60 of Appendix A and the Design Objective given in Section 1I-D of Appendix I ýtolO0 CFR50).For the purpose of dose projection, the following assumptions shall be made with regard to concentrations of non-gamma emitting radionuclides subsequently analyzed:-a) [H-3] < H-3 Concentration found recent condensate storage tank analysis b) [Sr-89] !5 4 xCs-137 Concentration c) [Sr-90] < 0.5 x Cs-137 Concentration d) [Fe-55] I i x Co-60 Concentration Assumed Scaling Factors used in b,. c, and d above represent conservative estimates derived from analysis of historical data from process waste streams. Following receipt of H-3, Sr-89, Sr-90 and~e-55 analysis information, dose estimates shall be revised using actual:.radionuclide concentrations and actual tank volumcs discharged. Unit I ODCM Revision 28 1I 6 September 2006 1.1.4.5 Service Water System Contamination Service water is normally noni-radioactive. If contamination is suspected, as indicated by a significant increase in service water effluent monitor response, grab samples will be obtained from the service water discharge lines and a gamma isotopic analysis meeting the LLD requirements of Part I Table D 4.6.1.5-1. completed. If it is determined that an inadvertent radioactive discharge is occurring from the service.water system, then: a) A lOCFR 50.59 review shall be performed (ef. M&E Bulletin 80-10), b) daily service water effluent, samples shall be taken and analyzed for principal gamma emitters until the release is terminated,:c) an incident composite shall be prepared for H-3, gross alpha, Sr-89, Sr-90 and Fe-55 analyses and, d) dose projections shall be performed in accordance with Section.Il-1.3of this manual (using estimated concentrations for H-3, Sr-89, Sr-90 and Fe-55 to be conservatively determined by supervision at the time of theincident). Additionally, service water effluent monitor setpoints may be recalculated using the actual distribution of isotopes found fromn sample analysis.When contamination is indicated by quantitative-non-gamma emitter results, sample and analyze gamma and. non-gamma emitters weekly.1.2 Liquid Effluent ConcentratiouCaleculation This calculation documents compliance with Part I Section DLCO 3.6.15.a (1).The concentration of radioactive material released in liquid effluents to unrestricted areas (see Figure 5.1.3-1) shall be limited to ten times the effluent concentrations specified in I OCFR20, 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 E-4 microcurie/milliliter (ltCi/ml)total activity at the point of discharge. For dissolved and entrained noble gases, this limit may also be satisfied by using 2E-4 4Ci/ml as the, MEC for each noble gas.The concentration of radioactivity from Liquid Radwaste batch releases and, if applicable. Service Water System and emergency condenserstart-up vcnt discharges are included inthe calculation. The calculation is performed for a specific period of time. No credit taken for averaging. The limiting concentration is calculated as follows: FMEC = i [(X S'CiY) / (MECi E Fj) ]Unit I ODCM Revision 28 1.17 September 2006 1.2.Liquid Effluent Concentration 'Calculation: (Contrd)Where: FMEC The fraction of Ma:cimum Effluent Concentration, the.raiio at the point ofdischargeof the actual concentration, to ten times the Effluent.Concentration of 10 CFR 20, Appendix B,.Table 2, Column: 2 for radionuclides other than dissolved or entrained noble gases.For noble gases, theconcentration shall be .limited io 2 E-4.microcurie/ml total activity.Ciý-(p.Ci/ml) 5 = The concentration ofnuclide.i. in particular effluent stream s,, p.Ci/m!.F, = The flow rate of a particular effluent-stream s, gpm.MECi = Maximum Effluent Concentration,.ten times the Effluent Concentration of a specific nuclide i from I OCFR20, Appendix B, Table 2, Column 2 (noble gas limit is 2E-4 pCi/ml).(CiFý) The total activity rate.of nuclide i, in all effluent streams s.(F 5) The total flow rate of all effluent..streams s, gpm (including those streams which do.not contain radioactivity). A value of less than. one for EMEC is considered acceptable forcompliance with Part I Section DLCO 3.6.15.a.(1). 1.3 Dose Determinations

1.3.1 Maximum

Dose Equivalent Pathway A dose assessment report was prepared for. the Nine Mile-Point Unit I facility by Charles T.Main, Inc., of Boston, MA. This report: presented the calculated dose equivalent rates t-o individuals as well.as the population within a 50-mile radius uf the facility based on the.radionuclides released in liquid, and gaseous effluents during the time periods of 1 July 41980 through 3.1 December 1980 and from January 1981 through 31 December. 1981. The.radwaste liquid releases are based on a canal discharge rate of 590 ft 3/sec which affects near field and far field dilution; therefore, this report'is specific to this situation. Utilizing the effluent data contained in the Semi-Annual Radioactive Effluent Release Reports as source terms, dose equivalent rates were determined using the environmental pathway models specified in Regulatory Guides 1.109 and:.l, 1I11 as incorporated in the NRC computer codes LADTAP for liquid pathways, and XOQDOQ and GASPAR for gaseous effluent pathways, Dose equivalentrates were calculated for the total body as well as seven organs and/or tissuesfor-the adult, teen, child, and infant age groups. From the standpoint of liquid effluents, the pathways evaluated included fish' and'drinking water ingestion, and external exposure to water and sedim ent.Unit I ODCM Revision. 28 118 September 2006

1.3.1 Maximum

Dose Equivalent Pathway (Cont'd)The majority of the dose. for a radwaste liquid batch release was received via the fish pathway. However, to comply with Part I Specifications for dose projections, the drinking water and. sediment pathways. are included. Therefore,-aHl doses.due to liquid effluents are calculated monthly for the fish and drinking water ingestion pathways and the sediment external pathway from all detected nuclides in liquid, effluents released to the unrestricted areas to each organ. The.dose projection for liquid batch releases will also include discharges from the emergency condenser vent as applicable, for all pathways. Each age group dose factor, Aiat is given in Tables 2-1 to 2-8. To expedite time, the dose is calculated for a maximum individual instead of each age group. This maximum indiV idual will be a composite of the highest dose factor of each age group for each organ, hence Aij.The' fo1l1wingexpression from NUJREG 0133, Section 4.3 is used to calculate dose: A i [Ail L(ATLCLFL)] Where: Di = The cumuilative.dose commitmentfto the total body or any organ, from the liquid effluents for the total time period"(ATL), troem.ATL = The length of the L th time period over which CiL and FL are averaged for all liquid releases, hours.CJL The average concentration of radionuclide, i, in undiluted liquid effluents. during time period ATL from any liquid release, ,tCi/ml.A 1 r The site related ingestion dose commitment factor to the total body or any organtifoir each identified principal gamma or beta emitter fora maximum individual, mrem/hr per ptCi/ml.FL The near field average dilution factor for CGL during any liquid effluent release.Definedas~the ratio of the maximum undiluted liquid waste.flow during release to the average flow from the. site discharge structure to unrestricted receiving waters, unitless.Aia, values for radwaste liquid batch releases at a discharge rate of 295 ft 3/sec (one circulating water pump in operation) are presented in tables 2-I to 2-4. Ai,, values for an emergency condenser vent release are presented, in tables.2-5 to 2-8. The.emergency condenser vent releases are assumed to travel to the perimeter' drain system and released from the discharge structureat a rate of .33 J1 3/sec. See Appendix A forthe dose factor Alot derivation. To expedite time the dose is calculated to a maximum individual. This maximum individual is a composite ýof the highest dose factor Aiut of each age group a for each organ tand each nuclide i. If a nuclide is detected for which a factor is not listed, then it will be calculated and included in a revision to the ODCM.SAll doses calculated in 'ltis manner:for each batch of liquid effluent will be summed for comparison with quarterly and annual limits, added to the doses accumulated from other releases in the quarter and year of interest. In all cases, the following relationships will hold: Unit I ODCM Revision 28 11,9 September 2006

1.3.1 Maximum

Dose Equivalent: Pathway (Cont'd)For a calendar quarter: Dt< 1.5 mrem total body Dt_< 5 mrem for any organ For the calendar year: D, <3 .0 mrem total body D, <10 mrnrem for any organ, Where: Dt = total dose received to the total body or any organ due to liquid effluent releases.If these limits are exceeded, a special report will be submitted to the NRC identifyingthe

cause and proposed corrective actions. In addition, if these limits are-exceeded by a factor of two, calculations shall be made. to determine if the dose .limitscontained in 40 CFR 190'have been exceeded.

Dose limits, as.contained in.40 CFR 190 are 'total body and.organ doses of 25. mrem per yearanda thyroid dosc of 75 mrem per year, These calculations will include doses as aresult of liquid and gaseous pathways as well as doses from direct radiation. The liquid pathway analysis will only 'includethe fish and sediment pathways since the drinking'water pathway is insignificant. This pathway.is only included in the station's effluent dose projections to complywith Part I Specifications. Liquid, gaseous and direct radiation: pathway doses will consider. the James A. FitzPatrick and Nine Mile Point Unit 2 facilities.as well as Nine Mile Point Unit I Nuclear Station.In the, event the calculations.demonstrate that the, 40 CFR 190 dose limits,.as defined above, have been exceeded, then a report shall be prepared and submitted to the COmmrissiun within 30 days as specified in Part I Section DLCO 3.6.15Ad.Section 3.0 of the ODCM contains more information concerning calculations for an evaluation of whether 40 CFR .190 limitshave been.exceeded. Unit 1 ODCM Revision 28 1110 September 2006 1.3.2 Dose Projections -Determinations of theNeed to Operate the Liquid Radwaste Treatment System 1.3.2.1, Requirements DLCO 3.6.16.a requires that the liquidradwaste system be-used toreduce the radioactive materials in liquid wastes prior to. their discharge when the projected doses due to the liquid effluent would exceed 0.06 mrem to the total body or 0.2 mrern to any organ for the batch.This Control implements TeulhniLal Specification 6.5.3.f thatrequires the Radioactive Effluent Controls Program to include limitations on the functional capability and use of the liquid effluent treatment. system to ensure the appropriate portion of this system is used to reduce releases of radioactivity. This is required when the projected doses would, exceed 0.06 mrem to the totalbody and 0.2 mrem to any organ. Since releases are performed much less frequently than once permonth, doses are to be projected prior to each release and the above limits: will be applied on a batch basis.1.3.2.2 Methodology The dose projection for each batch is calculated in the same manner as cumulative dose calculations for the current calendar quarter and current calendar year. See 111.1.4.4 and 11-1.3.1. If the calculated dose is greater than 0.06 mrem to the total body or 0.2 mrem to any organ, the appropriate subsystems of the liquid radwastc system shall be used to reduce the radioactivity levels of the batch prior to release.1.3.2.3 Continuous Liquid:Release Dose Projections Each month thata continuous liquid release is in progress, or is anticipated, the expected dose to man can beaccounted for or projected. Since a continuous release~does not result from not operating a portion of the Liquid Radwaste System,,projections are not required to determine or evaluate Radwaste System Functionality. Dose projections may he relevant to planning repairs, and in reporting intended actions. See 1-1.1.4.5. Unit I ODCM Revisioi 28 1 11 September 2006

2.0 GASEOUS

EFFLUENTS 2.1 Setpoint Determinations

2.1.1 Basis

Stack gas monitor setpointswill be established such that the instantaneous release rate of radioactive materials in gaseous effluents does not exceed the 10: CFR 20 limits for annual release rate. The setpoints will be activated if the instantaneous dose rate at or beyond the (land) site boundary would exceed 500 mrem/yr to the whole body or 3000 mrem/yr to the skin from thelcontinuous release of radioactive noble gas in the gaseous effluent.The offgas (condenser air ejector activity) monitor setpoints provideassurance that thetotal body exposure to an.individual at the exclusion areaboundary does not exceed a small fraction ofthe dose guidelines of 10 CFR 100.Emergency condenser vent monitor setpoints will be established such that the release rate for-radioactive materials in gaseous effluents do not exceed the Technical Specificationdose rate limits. Monitor setpoints foremergency condenser vent monitors are conservatively fixed at 5 mr/hr for reasons described in Section 11-2.1.4'and therefore do not require periodic recalculations. Monitor. setpoints.. from continuous.release points will be determined once per quarter under normal release rate conditions and will be based on the isotopic composition of the actual release in progress, or an offgas isotopic distribution or a more conservative default composition specified in the pertinent procedure. If the calculated setpoint is higher than the existing setpoint it is not mandatory that the .setpoint be changed.Under. abnormal site release rate conditions, monitor alarm setpoints from continuous release points:. will be recalculated and, if necessary, reset at:more frequent intervals. as deemed necessary by Chemistry Supervision. In particular, contributions ftom both JAF and NMP-2 and the Emergency Condenser Vents shall be assessed.During outages and until steady state power operation is again realized, the last operating stack and off gas monitor alarm setpoints shall be used.Since monitors respond.to noble gases only, monitor alarm points are set to alarm priorto exceeding the corresponding whole body. dose rates.The skin-dose rate limit isnot used in setpoint calculationsbecause it is never.limiting.

2.1.2 Stack

MonitorSetpoints The detailed methods for. establishing setpoints shall be contained. in the station procedures. These methods shall apply the following general ci'teria: (1) Rationale for Stack monitor settings is based on the general equation: release rate. actual release rate. max. allowable corresp. dose rate, actual corresp. dose rate, max. allowable Zi (Q) max_jQ (Vi + (SF)K 1 (X/Q),) 500mretn/yr Unit 1 ODCM Revision 28 H 12 September 2006

2.1.2 Stack

Monitor Setpoints (Cont'd)W/here:= release rate for each isotope i, gCi/sec.V= gamma whole body dose factor in units of mrem/yr per RCi/sec..(See Table 3-2).instantaneous',release rate. limit 0Ci/sec.SF, K, X/Q = See Section 11-2.2.1 LI.(2) To ensure that Part I dose rate limits are not exceeded, the Hi Hi alarms ,on the stack monitors shall be set lower than or equal to (0.9) (Q),14,. Hi alarms shall .be set lower than or equal to (0.5)(Q)mx (3) ~Based on the above conservatism, the dose contribution from JAF and NMP-2 can usually be ignored. During Emergency Classifications at JAF or NMP-2 due to airborne effluent, or after..emergeincy condenser vent~releases of significant proportions, the 500 mrci/yr value may be reduced accordingly. (01) To convert monitor gross count rates to pCi/sec release rates, the following general formula shall be applied: (Cm-B)Ks.=Q = Q Ci/sec, release.,rate Where: CM- = monitor gross count-rate inrcps. or cpm B = monitor background count rate K, .. = stack monitor efficiency factor with units of ItCi/sec-cps or gCi/sec-cpm. (5) Monitor K,.factors shall be determined using the general formula: K, = E iQI(C,,B)Where: Q = individual radionuclide stack effluent release rate as determined by isotopic. analysis.K. factors more conservative than those calculated by the above methodology may be assumed.Alternatively, when stack release rates.are near the lower linmit of detection, the following general formula may be used to calculate K':*1/K = F (- F, k Y E k) (3.7E4 dis)f f sec -11i Where: f stack flow in co/see.E = efficiency in units of cpm-cc/gCi or cpsa-cc/pCi (cpm = counts per minute; .cps = counts per second).Ek = cpm-cc/bps or cps-cc!'ps. From energy calibration curve produced during NIST traceable primary gas calibration or transfer source calibration (bps =.beta per second;yps:= gammas per second).Unit 1 ODCM Revision 28 1113 September 2006

2.1.2 Stack

Monitor Setpoints (Cont'd)Yk = b/d (betas/disintegration) or 7/d (gammas/disintegration). F = Activity fraction of nuclide&i in the mixture.i. = nuclide counter.k -discrete energy'beta or gamma cmitter per nuclide counter.s = seconds, This monitor calibration method assumes a noble gas distribution typical of a recoil release mechanism. To ensure that the calculated efficiency isconservative, beta or gamma emissions. whose energy is above the range of calibration of the.detector are not included in the~calculation.

2.1.3 Recombiner

Discharge (Off Gas) Monitor Setpoints (1) The Hi-Hi alarm points shall activate, with recombiner discharge rates equal to or Jess, than:'500,000 ACi/sec. This.alarm point~may be.setequal to or lessthan .1 Ci/sce for.a period of time not to exceed 60, days provided the.offgas. treatment system is in operation, According to Part I, Note (c) to Table D 4.6.14-2, the channel functional test of the condenser air ejector radioactivity monitorshall demonstrate thatautomatic;isolation of this pathway occurs if either of the following conditions exist: i) Instruments indicate two channels above the Hi-Hi alarm setpoint,:ii) Instruments indicate one channel above Hi-Hi alarm, setpoint and one channel downscalc. This automatic isolation, function is tested once per operating cycle in accordance with station procedures. (2) The. Hi alarm points shall be set to activate at equal to or less than five (5) times normal full power background. Ifthe monitor alarms at this-setpoint, the offgas will be immediately sampled and analyzed, followed by an analysis. of reactor coolant sample.-(3) To convert monitor mR/hr readings to gtCi/sec, the formula below shall be applied: (R)(KR) = QR iCi/sec recombiner discharge release rate Where: R = mR/hr monitor indicator. KR = efficiency factor in units of ItCi/sec/MR/hr determined priorto setting monitor alarm.points.(4) MonitorKR factors shall be determined using the general formula: K = R Where:=individual.radionuclide recombiner discharge release rate as determined by isotopic analysis and flow rate monitor.KR factors more conservative.than those calculated by the above methodology may be assumed.Unit I ODCM Revision 28 1 14 September 2006

2.1.3 Recombiner

Discharge (Off Gas) Monitor Setpoints (Cont'd)(5) The setpoints chosen provide .assurance that the total body exposure to an individual at the exclusion area boundary will notexceed a very small fraction of the limits of 10CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment (thereby implementing the requirements of General Design Criteria 60 and 64 of Appendix A to IOCFR Part 50). Additionally, these setpoints serve to limit buildup of fission product activity within the station systems which would result if high fuel leakage were to'be permitted over extended periods.2.1.4 Emergency Condenser Vent Monitor Setpoint The monitorsetpoint was established by calculation ("Emergency Condenser Vent Monitor Alarm Setpoint", January 13, 1986, NMPC File Code #16.199). Assuming a hypothetical case with (1) reactor water iodine concentrations higher thantthe Technical Specification Limit, (2) reactor water noble gas concentrations higher than would be expected at Technical Spec ification iodine levels, and (3) leakage of reactorsteam into the emergency condenser shell at 300% of rated flow (or 13 E6 lbs/hr), the calculation predictslan emergency condenser vent monitor response of 20 mR/hr. Such a release would result in less than 10 CFR 20. dose rate values at the site boundary and beyond for typical emergency condenser cooldown periods.Since a:20.mRlhr monitor response can, in theory. beachievableonly when reactor water iodines are higher than permitted by Technical Specifications, a conservative monitor setpoint of 5 mr/hr has been adopted.2.1.5 Discussion 2.1.5.1 Stack Effluent Monitoring

System Description

The NMP-1 Stack Effluent Monitoring System consisted of two subsystems; the Radioactive Gaseous Effluent Monitoring System (RAGEMS) and the Offgas Effluent Stack Monitoring System (OGESMS). The OGESMS shall be used to:monitor station noble gas effluents and collect particulates and iodine samples in compliance with Part I requirements. The RAGEMS was designed to be promptly activated from the Main Control Room for use in high range monitoring during accident situations in compliance with NIREG 0737 criteria. In accordance with a letter dated September 11, 2002 fromn the NRC to NMPNS, LLC, "Nine Mile Point Nuclear Station Unit I -Use of the Offgas Effluent Stack Monitoring System to Meet Regulatory Guide 1.97, Revision 2 and NUR-EG-073 7,"'OGESMS meets the objective and purpose of NtREG-0737 and RG 1.97. The sample line to RAGEMS will now be used as an additional auxiliary sample point.2.1.5.2 Stack Sample Flow Path -RAGEMS Auxiliary Sample Point: The effluent sample is obtained inside the stack at elevation 530' using an isokinetic probe with four orifices. The sample line then bends radially out and back into the stack; dcsccnds down the stack and out of the stack at approximately eevation 257'; runs horizontally '(enclosed in heat tracing) some 270' along the off gas tunnel; and enters Turbine Building 250' and Offgas Building 247'.Unit I ODCM Revision 28 1115 September 2006 2.1.5.3 Stack Sample Flow Path -OGESMS The OGESMS sample is obtained from the same stack sample probe as the RAGEMS Auxiliary Sample Point. From the exit of the stack at elevation 257', the sample line runs east. approximately 20' and then vertically approximately 8' to the OGESMS skid. In the OGESMS,.sample llows:thru a particulate/iodine cartridge housingand four noble gas scintillation detectors (i.e., 07 and 08 low range beta detectors:and RN-03A and RN-03B high range gamma detectors) .From OGESMS, the stack sampleflows back into the stack at approximately elevation; 257'.All OGESMS detector outputs.are monitored and recorded remotely in the Main Control Room. Alarmingcapabilities are. provided to alert Operators of high release rate conditions prior toexceeding PartlI Control DLCO 3.6.15b (1)(a) whole body dose.rate limits.Stack particulate and. iodine samples are retrieved manually fromrthe OGESMS and analyzed' in the laboratory using gamma spectroscopy at frequencies, and LLDs specified in Part I Table D 4.6.15-2.2.1.5.4 Sampling Frequency/Sample Analysis Radioactive gaseous wastes shall be sampled. and analyzed in accordance~with the sampling and analysis program specified inPart I Table D 4.6.15-2. Noble gas;sample and analysis frequencies are increased during elevated release rate.conditions. Noble gas sample and analysis are also performed following startup, shutdown and in conjunction with :each drywell purge. Particulate samples are saved and. analyzed for principal gamma emitters, gross alpha, Fe-55, Sr-89, Sr-90 at monthly intervalsmiirimally, and in response to an increase in noble gas release rate. The latter three analyses are performed off-site from a composite sample.Consistent with.Part I Table D 4.6.15-2, stack effluent tritium is sampled monthly, during each drywell purge, and weekly when fuel is off loaded until stable release rates are demonstrated. Samplesare analyzed off-site.Line loss correction. factors are applied to allparticulate and iodine results. Correction factors of 2.0 and 1L5 are used for data obtained from RAGEMS Auxitiary Sample Point and OGESMS respectively. These correction factors are based on empirical data from sampling conducted at NMP-I in 1985 (memo from J. Blasiak to RAGEMS File, 1/6/86,"Stack Sample Representativeness Study: RAGEMS versus In-Stack Auxiliary Probe Samples"). 2.1.5.5 1-133 and 1-135 Estimates Monthly, the stack effluent shallbe sampled for iodines over a.24 hour period and the 1-135/1-131 and the 1- 133/I-13 1.ratios calculated. These ratios shall be used to calculate 1-133,.1-135 release for longer acquisition samplescollected during the month.Unit I ODCM Revision 28 1116 September 2006 .2.1.5.5 1-133 and 1-135 Estimates (Cont'd)Additionally, the 1-135/.1-131 and 1-133/I-13.1 ratios should also be determined after a significant change in the ratio is suspected.(eg, plant status changes from prolonged shutdown to power: operation or fuel damage has occurred). 1-135 will be included in the Radioactive Effluent. Release Report in accordance with Regulatory Guide 1.21 but if will not be included when totaling dose rate or dose.2.1.5.6 Gaseous Radwaste Treatment System Operation Part I Control DLCO 3.6.16.b requires that the gaseous radwastetreatment system shall be functional and shall be used to reduce radioactive materials in-gaseous waste prior to their.discharge as necessary to meet the requirements of Part.l Conirul DLCO 3.6.1 5.b.To ensure Part I Control DLCO 3.6.15.b limits are not exceeded, and to confirm proper radwastc treatment system operation as applicable, cum~ulative dose contributions forthe current calendar quarter and current calendar year shall be determined monthly in accordance with section 2.2 of this manual. Initial dose calculations shall incorporate the following assumptions with regard to release rates of non-gamma emitting radionuclides subsequently analyzed off-site: a) H-3 release rate < 4 fCilsec b) Sr-89 release rate < 4 xCs-137 release rate c) Sr-90 release rate. 0.5 x Cs- 137 release rate d) Fe-55 release rate < I x Co-60 release. rate Assumed release rates represent conservative estimates derived from analysis of historical data from effluent releases and.process waste streams (See NMP 34023, .C. Ware to J.Blasiak, April 29, 1988, "Dose Estimates for Beta-.Emitting Isotopes"). Following receipt of off-site H-3, Sr-89, Sr-90, Fe-55 analysis information, dose estimates shallbe revised using actual radionuclide concentrations. 2.2 Dose and Dose Rate Determinations In accordance with Technical Specifications 6.5.3, "Radioactive Effluent Controls Program, and ODCM.Part I Controls DSR 4.61.54b.(.), DSR 4.6.15.b.(2), and DSR 4.6. 5.b.(3) dose.and dose rate determinations will be made monthly to determine: (1) Whole body dose ratesand gamma air doses at the maximum X/Q land sectorsite boundary interface. (2) Skin dose rates.and beta air doses at the maximum X/Q land sector site boundary interface. (3) The critical organ dose and dose rate at a critical receptor location beyond the site boundary.Average meteorological data (ie, maximum five year annual average XiQ and D/Q values in the case of elevated releases or1985 annual average X/Q and D/Q values, in the~caseof ground level releases) shall.be utilized for dose and dose rate calculations. Where average meteorological data is assumed, dose and dose rates due to noble gases at locationsbeyond the site boundary will be lower than equivalent site boundary dose, and dose rates. Therefore, under these conditions,. calculations of noble gas dose and dose rates beyond the maximum X/Q land sector site boundary locations can be neglected.. Unit 1 ODCM Revision 28 1117 September 2006 2.2 Dose and Dose Rate Determinations (Cont'd)The frequency of dose rate calculations will be upgraded when elevated release rate conditions specified in subsequent'sections 11-2.2. 1.1 and 1H-2.2.1.2 are realized.In accordance with Technical Specification 6.5.3.g, noble gas dose ratetothe whole body and skin will be calculated at the site boundary. In accordance With Technical'Specification 6.5.3.h, gamma and beta air doses may be calculated at a point beyond the site boundary.To demonstrate compliance with Technical Specification 6.5.3, "Radioactive Effluent.Controls Program", critical organ doses and dose rates may be conservativelycalculated by assuming the existence of a maximum individual. This individual is a composite of the highest dose factor of each age group, for each organ and total, body, and each nuclide. It is assumed thatall pathways are applicable and the highest X/Q and/or D/Q value for actual pathways as noted in Table 3-1 are in effect. The maximum individual's dose is equal to the same dose that person would receive if Lbey were simuitaneously subjected to the highest pathway dos at each critical receptoridentified for each pathway. The Pathways include grass-(cow and goat)-milk, grass-cow-meat, vegetation, ground plane and inhalation. To comply with Part I requirements the maximum individual dose rate will be calculatedat this hypothetical critical residence. ltdose or dose rates calculated, using the assumptions noted above, reach:Part I limits, actual pathways will be evaluated, and dose/dose rates may be calculated at separate critical receptor locations and compared with applicable limits.Emergency condenser vent release contributions to the monthly dose and dose rate determinations will be considered only when the emergency condenserretum isolation valves have been opened for reactor cooldown, if Emergency Condenser tube leaks develop with or without the system's return isolation valve opened, orif significant activity is detected in the Emergency Condenser Shell.Without tube leakage, dose contributions from emergency condenser vent releases are to be determined based on condensate: storage tank and emergency condenser shell .isotopic distributfions, When releases from the emergency condenser have occurred, dose rate and dose determinations shall be performned using methodology in 11-2.2.1 and 11-2.2.2. Furthermore, environmental sampling may also be initiated to refine any actual contribution to doses. See Section 11-2,4.2.2.1 Dose Rate Dose rates will be calculated monthly, at a minimum, or when the Hi-Hi stack monitor alarm setpoint is reached, to demonstrate that dose rates resulting from the release of noble gases.tritium, iodines, andparticulates with half lives greater than 8 days. arewithin the limits specified in Technica! Specifications Section 6.5.3,,"Radioactive Effluent Controls Program".These limits are: Noble Gases Whole Body Dose Rate: 500 mrem/yr Skin Dose Rate: 3000 mrem/yr Tritium. lodines and Particulates Organ Dose Ratc: 1500 mrcm/yr Unit I ODCM Revision 28 HI 18 September 2006 2.2.1.A Noble Gases The following noble gas dose rate equation includes the contribution.from the stack (s)elevated release and the emergency condenser vent (v) ground. level release when applicable (See section 11-2.2).For whole body dosenrates (mrem/sec): DR, (mrem/sec)= 3.17E-8 i [(V, + (SF) Ki (X/Q)j) Qý + (SF)Kj (X/Q),Qiv] For skin dose rates (mrem/sec): D.,rj(mrem/seC)= 3.17F-gA i.[(lI(X/Q). + 1. 11 (SF)(B, + Mi(,X/Q).))Qi, +(Lj + 1.1 1 (SF)Mj)(X/Q),Qjj] Where: DR 7 = whole body gamma dose rate (mrem/sec). DRTEp = skin dose rate from gamma and beta radiation (mrem/sec); Vj = the constant accounting for the gamma whole body dose rate from stack radiation for an elevated finite plume releases for each identified

noble gas nuclide, i. Listed on Table. 3-2 in mrem/yr per P, Ci/sec'.Ki the constant accounting for the gamma:whole body dose rate from.immersion in the semi-infinite cloud for each identified noble gas.nuclide, Ii. Listed in Table 3-3 in mrem/yr per j+/-Ci/m 3 (from Reg.Guide 1.109)Qi,,Qi, = the release rate ofisotope i from the s.tack(s) or emergency condenser vent(v); (gtCi/sec)

SF structural shielding factor.X/Q the relative plume concentration. (in units of sec/rn 3) at the land sector site boundary or beyond. Average meteorological data (Table 3-1.) is used. "Elevated" X/Q values are used for stack releases (s = stack);"Ground" YX/Q values are used for Emergency Condenser Vent releases (v = vent)..Lj the constant accountingfor the beta skin dose rate from immersion in the semi-4infinite cloud for each identified noble gas nuclide, i. Listed in Table 3-3 in mrem/yr per gLCi/V 3 (from Reg. Guide 1. 109), B, the, constantI accounting for the air gamma radiation from the elevated Finite:plume resulting from stack.releases. for each identified. noble.gas nuclide, i. Listed in Table 3-2 in mrad/yr per ptCi/sec.Unit I ODCM Revision 28 1. 19 September 2006 2.2.1.1 Noble Gases (Cort'd)Mi = the constant accounting for the gamma air dose rate from immersion in the semi-infinite cloud foreach identified noble gas nuclide, i. Listed in Table 3-3 in mrad/yr per .iCi/m 3 (from Reg. Guide 1.109)See Appendix B for derivation of Bi and Vi.To ensure that the site noble gas dose rate.limits are not exceeded, the following procedural actions are taken if the offsite dose rates from Unit I exceed 10% of the liimits: 1) Notify Unit I SSS (Station Shift'Supervisor) and Unit 1 Supervisor Chemistry.

2) Notify Unit 2.SSS and Unit 2 Supervisor Chemistry and request the,Unit 2 contribution to offsite dose rate.3) Notify SSS of the James A. Fitzpatrick Nuclear Plant and request the Fitzpatrick contribution to offsite dose rate.4) Increase the frequency of perforrmingnoble gas dose and dose rate calculations, if necessary, to ensure Site (Nine Mile Point Units 1 and 2 and Fitzpatrick) limits are not exceeded.Additionally, alarm setpoints are set at 90% of the dose rate limit to ensure that site limits are-not exceeded.

This alarm setpoint is. adjusted if the noble gas dose rate from Unit .1 is greater than /o 0 of the limit.2.2.1.2 Tritium, Iodines and Parlticulates To ensure that the 1500 mrern/year site'dose ratelimit is not exceeded, offsite dose rates for tritium,, iodine and particulates with half lives greaterthan 8 days shall be calculated monthly and when release:rates (Q) exceed 0.34 VCi/sec using. the following equation.DA, (morem /sec)= 3 i7E-8 A [ Rj [WM Qi, + WV, Qij]Where;D~k Total dose rate to each organ k of an individual in age group a (nrem/sec). Wi dispersion parameter either X/Q (sec/m 3) or DIQ (I(/nm) depending on pathway and receptor location assumed.: Average meteorological data is used (Table 3-.). "Elevated" Wj, values are used for stack releases (s = stack); '!Ground" Wj Values are used for Emergency Condenser Vent :releases (v = vent).Qi = the release rate of isotope i,. from the stack (s) or vent(v); (IiCi/sec). Unit I ODCM Revision 28 II 20 September 2006 2.2.1.2* Tritium, Todines and. Particulates (Cont'd)Rijak thedose factor~for each isotope i, pathway j, agc group a, and organ k (Table3-4, through 3-22; m 2-mrem/yr per gCi/sec for all pathways except.inhalatioln, rnrem/yr per RiCi/m3. The .R values contained in Tables 3-4 through 3-22 were calculated using the methodology defined in.NUREG-0133 and parameters from Regulatory Guide 1.109, Revision 1.; as presented in Appendix C.3.17E-8 = the inverse of the number of seconds in a year.The use of the 0.34 gtCi/secrelease rate threshold to perform Unit I dose rate calculations is justified as follows: (a) The 1500 mrem/yr organ dose rate-limit corresponds to a minimum release rate limit.of 0.34 gCi/sec calculated using the:equatiOn:. 1500 = (Q, gtCi/see) x (RijWj)maX Where: 1500 = site boundarydose rate limit (mrem/year).(RijWj)m., = the maximum curie-to-dose conversion factor equal to 4.34E3 mrem-sec/ltCi-yr for Sr-90, child bone for the vegetation pathway at the critical residence receptor location beyond the site boundary for an elevated release.(b) The use of 0234 .Ci/sec release rate threshold and the4.34E3 mrem-sec/iCi-yr curie-to-dose conversion factor is conservative since curie-to-dose conversion factors for other isotopes likely to be presentare significantly lower.If the organ dose rate exceeds 5% of the annual limit, the following procedural actions will be taken: 1) Notify Unit 1 SSS (Station Shift Supervisor) and Unit I Supervisor Chemistry.

2) Notify Unit 2SSS and Unit 2 Supervisor Chemistry and request the Unit 2 contribution to offsite dose rate.3) Notify SSS of James A. Fitzpatrick Nuclear Plant and request JAF's contribution to.offsite dose rate.4) Increase the frequency of performing dose and dose rate calculations if necessary to ensure.site (NineMile Point Units 1 and 2 and Fitzpatrick) limits are not exceeded.Unit 1ODCM Revision 28 1121 September 2006 2.2.2 Dose Calculations willbe. performed monthly at a minimum, to demonstrate that doses resulting from ýthe release of noble gases, tritium, iodines, and particulates with half lives greater than 8 days are within the limits specified in 10 CFR .50, Appendix l, These limits are: Noble Gases:5 mR gamma'calendar quarter 10 mrad beta/calendar quarter 10 mR gamma/calendar year 20 mrad beta/calendar-year Tritium, lodines and Particulatcs 7.5 mrem to any organ/calendar quarter 15 mrem to any organ/calendar year 2.2.2.1 Noble Gas Air Dose The following Noble Gas air dose equation includes contributions from the stack (s)elevated release and the emergency condenser vent (v) ground level release when applicable (see section 11-2.2):: For gamma radiation' (rnrad): D, (mrad) = 3.17E-8 Y i [(Bli+ Mj(X/Q)) Qi, + Mi(X/Q), Q 1] t For beta radiation (mrad): Dp (mrad) 3.17E-8 E iNi[(X/Q)s Ql .+ (X/Q0) Qi, JIt., Where: D, ' :gamma air dose(mrad).

DI= beta air dose (mrad).Note that the units for the gamma air dose are in mrad compared to the units for the limits are in.mR. The NRC recognizes that I iR=RI urad, for gamma radiation. Bi the constant accounting for the air gamma radiation from the elevated finite plume resulting from. stack releases for each identified noble gas nuclide, i. Listed in Table 3-2 infmad/yr per gCi/sec.Ni = the constant accounting for the air beta dose from immersion in the semi-infinite cloud for each identified noble gas nuclide, i.Listed on Table 3-3 in mrad/yr per pCi/m 3 (from Reg. Guide 1109).Qis, Qiý, therelease rate of isotope i, from the stack: (s) orvent,(v);(paci/sec). Unit I ODCM Revision 28 Fi122 September.2006 2.212.1 Nnble Ga~q :Air Dose (Cont'd)3J7E-8 the inverse of the-numberof seconds in a year.Mi the constant accountting forthe air gamma dose from immersion in the semi-infinite cloud for each identified noble gas nuclide,. i.Listed on Table 3-3 in mrad/yr per tCi/mr 3 (from Reg. Guide 1.109).t = total time during release period, sec.All other parameters are as defined in section 1-.2.2.. 1.2.2.2.21 Tritium, lodines and Particulates To ensure that the 15 mremlyr facility dose limit is not exceeded, offsite doses for tritium, iodines,.and particulates with half lives greater than 8 days shall be calculated monthly using the following equation: Dak (mrem) =3.17E-81 j[ E i:.Rijj, [WS. Qks + W1 Q'i] ]t Where:total dose to each organ k of an individual in age group a(mrem).Wj dispersion parameter eitherX/Q (sec/rn) or-D/Q (/rn,)depending on pathway: and receptor location assumed. Average meteorological data is used (Table 3-1). ".Elevated" Wj values are used for stack releases.(s = stack); "Ground" Wj values are used for Emergency Condenser Vent releases (v = vent).=Q the release rate of isotope i from stack(s) or vent (V'); (pCi/sec). Rija = the dose. factor for each isotope i, pathway j, age group a, and organk (Tables 3-4. through 3-7, mnrem/yrper Tables 3-8 through 3-22, m 2-mrem/yr per gCi/sec). R values contained in Tables 3-4 thruugli 3-22 were calculated using the methodology defined in NUREG-01331and parameters from Regulator), Guide 1.109, Revision 1; as presented in Appendix C.3.17E-8 the inverse of the number. of.seconds in a year.total time during the release period, see.Unit 1 ODCM Rlevision 28 11 23 September 2006 2.2.2.3 Accumulating Doses Doses will be calculated monthly, at a minimum, .for.>mma air, beta. air, and the critical organ for each age group. Dose estimates will, also, be calculated monthly prior to receipt of any offsite or onsite analysis data i.e., strontium, tritium, and iron-5.5 Results wiltbe summed .for each calendarquarter and year.The critical doses are based on the following:,-noble gas plume air dose direct radiation from ground planc. deposition inhalation dose-. cow milk ingestion dose-goat milk ingestion dose cow meat ingestion dose vegetation (food crops) ingestion dose The quarterly and annual results shall be compared to-the lim its listed in paragraph 11-2.2.2. If the limitsare exceeded, special reports, as required by Part. I Section D 6.9.3 shall be submitted. 2.2.3 Dose Projections -Determination of Need to Operate Gaseous Radwaste Treatment System and Ventilation Exhaust Treatment System 2.2.3.1 Requirement DLCO 3.6.16.b requires that the Gaseous Radwaste Treatment System be used to reduce the radioactive materials in gaseous waste prior to their discharge as necessary to meet the requirements of DLCO 3.6.15. DLCO3.6.16.b(2) requires that the Ventilation Exhaust Treatment System be used to reduce releases of radioactivity when the projected doses in 31 days would exceed 0.3mrem-to any organ. These Controls implement Technical Specification 6.5.3.f that requires the Radioactive Effluent Control Program to include limitations on the functional capability and use of the gaseous effluent treatmentsystems (Gaseous Radwaste Treatiuent System AND Ventilation Exhaust Treatment'Systcm) to ensure the. appropriateportions. of these systems are used to reducereleases of radioactivity. TheGaseous Radwaste TreatmentSystemnisexpected to be in service. For the Ventilation Exhaust Treatment System, use is required when the projected doses in a period of 31 days would exceed 2%of the guidelines for the annual dose or dose commitment, conforming to IOCFR50, Appendix I, i.e., 3 mrem to any .organ. Whaen Treatment systems are not in use, doses are lobe projected every 31 days.The appropriate components, which affect iodine or particulate release, to be in use are: Rad Waste: Building FLT-204-24 FLT-204-25 FLT-204-69 FLT-204-70 RSSB FLT-204-147 Unit I ODCM Revision 28 II 24 September 2006 2.2.3.2 Methodo1gy Due to system design and operating procedures the charcoal beds are always operated When the offgas system is in operation. Therefore, dose projection is not relevant to determining need to operate..If the Gaseous Radwaste Treatment System becomes nonfunctional. for more than. seveni days a Special Report to the NRC is required. This -report will includeappropriate dose assessments (cumulative and projected). If Ventilation Exhaust Treatment System components become nonfunctional which prevent building:effluent from being filtered, dose projections will be performed monthly using the methodology of Section I-2.2.2.2. Assumptions for released activity will be added to historical routine stack emissions for calculating dose during the anticipated .period of component unavailability. The calculated projecteddoses for iodine.and particulates:will be compared to the DLCO 3.6.1 5.,b limitsand Technical:Specifications Section 6.5.3.f limit, 0.3 mrem to any organ.2.3 Critical Receptors in accordance with the provisions of 10 CFR 20 and 1..0 CFR 50, Appendix I, the critical receptors have been identified and are contained in Table 3-1.Forelevated noble gas :releases the critical receptor isthe site boundary.When 1985 average annual X/Q values are used.for ground levelnoble gas releases, the critical receptor is the maximum XIQ land sector site boundary interface..For tritium, iodines, and particulates with lhalf lives greater than eight days, the critical pathways are grass-(cow and goat)-m ilk, grass-cow meat, vegetation, inhalation and direct radiation (ground plane) as a result of ground deposition. The grass-(cow and goat)-m ilk, and grass-cow-meat pathways will be based on the greatest D/Q location. This location has been determined in conjunction with the land use census (Part I Control DLCO 3.6.22) and is subject to change. The vegetation'(food crop) pathway is based on the greatest D/Q garden location from which samples are taken. This location may also be modified as a result of vegetation sampling surveys, The inhalation and ground plane dose pathways will be calculated at the critical residence. Because Part'I states to calculate "at the site boundary or beyond", the doses and/or dose rates must be calculated for a maximum individual who is exposed to applicable pathways at the critical residence.. The maximum individual is a composite of the highest dose factor of each age ,group, for each organ and total body, and each nuclide.Unit 1ODCM Revision 28 TI 25 September 2006

2.4 Refinement

of Offsite Doses Resulting from Emergency Condenser Vent Releases The doses resulting from the operation of the emergency condensers and calculated in accordance with 11-2.2.2 may be refined using datafrom actual environmental samples.Ground deposition samples will be obtained from an. area or areas of maximum projected deposition. These areas are anticipated to be at or near the site bo undary and near projected plume centerline. Using the methodology found in Regulatory Guide 1. 109, the dose will be calculatedjto tle maximum exposed individual. This dose.,will then be compared tothe dose calculated in accordance with 11-2.2.2. The comparison will result in an adjustment. factor of less than or greater than one which Will be used to adjust the other doses from other pathways. Other environmenta samples may also be collected and the resultant calculated doses to the maximum exposed individual compared to the dose calculated per IT-2.2.2. Other environmental sample media may include milk, vegetatiorn (suchas garden broadleaf vegetables), etc. The adjustment factors from these pathways may be applied to the doses calculated per 11-2.2.2 on a.pathway by pathway basis or several pathway adjustment factors may be averaged and used to adjust calculated. doses.Doses calculated from actual environmental sample media will be based on the methodology presented in Regulatory" Guide 1.109. The regulatory guide equations may be slightly modified to account for short intervals of time (less than one year) or modified for simplicity purposes by deletingdcceay factors. Deletion of decay factors would yield more conservative results.Unit I ODCM Revision 28.[1 26 September 2006 1.0 40 CFR 190 REQUIREMENTS The "Uranium Fuel Cycle" is defined in 40 CFR Part 190.02 (b).as follows: "Uranium fuel cyclemeans the operations of milling of uranium ore, chemical conversion of uraniuim, isotopic enriclhmcnt of uranium, fabrication of uranium fuel, generation of electricity by a light-water-cooled nuclear power'plant using uranium fuel, and reprocessing of:spent uranium fuel, tothe extent that these directly support the production of electrical power for public use utilizing nuclear energy, but excludes mining operations, operations at waste disposal sites, transportation of any radioactive material in support of these:operations, and the reuse of cecovered non-uranium special nuclearUand by-product materials from the cycle." Control DLCO 3.6. 15.d of Part I requires that whenthe calculated doses associated with the effluent releases exceed twice the applicable quarter or annual limits,, the licensee shall evaluate the calendar year doses and, if required, submit a Special Report to the NRC and limit subsequent releases such that the dose commitment to a real individual-from all uranium fuel cycle sources is limited to 25 mrem to the total body or any organ (except the thyroid, which is limited.to 75 mrem). This report is to :demonstrate that radiation exposures to all real individuals from all uranium fiel. cycle sources (including all liquid and gaseous effluent pathways and direct radiation) are less than the limits in 40 CFR Part 190.If releases that resultin doses exceeding the 40 CFR 190 limits have occurred, then a.variance from the NRC to permit such releases will be requested and if possible, action will.be taken to reducesubsequent rclcases.The report to the NRC shall contain: 1) Identification of all uranium fuel~cycle facilities or operations.within 5 miles of the nuclear power reactur units at the site that contribute to the annual dosc of the maximum exposed member of the. public.*2) Identificationof the maximum exposed member of the public and a determination of the total annual dose tothis person from existing pathways and sources ofradioactive effluents and direct radiation. The total body and organ doses resulting from radioactive material in liquid:effluents from.Nine Mile Point Unit 1 will be summed with the maximum doses resulting from the releases of noble gases, radioiodines, and particulates for the other calendar quarters (as applicable) and from the calendar quarter in which twice the limit was exceeded. The direct dose components will be determined by either calculation or actual measurement. Actual measurements will utilize environmental TLD.:dosimetry. Calculated measurements will utilize engineering calculations to determine a projected direct dose component. In the event calculations are used, the methodology will be detailed as required in Part I Section D 6.9.1.e.Unit 1 ODCM Revision28 I1 27 September 2006 3.0 40 CFR 190 REQUIREMENTS (Cont, d)The doses from Nine Mile.Point Unit I will be added to the doses to the maximum exposed individual that are contributed from other uranium fuel cycle.operations:within

5 miles of the site. Other uranium fuel cycle facilities within 5 miles of the Site includeNine Mile Point.Nuclear Station Unit 2 and the James A.. Fitzpatrick Nuclear Power Plant. Doses from other facilities wiil be calculated in accordance with each facilities' ODCM.For the purpose~of calkulating doses, the results of the Radiological Environmental Monitoring Program may be included for providing more refined estimiates of doses to a real maximum exposed individual.

Estimated doses, as calculated.from stationeffluents, may be replaced hydoses calculated from actual environmental sample results. Reports will include all significant details of the dose determination if radiological sampling and analyses are. used to determniie if the dose limnits:of 40CFR190 are exceeded.3.1 Evaluation of Doses From Liquid Effluents For the evaluation of doses to real members of the public from liquid effluents, the fish consumption and shoreline sediment ground dose Will be considered. Since the doses from other aquatic pathways are insignificant, fish consumption and shoreline sediment are the only two pathways that will be considered. The dose associated with fish consumption may be calculated using effliuent data and Regulatory Guide L.109methodology or by calculating a dose to man based on actual fishsample analysis data. Because of the nature of the receptor location and the extensive fishing in the area, the critical, individual may be a teenager or an adtilt. The dose associated with shoreline sediment is based On the assumption that the shoreline would'be utilized as a recreational area. This dose may be derived from liquid effluent data and Regulatory Guide 1. 109 methodology or from actual shoreline sedimentsample analysis data.Equations umed to evaluate doses from actual fish and shoreline sediment. samples are based on Regulatory Guide 1.1.09 methodology. Because of the sample medium type and the half-livesof the rad ionuclides historically observed,. the decay corrected portions of the equations are deleted. 'This does not reduce the consentatism of the calculated doses but increases the simplicity from an evaluation point of view. Table 3-23 presentsthe parameters used for calculating doses from liquid effluents. The dose from fish sample media is.:calculated as: R.Rpj i , [Cif.(U)(Dý,jp) f] (.l E-3)Where: Rapj = The total annual dose toorganj, of an individual of agegroup a,from nuclide i, via fish pathway p, in mrem per:year.Cif = The concentration of radionuclide i in fish samples in.pCi./gram. U = The consumption rate of fish inkgiyr.!E+3 = Grams per kilogram, Unit I ODCM Revision 28 II 28 September 2006

3.1 Evaluation

of Doses From Liquid Effluents (Cont'd)(Pai~j) = The ingestion dosc factor for age group a, nuclide i, fish pathway p, and organ j, (Reg. Guide 1.109, Table E-l I) (mrem/pCi). f The fractional portion of the year over which the dose. is applicable. The dose from shoreline sedimeni sample media is calculated.as: Rapj = ' (U)(.4E+4)(0.3)(Daipj) f]Where: Rpj. The total annual dose to organ j,-of an individualof age group a, from nuclide i,.via the sediment pathway p, inf mrem per year.Ci. = The concentration ofradionuclide i in shoreline sediment in pCilgram.U = The usage factor, (hr/yr).(Reg. Guide 1.109).4E+4 = The product of the assumed density of shoreline sediment (40 kilogram per square meter to a depth of 2.5 cm) times the number of grams per kilogram.03 = The shore width.factor for a lake.Daipj = The dose factor for age group a, nuclide i, sediment pathwa Z s,I and organ j. (Reg. Guide IA 09, Table E-6)(mrem/hr per pCi/.m).f = The fractionalIportionofthe yearover which tihe .dose :is applicable.

3.2 Evaluation

of Doses From Gaseous Effluents For the evaluation of doses to real members of the public, from gaseous effluents, the pathways contained in section IT-2.22.3 of the.OT)CM will be considered'. These include the deposition, inhalation cows milk, goats milk, meat, and food products. (vegetation). However, any updated field data may be utilized that concerns 'locations of real individuals, real time meteorological data,: location of critical receptors, etcý Data from the most recent census and sample location surveys should be utilized. Doses may also :be calculated from actual environmental sample media, as available. Environmental sample media data such as TLD, air sample, milk sample and vegetable (food crop) sample data may be utilized, in lieu of efiluent calculational data.Doses tomember of the public from the pathways contained in ODCM section 11-2.2.2.31 as a result of gaseous effluents will be calculated using the dose factors of Regulatory Guide 1.109 or the methodology of the ODCM,.as applicable. Doses calculated from environmental sample media will be based.on .the, methodologies foundin Regulatory Guide. 1.109.Unit1 ODCM Revision 28 11 29 September 2006 3-3 Evaluation of Doses From Direct Radiation The dose contribution as a result of direct radiation shall bcconsideredwhcn evaluating. whether the dose limitations of 40 CFR 190 have been exceeded.Direct radiation doses as a result of the reactor, turbine and radwaste buildings and outside radioactive storage tanks (as. applicable)may be evaluated by engineering calculations or by evaluating environmental TLD results at critical receptor locations, site boundary or other special interest locations. For the evaluation of direct radiation doses utilizing environmental TLDs, the critical. receptor in question, such as the critical residence, etc.:, will be compared:to the control locations. The comparison involves the difference .in environmental TLD results between the receptor locationand the average control location result, 3.4 Doses to Members of the Public Within the Site Boundary The Radioactive Effluent Release Report shall include an assessment. of the radiation doses from radioactive liquid and gaseous effluents.to members of the public dueto their activities inside the site boundary as defined by Figure 5.1-1 of the Technical Specifications. A member of the public, as defined iniPart I, would be represented by an. individual who Visits the site's Energy Center for the purpose of observing the educational displays or for picnicking and associated activities. Fishing Iis a major recreational activity in the area and on'the Site as a result ofthesalmonoid and trout populations in Lake Ontario. FiShermen have been observed fishing at the shoreline near the Energy Center from April through-December in all weather conditions. Thus, fishing is the major activity performed by members of the public within the site Based on the nature of the fishermen and undocumented observations, it is conservatively assumed that the maximum exposed individual spends an average of 8 hours per week. fishing from the shoreline at a location between the Energy Center andthe Unit I facility. This estimate is considered conservative but not necessarily excessive and accounts for occasions where individuals may fish more on weekends or on a few days in March of the year.The pathways considered for the evaluation include the inhalation pathway, the ground dose pathway with the resultant whole body and skin dose and the direct radiation dose pathway with the associated whole body dose. The~direct radiation dose. pathway,,in actuality, includes several pathways. These include: the direct radiation gamma dose to an individual from an overhead plume, a gamma submersion plume dose (as applicable), possible direct radiation dose from the facility and aground plane dose (deposition), Because the location is in close proxilmity to the site, any beta plume submersiondose is felt to be insignificant. Other pathways, such as the ingestion pathway, are not applicable since these doses are included under calculations for doses to members of the public outside of the site boundary.In addition, pathways associated with water related recreational activities, other than fishing, are not applicable here. These include swimming, boating and wading which arc prohibited at the facility.Unit I ODCM Revision 28 HI 30 September 2006

3.4 Doses

to Members of the Public Within the Site -Boundary (Cont'd)The:iinhalation pathway is evaluated by identifying the applicable

radionuclides (radioiodine, tritiuni and particulates) in the effluent for the appropriate time period. The radionuclide concentrations

are then multiplied by the appropriate X/Q value, inhalation dose factor, air intake rate, and the fractional portion of the year in question.

Thus, the inhalation, pathway is evaluated using thefollowing equation adapted from Regulatory Guide 1.109. Table 3-23 presents the :reference for the parameters used in the following equation.NOTE: The f6llowingequation is adapted from equations C-3 and C-4 of.Regulatory Guide 1.109. Sincemany of the factors are in units ofpCi/3 , m 3 ,/sec., etc., and since the radionUclide decay expressions have been deleted because of the short distance to the receptor location, the equation presented here is not identical toathe Regulatory Guide equations. Dja ~ ~~~ = -: [CF(X/Q)(DFA)ij.(.BR);J Where: Dj. TThe maximum dose from all nuclidesto the organ j and age group (a) in mrem/yr.SCi, The average concentration in the stack release of nuclide i for the period in pC/m3.F Uniti average stackflowrate in m3/sec.X/Q = The plume dispersion parameter for a location approximately 0.50 miles west ofNMP-,; the plume dispersion parameter is 8.9E-06 seclm 3 (stack) and was obtained from the C.T. Main five year average annual.X/Q tables. The stack (elevated) X/Q is conservative whenbased on 0.50 miles because of the close proximity of the stack and the receptor location.(DFA).ja " The dose factor for nuclide i, organ j, and age group a inmrem per pCi (Reg. Guide 1. 109% Table .E-7).(BR)a Annual air intake for individuals in age group a in mn 3 per year (obtained from Table E-5 of Regulatory, Guide 1.109).=t Fractional portion of the year for which radionuclide i was detected and for which a dose is to be calculated (in years).Unit 1 ODCM Revision 28 11 31 September 2006

3.4 Doses

to. Members of the Public Within the Site Boundary .(Cont'd)The ground dose pathway (deposition) will. be evaluated by obtaining at least one soil or shoreline sediment. sample in the area where fishing occurs. The dose will then be calculated using the sample results, the time period in question, and the methodology based. on RegulatoryGuide 1.109 as presented in Section II-3.1.. The resultant dose may be adjusted for a background dose by subtracting the applicable off-site control. soil or shoreline sediment.sample radionuclide actiVities. In the event it is noted. that:fishingis not performed from the shoreline, but is instead performed in the water (i.e., the use of waders), then the ground dose pathway (deposition) may not be evaluated, The direct radiation gamma dose pathway includes any gamma doses from an overhead plume, potential submersion in theplume. possible direvt radiation from' the facility and ground plane.dose (deposition). This general pathway will be evaluated by average environmental TLD readings. At least two environmental TLDs will be: utilized at one location in the approximate area where fishing occurs. The TLDs will be placed in the field.on approximately the beginning of a calendar quarter and removed on approximately the end of the calendar quarter. For the purposes of this evaluation, TLD data from quarters 2, 3, and 4 W-Il be utilized.The average TLD readings wilibe adjusted by the average control TLD readings. This is accomplished by subtracting the average quarterly control TLD value from the average fishing location TLD value. The applicable quarterly control TLD values will be utilized after adjusting for the appropriate time periodý (as applicable). In the event of loss or theft of the TLDs, results from a TLD or TLDs in a nearby arearmay be utilized.Unit I ODCM ReVision 28 11 32 September 2006 4.0.ENVIRONMENTAL MONITORING PROGRAM 4.1 Sampling Stations The current sampling locations are specified in Table 5-1 and Figures: 5.L-1, 5.1-2. The meteorological tower is shownin Figure .5.1-1. Thelocationis:shown as TLD location 17.The Radiological Environmental Monitoring Program is ajoint effort between the owners and oPerators of.the Nine'Mile Point Unit1 and~the James A. FitzPatrick Nuclear Power Plant.Sampling locations are chosen on .the basis of historical average dispersion or deposition parameters from both units. The environmental sampling location coordinates shown on Table 5-11 are based on the NMP-2 reactor ccnterlinc. The average dispersion and deposition parameters .have been calculated for a 5 year period, 1978 through 1982. These average di.,per.ion or deposition parameters for the site are used to compare results of the annual land use-census. If it is determined that sample locations required by Part.I are unavailable or new locations'are identified that yield a significantly higher(e.g. 50%) calculated D/Q value, actions will be taken as required by Controls DLCO. 3.6.20 and DLCO 3.6.22, and the Radiological Environmental Monitoring program updated accordingly.

4.2 Interlaboratory

Comparison Program Analyses shall be performed on samples containing known quantities of radioactive materials that are supplied as part of a Commission approved or sponsored lnterlaboratory Comparison Program, such as the EPA Crosscheck Program. Participation shall be only for those media, e.g.,air, milk, water, etc., that~are included in the Nine Mile Point Environmental Monitoring Program and.for whichCrosscheck samples are available. An attempt will be made to obtain a QC sample to program sample ratio of 5% orbetter. The site identification symbol or the actual Quality Control sample results shall be reportedlin theAnnual Radiological Environmental Operating Report so that the Commission staff may evaluate the results.Specific sample media for whichEPA Cross Check Program samples are available include the following: -gross beta in air particulate filters-gamma emitters in air particulate filters-gamma emitters in milk-gamma emitters. in water-tritium in water-1-131 in water Unit I ODCM Revision 28[H 33 Septembher 2006

4.3 Capabilities

for Thermoluninescent Dosimeters Used for Environmental Measurements Required detection capabilities for thernmolurminescent dosimeters used for environmental measurements required by Table D 4.6.20-1, footnote~b of Part I are based on ANSI Standard N545, section 4.3. TLDs are defined as phosphors packaged for fielduse. In regardtothe detection capabilities for thermoluminescent dosimeters, only one determination is required to evaluate the above capabilities per type of TLD. Furthermore, the above capabilitiesrmay be determined by the vendor whosupplies the TLDs. Required detection capabilitiesare as follows: 4.3.1 Uniformity shall be determined'by giving ThDs from the same batch an exposure~equal to that resulting from an exposure rate of 10 mR/hr during the field cycle. The responses obtained shall have a relative standard.deviation ofless than. 7.5%. A. total of at least 5 TLDs shall be evaluated.

4.3.2 Reproducibility

shall be determined by-giving TLDs repeated exposures equal to that resulting from an exposure rate of 1.0 uR/hr during the field cycle. The average of the Trelative.,standard deviations of the responses shall be less than 3.0%. A: total of at least 4 TLDs shall be evaluated. 4.3.3' Dependence of exposure interpiretation, on the length of a field cycle shall be examined by placing TLDsfor a period equal to at least a field cycle and aperiod equal tohalf the same field cycle in an area where the exposure rate is known to be constant. This test shall be conducted under. approximate average winterltemperatiires and approximate average summer temperatures. For these tests, the ratio of the response obtained: in the fieldcycle.to twice that obtained for half the field cycle shall not be less than 0.85. Ati least 6 TLDs sliall be evaluated.

4.3.4 Energydependence

shall be evaluated by the response of TLDs to photons for: several energies between. approximately 30 keV and .3 Me.V. The response shall not differ from that obtained with the calibration source by more than 25% for photons with energies greater than 80 keY and shall not be enhanced. by more than a factor of two for photons with energies less than 80 keV. A total, of at least 8 TLDs shall be.evaluated. 4,3.5 The directional dependence of the TLD response shall be determined by comparing the response of the TLD exposed in the routine orientation with. respect to the calibration source with the response obtained for different orientations., To accomplish this,theTLD shall be rotated through at least.two perpendicular planes. The response averaged over all directions shall. not differ from the response obtained in the standard calibration position by more than, 10%. A total of at least 4 TLDs shall be:. evaluated.

4.3.6 Light

dependence shall he determined by placing TLDs in the field for a period equal to the field cycle under the four conditions found in ANSI N545, section 4.3.6. The results obtained for the unwrapped TLDs shall uot differ from those obtained for the TLDs wrapped in aluminum foil by more than 10%. A total of at least 4 TLDs shall be evaluated for each of the four conditions. Unit 1 ODCM Revision 28 n 34 September 2006

4.3.7 Moisture

dependence shall be determined by placing TLDs (that is, the phosphors packaged for field use) for a period equal to the field cycle in an area where-the. exposure rate is. known to be constant. The TLDs shall be exposed under two conditions; (1) packaged in a thin,.sealed plastic bag, and (2) packaged in a thin, sealed plastic bag with sufficient water to yield observable moisture throughout the field cycle. The TLD or pho0sphor, as appropriate, shall be dried before readout. The, response of the .TLD: exposed inthe plastic bag containing. water shall not differ from that:exposed in the regular plastic bag by more than 10%. A total of at least 4TLDs shall be .evaluated for each condition. 4.3.8 Self irradiation shall be determined by placing TLDs for a period equal to. the field cyclein an area where thecexposure rate is less than .10 uR/hr and the exposure during.the field cycle. is known. If necessary, corrections shall be. applied for the dependence of exposure interpretation'onthe length of the field cycle (ANSI N545, section 4.3.3). 'he average exposure inferred from the responsesof the TLDs shall not differ from the known exposure by more than an exposure equal to that resulting from an exposure rate of 10 u.Rhr during the field cycle. A total of at least 3 TLDs shall be evaluated. Unit I ODCM Revision 28 T1 35 September 2006 TABLE 1-1 Average-Energy Per Disintegration ISOTOPE Ar-41 Kr-83m Kr-85 Kr-85m IK-87 Kr-88 Kr-89 Kr-90 Xe-131M Xe-133 Xe-133m Xe-i35.Xe-135m Xe, 137 Xe-138 Eymev/dis:1.294 0.00248 0.0022 0.159 0,79.3 1.95 12.2 2. ! G 0.0201 0.0454 0:042 0.247 0.432 0. 94!'ls (Ref)(3)(l) 0.0371 (1)G!)(1)(1)(2)(2)(1)(I)(1)(*J)(13.(1)(1)Eimev/dis(4) 0.464 0.250 0.253 1.32 0.377 1.37 1.01 0.143 0.135 0.19 0.317 0.095 1.64 0.61.1 (3)(1)(1)(1)(1)(.1)(2)(2)(1)(1)(1)(1)(1)(1.)(.1).(2)ORNL-4923, Radioactive Atoms -Supplement I, M.S. Martin, November 1973.NEDO-12037, "Summary of Gamma and Beta Emitters and intensity Data". M.E. Meek, R.S.Gilbert; January 1970. (The.average energy was computed from the maximum energy using the!CRP 11 equation, not the 1/3 valueassumption used in.this reference). NCRP Report No. 58, "A Handbook ofRadioactiVity Measurements Procedures"; 1978 The average energy includes conversion electrons. (3).(4)Unit I DCM Revision 28 September 2006 1136 NUCLIDE H3 Cr51 Cu 64 Mn 54 FE 55 Fe 59 Co 58 Co 60 Zn 65 Sr 89 Sr. 90 Zr 95 Mn 56 Mo 99 Na 24 1 131 1 133 Ni 65 1 132 Cs 134 Cs 136 Cs 137 Ba 140.Ce 141 Nb 95 La 140 Ce 144 BONE 1.31 E+l 2.8.4E-+1 1.72E+1 2.32E+3:1:74E+4 1.91E-i 2.37 3,03E+1 4..22 1.33E-3 1.58E-4 3.54E+2 4.05E31 4.91 E+2!.:50E+2 7.21 E-2 1,85E-2 L18E-2 2.79 TABLE 2-1 Ait VALUES -LIQUID*RADWASTE TANK INFANT torem.- ml hr- gCi LIVER T BODY THYR(2190E-1 2.90E4 1 2;90E-1 1- 1.219E-2 8.39E-3 1.13E-1 5.2313-2 I .87E+1 4.23 ..8.44 2.26 --4.96E-1+ 1.96E+1 --3.34 8.34 -1.02E+1 2.40E+1 -5,911321 2,73E+1 ---- 6.66E+1 -"- 4.43E+3 --4.66E-2 3.30E-2 --2.40E1-4 4.115E2- -2.34E+1 4.57 --2.37 2.37. 2.37 3.54E+1 E157E+1 1.17,E+4 6.15 1.80 .112E+3!.51E-,4 6.85E1-5 --3.21E-4, 1.14E-4 .1.501E-2 6.60E-+2 --1.19E24) 4,4513+1 -1 5.75E+2 4.07E+1 --11506E-1 7,74 --4.40E-2 5.17E13 -1.591E-2 9.18E.3 --4.67E-3 1.20E-3 --1.14 1.5712-1 --ODI KIDNEY 2.90E-1 1.83E23 1.91 E-1 4.14 NNW LUNG 2.90EI 1.63E-2 4.13 1.47E121 7...3 2.87E+1 5.02E-2 2.007E-4 3.50E+1 2.37 4.17E+ 1.7.23 3-.58E"4:1.70E+2 4.75E+1 1.54E+2 3157E-2 136ME-2 1.14E-2 4162E-.1 GI-TRACT 2.910E-1 3.75E-1 2.32 6.86 I.07 2.37E1+1 8.33 2.4213+1i 5.00E+I 4.77E+1, 2.1,7E+2 2.32E+1 2.118E-2 7.71 2-37 1.28 1.04 1.1 5E-2 2.60E-4 1.79 1.80 3.69E+1 2.27E÷+1.34E÷1 5.48E+1 1.60E+2 2.37 6.97Et1 9.7 1E+1 6.24E13+9.23E-2..* Calculated in accordance with NURFIG 0133, Section 4.3.1; andRegulatory.Guide 1.109,'Regulator, position C, Section 1.Unit I ODCM Revision 28 11 37 September 2006 TABLE 2-2 Ai 8 t VALUES -LIQUID*RADWASTE TANK, CHILD mrem -ml hr-- gCi NUCLIDE H3 Cr51 Cu 64 Mn 54 Fe 55 Fe 59 Co 58 Co 60 Zn 65 Sr 89 Srgo Zr 95 Mn 56 Mo 99 Na 24 1131 1133 Ni 65 1:132 Cs 134 Cs 136 Cs 137 Ba 140 Ce 141 Nb 95 La 140 Ce 144 BONE 2.13E22 2:5. E-6 6.92 9.21E+2 1.30E+3 1.89 1.12E'2 2,15E+4 3.261+4 4:26E+5 1.70 5.35E-3 1. 52232 2109E+2 3.39E+1 2.67E-1 6.13E-3 3.68E+5 3.52E+4 5.15E+5 3.61 E+2 1.50E- t 5.221E+2 1.50E-1 5.00 LIVER 4.39E-1 2. 13 B2.2.70 3.38E+3 4.88E+2 2.11E-.3 7.46E+1 3.28E22 5. 73E÷ 4 1 AOE:4 1.33 1.65E2-1 9.57E+1 I ,522+/-2 2.10E22 4.19F+1 2:51E-2 1.131E-2 6.04E+5 9.67E+4 4:93E+5 3.96E-1 1.07E-1 21.03E+2 5.93E-2 1.81 T BODY 4.3 9E- 1 1.40 1.63 9..06E+2 1.51E22 1.05E+3 2.24E+2 7.48E+2 3.56EE-4 9.32E+2 1.08E+5 1.32 3.73E-2 2.37E+1 1.522+2 1.19E+2 1.59E+1 1.47E-2 5..18E-3 1.27E+5 6-26E64 7.28E+4 2.112+1 6.99E-2 1,.45E+2 2.68E-2 THYROID KIDNEY LUNG GI-TRACT 4.392E-1 7.86E-1 2.51E-6 6.92 1.34 1.89 1.1213+2 3.85:1.10E-4 1.23 5.35E-3 1.:52E+2 6.94E+4 7.78E+3 5.22E21 3.54E+l 6.21F-l 5.372E1 7196E-2 6.34E-2 6.39E-1 1.03E-2 4.391E-1 2.30E-1 6.52.9.53E+2 1.34 1.89, 1.12E+2 3.6 1E+4 1.102-4 1.38.2.00E-1 2.04E+2 1-5213+2 3.45E+2 6,98E+1 1.72E-2 1.87E2-5 5.15E-ý4 1.6 iE÷5 1.82E-1 8.24&.2 1.91E+2 1.03E-2 1.16 4.39E-1 1.42.5 1.E-6 6.92 2,76E+2 6.12E+2 1.89 1,12E+2 1.10E-4 1.23 5.351-3 1.52÷+2 5.60E-2 1.38E-4 6.72E+4 7.68E+3 5,78E+4 2.68E-1: 6.34E2-6.39E-1 1.03E-2 4.3.9E-1 7.3 IE 11 1.27E+2 2.84E+3 9.05E+1 2.19E+3 4.261+2 1.31E+3 1.01E+4 I 26E+3 5.74E+3 1.08E+2 2.39E+1 7.9 131-1.52E+2 S.'87E+ 1 1.693+1 3.08 1.32E-2 3.29E 13 3.40E+3 3. 1 4E+3 1.83E22 3.75E+5 1.136E3 6.06E-1 3.58E-.1 3.58E-1 .3.80E+2* Calculated in accordance with NUREG 0133, Section 4.3.1; and Regulatory Guide 1.109, Regulatory position C, Section 1.Unit 1 ODCM Revision 28 1138 September 2006 TABLE 2-3 Ai, VALUES -LIQUID*RADWASTE TANK TEEN mrem -ml hr -ICi NUCLIDE H3 Cf 51 Cu 64 Mn 54 Fe 55 Fe 59 Co 58 Co :60 Zn 65 Sr 89 Sr 90 Zr 95 Mn 56 MO 99 Na 24 1 131 1 133 Ni 65.1 132 Cs 134 Cs 136 Cs,137 Ba 140 Ce 141 Nb 95 La 140 Ce 144 BONE 1.02EA1-i.20E-5 3.331E+1 6194E+2 1.07E+3 9.03 5.36E+2 2.,10E+4 2.44E+4 4.66E+5 6.20 2.56&-2 1.39EAL2 1.55E+2 2.53E+1 2.08E2-1 4.90E-2 3.05E-5 2.98E÷+4 4.09E+5 2.35.E÷2 3.46E1-1 4,44E+2 1.57E-i 3.99 LIVER 3.28E-1,:1.02E-1, 2.89 4.34E+3 4.92E+2 2.49E+3 9,82E134 7.96E+2 7 .28F2+4 5.24E,-4 6.00 1.8 IE-I 9.22E111 I .39E+2 2.1 7.E+2 4.29÷12+2.666-2 1.28E-2 7.18E1J+5 1. 1713+5 5.44E+ 5 4.1013-1 3.32E2-t 2.48E+2 1.02E-1 2.65 T BODY 3.281E-1 1.39 1.36 8.87E+2 1..15E12 9.64E+2 2.15E+2 1. 12E4 3 3.40F+4 6.98E+2 1. 1513+5 5.97 3.22E-2 1.76E+1 1.39E+2 1.1613+2 1.3 1 E+1 1.21E2-2 4.60E-3 3.3'3E-+5 7.88E1-4 I V90E+5 1.5513+1 3.07E11 1. 1813+2 6.35E-2 1.83 THYROID 3,28E-1.8.16E7 I 1.20E-5 3.3112+ 1 6.41 9.03 5.36E+12 1.84E1+5.24E-4 5,.90 2.56E-2 1.39E+2 631E+4 5.99E+31.69E+2+/-2.97ý2.517E+2 3.81 E-1, 3.0413-3.06 4.94E-2 1.71 KIDNEY 3.28E1-3,84E-1 7.32 1.32E+3 6.4.1 9.03 5.36E+2 4.66E+4 5.214E-4 6.04 2.29E-1 2. H1E+2 1.39t+2 3.73 1E+2 7.5213+1 2.02E-2 2.28E+5 6.38E+4 1.8513+5 4.79E-1 3.1 7E71 2.40E12 4.94E-2 2.27 LUNG 3.28E- 1 1.94 1.20E-5 3.3 11+ 1 3.12E+2 7.8913+2 9.03 1.8413+1 5.24E-4 5.90 2.56E-2 1.39E12 2.68E1-6.60E-4 8.7313+/-4 1.01E+4 7.21E+4 5.75E-1 3;04E-1 3.06 4.94E--2 1.71 GI-TRACT 3.2 SE-I 2.1613+2 2.24E+2ý886E÷+3 2.1313+2 5.87E+3 1.24E+/-3 3.4931+3'3.08E+4 2.910F+3 1.311E+4 2.28E+2 1.19E+1'I.65E+2 1.3 9E+2 4.301+÷I 3.2513+1 1.,44 5.59E-3 9.1012+3 9.4413+3 7.99E+3 13.63E+2 8.16E+1 1.05E+6 3.05E+3, 5.74E2+* Calculated in accordance-with NUREG 0133, Section 1.Section 43.1; and RegulatorY Guide 1.109, Regulatory position C, Unit 1. ODCM Revision 28 U 39 September.2006 TABLE 2-4 Aig VALUES -LIQUID*RADWASTE TANK ADULT torero-ml hr -paCi NUCLIDE H3 Cr 51 Cu 64 Mn 54 Fe 55 Fe 59 Co58 Co 60 Zn 65 Sr 89 Sr 90 Zr 95 Mn 56.Mo 99 Na 24.113.1 1.133 Ni 65 1132 Cs 134 Cs 136 Cs 137 Ba f40 Ce 141 Nb 95 La 140 Ce 144 BONE 1.82E-2 5.94 6..64E+2 1.03E+3 1.62 9.60Et+1 2.31E+4 2:25E+4 5.60E+5 1.36 4.58E-3 1.35F+2 1.45E+2 23 5EE+ I i.93E-1 4.68E-3 2.98E+5 2.96E+4 3.82E+5 2.24E+2 9.53E-2 4-39E+2 I. I E-1 2.48 LIVER 4.45E-1 1.82E-2 2.75 4.3 8E+3 4.58E+2 2.43E+3 9; 15E+1 2.57E+42 7.36E+4 9.39E-5 1.15 1.73E.-1 8.70E+1 1.35E+2 2.07E+2 4.09E+1 2.51 E-2 1 .25E-2 7.08Et ;5 1.17E+5 5.22M+5 3.49E-1 ,8.20E-2 2.44E+2 6.03E-2 1.22 T BODY 4.45E- 1 1.27 1.29 8.41 E÷2 I.07E+2 9.3 1 E+2 2.03E+2 6.711 E2 3.32E+4 6.4 5E+2 I..37E+5 1.12 3:07&E2 1.66E+I 1..35E+2 1.1 9E+2 IL25E÷1 1.14E-2 4.38E-3 5.79E+5 8.42E+4 3 42+/-+5 1..47E÷!5.75E-2 I 32E+2 2.24E,2 4.24E-1 THYROID KIDNEY LUNG, GI-TRACT 4.45E4 7.64E-1 5.94 1.15 1. 62 9.60E+ I 3.30 9.39E-5 1.06 4.58E-3 1.35E+2 6.79E84 6.02E+3 4.38E-1 3.03E+1 5.32E-4 4160E4i1 6.83E-2 5.44E-2 5.47E-1 8.84E-3 3.07E-1 4.45E-1 2.93Er-6.94 1.3 IE+3 1.15 1.62 0.60E÷1 4.92E+4 9.39E-5 1.21 2:.20E-:1 1 .97E+2 1.35E 2'3 .55E+2 7.14E+1 2.OOE-2 2.29E+5'6.51E+4 1.77E+5 1.64E&1 6.72E-2 2.4 iE-t2 8.84E-3 9.47E-1 4.45E-1 ,!1.67 5.§4 2.56E+2 6,79E+2 1.62.9;60E4-I 3.30, 9.39E-5 1,06 4.59SE-3 1.35E+2 480E-2 1.1 8E-4 7.61E-A-4 8.93,E+3 5.90E+4 2.29E- I 5.44E-2.5.47E- 1 8.84E-3 3.07E- 1'4.45E-1 3.1 4E+2 2.35E+2 1.34E+4 2.63E+2 8.09E+3 1L82E+3 4.99E+3 4,63E+4 3.60E-3 ,1.62E-+4 3.06E+2 5.52 2.02E-2 1.35E+2 5.47E+ 1 3.68E+1 6.36E-1 235E-3 1.24E+4 1.33E+4 1,02E+4 4z61E+2 1.06E+2 1.48E+/-6 3.78E+3 7.37E+2* Calculated in accordance with NUREG 01331 Section 4.3.1;Regulatory position C, Section l.and Regulatory Guide 1.109, Unit I ODCM Revision 28 September.2006 1140 TABLE 2-5 Aiat VALUES -LIQUID*EMERGENCY CONDENSER VENT INFANT NUCLIDE H3 Cr51 Cu 64 Mn 54 Fe 55 Fe59 Co 58 Co 60 Zn 65 Sr 89 Sr 90 Zr 95 Mn 56 Mo99 Na 24 1 131 1 133 Ni 65 1 132 Cs 134 Cs 136 Cs 137 Ba 140 Ce 1,41 Nb 95 La 140 Ce 144 BONE 3.35E-2 7.29E-2 4.42E-2 5;95 446E+1 4.90E-4 6.07E-3 7.77E-2 1.08E-2 3.41E-6 4.05E-7 9.081- 1 1.04E- 1 1.26 3.85E-1 L1.8513-4 9,88E-5 3.03E-5 7-16E-3 mrem -ml hr -ICi LIVER T BODY 7T3E-4 7.43E-4-- ..330E-5 2.89E-4 1.34E-4 4.79E-2 1. 08E-2 2.16E-2 5.78E-3 I 271E- 1 5.02E-2 8.58E-3 2.14E-2 2.60E-2 6.15E-2 1.52P-1 6.99E-2-- 1.71E-1-- 1.14E+lI 1. i 9E-4 8.417E-5 6.17E-7 1.06E-7 6.00E-2 1.1 7E-.2 6.07E-3 6.07E-'3:9.166E-2 4 403E-2 1.58E-2 4,62E1-3 3.86E-7 1.76E-7 8.22E-7 2.93E-7 3 1.69 1.711E-1 3.06E-1 1.14E-I1 1.47 1.04E- I 3.85E-4 1.99E-2 1. 13E4 1.33E-5 4.07E-5 2.35E-5 1.20E-5 3.0813-6 2.93E-3 4.02E-4 FHYROID 743E-4 2.1SE-5........... ........... KIDNEY 7.43E-4 4.7013-6.4.89E-4 1 .06E-2 6,0713-3 3.101E+ 1~.87 1.85E3-5 7.35E-2 1.29E-4 5.30E1-7 98.97E-2 6.07E-3 1: k7E-I 1.85E,2: 9.17E-7 4.365-1:1.22E-1 3195E-1 9.15E-5 3 .48E-5 2.92E-5 1.19Ek-3 LUNG GI-TRACT 7,43E-4 7.43E-4 4..18E-5 9.61E-4-- 5ý94E-3-- 1;.76E-2.1.,06E-2 2.751E-3 3.76E-2 6.0813-2-- 2.14E&2-- 6. 1 9E-2-- .1.28E-I-- 1.221E-1-- 5.57E-1-- 5.95E1-2-- 5.60E-5-- 1.99D-2 6.07E;33 6.07E3-3-- 31.271-3-- 2.67E-3-- 2.94E-5-- 6.66E-7 1.79E-1 4.60E-3 ,2.49E-2 4.64E-3'1.60E-I 4.61E33 2.37E-4 9.47E-2-- 5.82E-23.43E-2-- 1.41E-1-- 4.1.1EI Calculated in accordance with NUREG Oi 133, Section 4.3.1 .and Regulatory Guide I .109. Regulatory position C, Section 1.Unit 1 ODCM Revision 28 Ii 41 September 2006 TABLE 2-6 Aj., VALUES -LIQUID*EMERGENCY CONDENSER VENT'CHILD mrem -Ml hr -pCi NUCLIDE BONE LIVER T BODY THYROID KIDNEY LUNG GI-TRACT H3 -- 1.44L1 t.44E51 1.44E-1 .1.44E--1 1.44E-1 1.44E4t Cr 51 3.78E-5 3.78E-5 1.37 7.58E,-1 2.07E-1 138 7.24E+I Cu64 -2463 159 -- 6.35 -- 1.23E 12 Mn 54 1.23-2 .2 1.23F-2 82+3-5,36F÷3 S.95F÷2 1:.27F-2 9-42E÷+2 12E2 Z2 Fe 55 9.04E÷r2 4.79E+2 1.49E+2 -- -2.71 E+2 8.88E1+I Fe, 59 1.28E13 2.017E+3 1.013E+3 2.38E-3 2.388E-3 6.00E÷2 2.155E+3 Co58 3,36E-3 7.01E+1 2.15E+2 3.36E-3 3.36E-3 3.36E-3 4.09E-+2 Co 60 1.99E,1-] 2.08E12 6,14E+2 1.9964. i.99E.4 1.99E-1 1.1 5E+3 Zii 65 2.15E+4 5.73E+-i4 3.56E+4 6.84E-3 3.61E-,4 6.84E-3 1.01E,14*Sr 89 3.07E+4 -- A7 7F.+2 ....... 1.19E5+3 Sr90 4.01E+5 -1.02E+5 ........ 5,40E+3 Zr95 3.01E-1. 6.78E-2 6.06E-2, 2.191E-3 9.6iE-2 2.19t-3 6.84E+1 Mn 56 -- i.65E-1 3.73E-2 -- 2.00E-1 -- 2.39E1..Mo 99 -- 8J16E-I 2.02E+1 -1.74E+2 -- 6.75E+1 Na 24 1.50E+2 150E+2 1,50E4,2,. 1.50E 12 1.50E1:i2 1.50E112 1.501E+2 1131 1,86,E+2 1.87r.+2 1.06E-+2 6.19E-,4 3.08E÷2 -- 1.67E+1 1133 3.08E1+ 3.81E+I -1144E1l 7,07E÷3 6.35E÷1 -- 1.53E+1 Ni 65 2.66E-t 2.50E-2 1.46E1-2 ......3.07 1 132 6.01E-3 1.103E-2 5.08E-3 5.12E-1 1.69E-2 : .1.30E,2 Cs 134 3.68E+5 6.04E+5 1.27E+5 6.29E-2 1.87E+5 6.71E+4 3.25E+3 Cs 136 3151Ei 4 9.66E+4 6.25E+4 1.10E,3 5.14E+4 7.67E+3 3.40E-+3 Cs 137 5.14E15 4.92E+5 7.27E+4 9.55E-2 1.60E+5 5.77E+4 3.08E13 Ba 140 2.48E+2 2.173E-1 1.45E+1 1.42E-4 :7.09E-2 1.30E,-1 1.26E+2 Ce 141 3,08E-2 1.54E.2 2.39E-3 1.13,84 6.83&33 .:13 E-4 1.91E+1 Nb 95 5.21E+2 2.03E+2 1.45E+2 1.14E-3 1.90E+2 1.14E1-3 3.7513+5 La 140 1.31E-1 4.59E-2 1.555E-2 1.83E-5 1.83E-5 I.3E-5 1.28E1+3 Ce 144 1.64 5.15E-1 8.811E-2 6.36E1-4 2.85E-1 6.366E-4 1.34E+2* Calculated in accordance with NUREG 0133, Section 4.3.1; and :Regulatory Guide 1.109, Regulatory' position C;Sec~tion 1.Unit 1 ODCM Revision 28 1 42 September 2006 TABLE 2-7 Aat VALUES -LIQUID*EMERGENCY CONDENSER VENT TEEN mrem -ml hr -ttCi NUCLIDE BONE LIVER H3 Cr51 Cu 64 Mn 54 Fe55.Fe 59 Co 58 Co 60 Zn 65 Sr 89 Sr 90 Zr 95 Mn 56 Mo 99 Na24 1 131 1 133 Ni 65 1 132 Cs 134 Cs 136 Cs 137 Ba 140 Ce 141 Nb 95 La .140 Ce 144 1.81E-4 5.89E-2 6.89E+2 1.05E+3.611E-2 9.53E- I 2.101E+4 2.38E+4 4.54E-+5 2.56E-:1 1.38E÷+2 1.47E+2 2.42E+ I 2.08E-1 4.86E-3 3.05E+5 2.98E+4.4,09E+5 1.9613+2 2.431E-2 4.41 E+2 1-05E-1 1.27 1.74E-1 1.8.1E-4 2.86 4.29F,+3 4.88E+2, 2.46E+3 8.78E+1 2.57E+2 7.28E+4 8.80E3-2 1.81E-1 8.57E4+1 1.238E42 2.06E+2 4.11 E+1.2.66E-2 1.27E-2 7.18E-4-5 1.17E3+5 5.44E+5 2.4 7E-2 1.64E-2 2.45E+2 5,17E-2 5.28E-1 T BODY 1.74E-1 1.28 L35 8 .52F+2 1.14E+2 9.50E+2 12.02E-2 5i78E+2 3.39E 4 6. 8.1 FA2 1.12E÷5 6.38E-2.3.22E-2 1.631E+1 1.381+2, 1.10E3+2 1.25E1+1 S.21E-2 4.56E-3 3.33E+5 7.8713-4-4 1.89E+5 1'.127E+1 2.36E-3.1 15E+2 1.380r2 7.12E-2 THYROID L.7413-1 7.12E-1 5.893-2 1.14E-2 1.61 E-2 9.53E-1:3.28E2!1 .05E-2 1.38E4-2 6.00E,+4 5.74E+3 4.29E-1 3.01E-I 5.28E-3 4.57E-1 6.77E-4 5.40E-4 5.43 E-3 8.78E-5 3.04E133 KIDNEY LUNG GI-TRACT 1.74E-I1 24818-1 7.24 t1.28E#+3 1. 1.4E-2 1 .61E&2 9.533- 1 4.66E j 4 1.24E1-.!2.29E-1 1.96E+2 1.38E4-2 3.514E+2 7.21E+1 2.OOE-2 2-28b+5 6.38E+4 1 f85E+5 8.23E-2 8.02E-3 2.37E+2 8.781E-5 3.17E-t 1,74E-1.1.:83 5.89E-2 3 10E+2 7.76E+2 1.61 E-2 9.53t3-1 3.28E-2 1.0513-2 1 ;38E+2 4.77E-4 8.7 1 E+4ý827IE+4 1.01E÷+/-4 7.191E+4 1.62E--1 5.40E-4 5.43E-3 8,78E-5 ,3;04E-3 t .74E1-.2.,15E+2'2.22E+2 8.811E+3 2,11,11+2 5.82E+3 1.211E+3 3.34E-+3 3.08&,-4 21831E+3, 1,27E+4 1.79E+2 1. 19E+1 1.54E+2 1.38E+2 4.07E+1 3.1i1E+l 1.44 5.54E-3 8.93E+3 9.43E+3 7.73E+3 3.03E+2 4.54E+1I 1.10513+6 2196E+3 3.19E.+2* Calculated in accordance with NVUREG 0133, Section 4.3.1; and Regulatory Guide 1.Section 1.109, Regulatory position C, Unit 1 ODCM Revision 28 September 20066'1143 TABLE 2-8 A;,t VALUES -LIQUID*EMERGENCY CONDENSER VENT ADULT 'NUCLIDE H3 Cr 51 Cu 64 Mn 54 Fe 55 Fe 59 Co058 C o'60 Zn 65 Sr 89 Sr 90*Zr 95 Mn 56 Mo 99 Na'24 I 133 Ni'65 1 132 Cs 134 Cs 136 Cs 137 Ba 140 Ce 141 Nb 95 La 140 Ce 144 BONE mI1 " 11 .. .. ......3.24E-5 1.06E-2 6.58E+2 1.02E+3 2.88E-3 1.71 E-41 2.31 E4-4 2.18E+4 5.44E+5 2.40E3-1 1.34E+2 1,31E4-1 2.25E+1 1.93E- I 4.64E-3 2.98E135 2.96E+4 3.82E+5 1.84E+2 2.21,E-2 4.38E+2 9.90E-2 1.17 LIVER 2.27E13.3.24E1-5 2.72 4,37E+3 4.55E+2 2.411E+3 8.83E-+1 2.56E+2 7.36&÷4 7.8 1E-2 1.73E-1 8&04EI1 1.34E+2 1.96E÷122.50E-2 1.24E-2 7.08E+/-5 117E+5 5.22E+5 2.321-1 1.50E-2 2.44E+2 4.99E-2 4.890-1.mrem:- ml hr -gCi T BODY 2.27E-:1 1.24 1.28 8.33.E+2 1.06E+2 9.22E+2 1 .98E3+2 5165E+2.3.3 2E+4.6.27E+2 1.34E+5 5.3512-2 3.07E-2 1.53E+1 I -34E+2 1.12E÷+2 1.19E÷1 1.141E-2.4.34E-3 5.7913+5 8.42E+4 3.42E+5 1.2113+1 1.78.1-3 1.3 11E+2 1.3213-2 6.33E-2 2.27E-1 7.43E4 1.06E1-2 2.04E-3 2.88E-3 1.7!1E-5.87E3-3 1.88E-3 134E+2 6.43EP4.5.75E+3 4.34E-1 5.39E-2 9.46E-4 8.19E,2 1.21 E-4 9.67E-5 9.73E-4 1.571E-5 5.45.-4 2.27E-I 2.74E-1 6.86 1.30E343 2104E-3 2.88E-3 1.71E-1 4V92E+4 1.22E-1 2.20E-1 1.82E1-2 1-34E+2 3 36E+2 6.82E+1 1.98E-2 2.29E+5 6.51iE+4 1.77E+5 7.88E-2 7.00E&3 2.41 E+2 1.57E-5 2.90E-1 2.2713-1 2.27E-1 1.65 3.1213 12-- 2321E+2 1.06E-2 1.34E+4 2.54E+2 2.61 E+2 6.7213+2 8.02E+3 2.8813-3 1.79E+3 1.7115-1 4.88-1E+3 5187E-3 4f63E+4-- 3.5013+3-- 1.5713+4 1 .88E-3 2.42E+2-- 5.52-- 1.86E+2 1.34E+2 1:34E+2-- 5,17E+1-- 3.51E+1-- 6.36E-1-- 2.33E-3 7.61EE+4 1.2413+4 8.92E+3 1.3'.-+4 5.89E+4 1.013E+4 1.33E- I 3.79E+2 9.67E-5 5.68E+1 9.731E-4 1,48E+6 1.57E-5 3;66E+3 5451E-4 3.95E+2 THYROID KIDNEY LUNG GI-TRACT Calculated in accordance with NUREG 0133, Section 4.3.1; and Regulatory Guide 1.109, Regulatory. position C, Section 1.Unit I ODCM Revision 28 TI 44 September 2006 LOCATION.Residences Dairy Cowst Milk Goats, Meat Animals Gardens Site Boundary DIR E (980)SE (130*)SE(130()ESE (1150)E (970)ENE (6.7')MILES 1.4 2.6 2.6 1.8: 1.8 0.4 TABLE 3-1 Critical Receptor Dispersion Parameters' For Ground Level and Elevated Releases ELEVATED ELEVATED X/Q (!sechm') .,2 ('1.8 E-07b 5.2 Eg09b 2.2 E-08c 7.0 E-10c 2.2 E-08c 7T0 E- I 0C 5.1 E-08e 1.7 E-09c 1.0 E-07 0 3.5 E-09'C 2.4 E-0 6 b"d 4.4 E,08'GROUNDe X/o(sec/mn 3)4.02 E-07 6.00 E-08 6o.0 E108 1.16 E-07 2.53 E-07 6.63 E-06 GROUNDe D/Q (m 1 8.58-E-09 11.64 E709.1.64 E-09 3.54 E-09 5.55 E-09 6.3.5 E-08 a. These values will be used in dose:calculations beginning inApril 1986 but maybe revised periodically to account for chianges in locations of farms, gardens or critical residences.

b. Values based on 5 year annual meteorological data (C;T. Main, Rev. 2)*c. Valuesbased on 5 year average grazing season meteorological data (C.T. Main Rev. 2)d. Value are based on most restrictive-X/Q

!and-based.sector (ENE). (C.T. Main, Rev. 2)e. Values are based on average annual meteorological data for the year 1985.f. Conservative. location based on past dairy cow and goat milk history,.Unit I ODCM Revision 28 1145 September 2006 TABLE 3-2 Gamma Air and Whole Body Plume Shine Dose Factors*For Noble Gases Nuclide Kr-985 Kr-85m Kr-87 Kr-88 Kr-89 Kr-83m Xe-133 Xe-133m Xe-I 35 Xe-135m Xe-137 Xe-138 Xe-1 31 m Ar-41.GAnmaAirB 1 xnradlyr 2.2 3E-;6 S.'75E-3 1.02E-2 2.23E-2.2.50E-2 2.26E-6 19.1 E-4 2.27E-4.2.62E--;3 5.20E-3 2.30E-3 1 .54E-2 1,.74E-15 I .64E-2" Gamma Whole Body V, m rem/vr PCi/s5C 1.68E-3 9.65E-3 2.17.E-2 1.71E-2 1.75E-4 1.871-4 2.50E-3 4.89E-3 2.20E-3 1I.03E-2 1.47E-6 1 .57E-2* Calculated in accordance with Regulatory Guide 1,109. (See Appendix B.)Unit 1 ODCM Revision 28 September 2006 Il 46 TABLE 3-3.IMMERSION DOSE FACTORS FOR NOBLE GASES*Nuclide Kr 83m Kr 85m Kr 85 Kr 87 Kr 88 Kr :'89 Kr 90 Xe ,13 Im Xe 133m Xe 133 Xe 135m Xe 135 Xc 137 Xe 138 Ar 41 KAf~mBody) 7.56E-02 1. 17EI L61E1 5M.9E3 1.471E4 1.66E4 1.56E,4-9.15EI 2.5 1 E2 1.94.E2 3.112E3, 1.~8 1F3 1..42E3 8.83VE3&.84E3 1.46E3 1 .341M 9.173 E3 2.3 7E3-1 X61E4.7.29E3 4.16E2 9914132.3.06E2, 7.J.1. E2 1.212E4 4. 13E3 2.69E3 aM(y:Air)* 1 23M3 6.17133 I 52E4 1.73E4 1..63134 S.156E2 3 .27E2 3.;53E2 3 .36E3*.1 .92E3 1.51133 9.21 E3 9.30E3 2.88E2 1.97E3 I 95E33 1.63E4 2.93E3 1.06E4 7'.83E3 1. 11 1E3 1.48E3 1.05E3 7.39E2.2,46E3 I.27E4 4.75E3 3.28E3* From, Table B-I.Regulatory Guide 1.109'Rev. 1* mremlyr per gCi/mr.** mrad/yr per 1ICi/,M 3.Unit I .ODCM Revision 28 September 2006 I1 47 TABLE 3-4 DOSE AND DOSE RATE Ri VALUES -INHALATION -INFANT'mrem yr.. .C im , NUCLIDE H3 C 14, Cr 51 Mn 54 Fe 55 Fe 59 Co 58 Co .60 Zn 65 Sr 89 Sr 90 Zr 95 Nb 95 Mo 99 1-131 1133 Cs 134 Cs. 137 Ba 140 La 140 Ce 141 Ce 144 Nd 147 BONE 2.65E4.1.97E4: 1.336E4 1.93E4 3.98E5 4.09E7 1.1565 1.5764.3.79E4 1.32E4 3.96E5.5.49E5 5.60E4 5ý05E2'2.77E4 3.19E6 7.94E3 LIVER 6.47E2 5.3183 2.53 E4 t.17E4 2.35E4 1.22E3 8.02E3 6.26E4 2.79E4 6.43E3 l.65E2 4.4E4 1.92E4: 7.03E5 6.1225 5.60E1 2.00E2 1,67E4 1.2 1E6.8.13E3 T BODY 6,47E2 5.3 13 8.95E1 4ý.98E3 3.33E3 9.48E3 1.82E3 1.18E4 3. H E4 1. 14E4 2.59E6 2.03F4 3.78E3 3.23E1 1.96E4 5.60E3 7.45E4 4:5 5E4 2.90E3 5.15E6 1.9963.1.76E5.5.00E2 THYROID 6.47E2 5.3 1 E3 5.75E1 KIDNEY 6.47E2 5.31E3 1.32E1 4.98E3 3.25E4 LUNG 6.47E2 5.3 1E3 1.28E4 1.00E6 8.69E4 1 .;02E6 7.77E5 4.5,1 E61.6A7E5 2.03E6 1.12E7 1.75E6 4.79E5 1.35E5.7.97E4 7.' I I E4 1.606 1.68E5 5.17E5 9.84E6 3.22E5 GI-LLI 6.47E2 5.31E3 3.57E2 7.06E3 1.09E3 2.48F4 1.11E4 3. 19E4 5.14E4 6.40E4 1.31 E5 2ý.7E4 1.27E4 4C87E4 1.06E3 2-16E3 1.33E3 1.33E3 3.84E4 8.48E4 2.1664 1.48E5 3.12E4 I A8E7 3.5686 3.ý11 F4 4.72E3 2.65E2 5.18E4 2.2464 1.90E5 1.72F5 1.34E1 5.25E3 5.38L5 3. 15E3 Th is and f6llowing Ri Tables Calculated in accordance with NUREG 0133, Section 5.3. 1, except C 14 values in accordance with Regulatory Guide 1. 109 Equatitn C-8.Unit I ODCM Revision 28 September 2006 1148 TABLE 3-,5 DOSE AND DOSE RATE R, VALUES -INHALATION -CHILD mremlvr ,Ci/mW NUCLIDE H 3 C 14 Cr 51 Mn 54*Fe 55 Fe 59 Co058 Co 60 Zn 65 Sr 89 Sr 90 Zr 95 Nb 95 Mo 99.1 131 1133 Cs 134 Cs 137 Ba 140 La 140 Ce 141 Ce i44 Nd 147 BONE 3.59E4 4.74E4 2.07E4 4.26E4 5.99E5 1.01E8 1'90E.5 2.35E4 4.81E4 1.66E4 6.51E5 9 .07E5 7.OE4 6,44E2 3.92E4 6.77E6 ,.08E4 LIVER 1.12E3 6.73E3 4.29E4, 2.5224 3,34E4 1.77E3 1.31E4 1.13E5 9.18El I.72E2 4.81E4 2.03E4 1.01E6 8.25E5 6.48E1 2.25E2 1.95E4 2.12E6.8.73E3 T BODY 1.12E3 6.73E3 1.54E2 9.5 tE3 7.77E3 167FA.3.16E3 2.26E4 7.03E4 1.72E4 6.44E6 3.7024 6.55E3 4:26E1 2.73E4 7.70E3 2.25E5 i 28F5 4.33E3 7.55E1 2.9023 3.61 E5 6.81 M2 THYR OID 1.12E3 6.73E3 8.55E1 1.62E7 3.85E6 KIDNEY 1.123 6.73E3 2.43E1 1.00E4 7.14E4 5.96E4 8.62E3 3.92E22 7.88E4 3.38E4 3.30E5 2.82E5 2. 1EM 8.55E3 1.17E6 4.81E3 LUNG 1.12E3 6.73 E3 1.70E4 1,.58E6 1 .27F.6 1.11E6 7.07E6 9.95E5 2.1 6E6 1.48E7 2.23E6 6.14E5 L 35E5-.!.21E5 1.04E5 1.74E6 1,.83E5 5.44E5 1;20E7 3.28E5 GI-LL!1.12E3 6.73E3 1.08E3 2.2 9E4 7-07E4 3A4E4 9.62E4 1.63F4 1.67E5 3.43E5 6. 11 E4 3.70E4 1.,27E5 2.84E3 5A48E3 3.85E3 3.62E3 1.02E5 2.26E5 5.666E4 3.89E5 8.21 E4 Unit I ODCM Revision 28 September 2006 11 49 TABLE 3-6.DOSE AND DOSE RATE Ri VALUES -INHALATION -TEEN mrem/vr llCi/m NUCLIDE H3 C 14 Cr 51 Mn 54 Fe 55 Fe 59 Co.58 Co60 Zn 65 Sr 89 Sr 90 Zr 95 Nb 95 Mo 99.1131 1133 Cs 134 Cs 137 Ba 140 La 140 Ce 141 Ce 144 Nd 147 BONE 160E4 3 .34E4 1.59E4 3.86E4 4.34E5 1.0Et 1.46E5 1.86E4 3-54E4 1.22E4 5-02E5 6.70E5 5.47E4 4379E2 2.84E-4 4.89E6 7.86E3 LIVER 1.27E3 4.87E3 5.1 1EA 2.38E4, 3.70E4 2.07E3 1.51E4.L.34E5 4.58E4 1.03E4, 1.69E2 4.91 E4 2.05E4 1.13E6 8.48E5 6.70E]236E2 1.90E4 2.02E6 8.56E3 T. BODY 1.27E3 4.87E3 1.35E2 8.40E3 5.54E3 1.43E4 THYROID 1.27E3 4.87E3 7.50E1 KIDNEY 1.27E3 4.87E3 3.07EI i1 .27FA 2178E3 i.98E4 6.24E4 1.25E4 6.68E6 3.155E4 5;66E3 3.22E1 2.64E4 6.22E3 5.49E5 3.11 E5 3.52E3 6.26E1 2.17E3 2.62E5 5.13 E2 1.46137 2.92E6 8.64E4 6.74FP4 1.00E4 4..l E2.8.4014 3-59E4 3.75E5 2:28E1 8.88E3 1.21E6:5.0203 LUNG 1.27E3 4.87E3 2.10E4 1.98E6 1.24E5 1.34E6 8.72E6 1.24E6 2.42E6 1.65E7 2.169E61 751 EE5 1-54E5:1.,46E5 1.21 E5 2035E6 2714E5 6114E5 1.134E7 3.72E5 Gb-LLJ 1.27E3 4.87E3 3.00E3 ,6.68E4 6.3913 I -7E5 9.52E4 2.59E5 4.66E4 3.71 E5 7..65E5 1491E5 9.68E4 2,690E5 6.49b.3 1.031E4 9.76E3 8.48E3 2.29E5 4:87E5.261E5 8.64E5 1.82E5 Unit I ODCM Revision 28 September 2006 If 50 TABLE 3-7 DOSE ANDDOSE RATE R, VALUES -INHALATION -ADULT mrem/vr A.Ci/m 3 NUCLIDE nH3 C 14 Cr.51 Mn 54 Fe 55 Fe 59 Co 58, Co 60 Zn 65 Sr 89 Sr 90:Zr 95 Nb 95 Mc 99 11.31 1,1.33 Cs 134 Cs 137 Ba 140 La 140 Ce 141 Ce 144 Nd 147 BONE 1.82E4 2A.6E4: 1.18.E4 3.24E4 m.485 9.92E7 11.07E5 1.41 E4..2.52E4.9.64E3 3 33E5 4.78E5 3 .90E4 3.-44E2 1.99E4 ,3.43E6 5.27E3 LIVER 1.26E3 3.41E3 3,96E4 1.70E4 2.78E4 1..58E3 1,15E4: 1,03E5 3.44E4 7.82E3 1.21E2 3.58EE4 1.48E4 8.48E5 6.2 1ES 4.90E1 i,74E2 1.35E4 6. 3E6 6,t0E3.T BODY 1.26E3 3.4110 11.00E2, U3bm 3.94E3 1 .06E4"2.07E3 1.48E4.4.66E4 8.72E3.'6.10EF6 2.33EF4 4.21IE3 230EI 2.05 E4 4.51E3 7.28E5 4.28E5 2.57E3 4.SSEI 1.53E3 1,84E5 3.65LE2 THYROID 1.26E3 3.4 IE3 5.95E1 KIDNEY 1.26E3 3.4 1E3 2.28EI 9.84E3 1. 1 9E7 2.15E6 6.90E4 5,47T-4 7.74E3 2.91 E2 6.13E4 2.58E4 2.87E5 2.22F5.1.67E1 6.26E3 8.148E5 3.56E3 LUNG I..26E3 3.41 E3 1.44E4 1.40E6 7.2 1E4ý1 02E6 9.28E5 5.97E6 8.64E5 1.40E6 9.60E6 1.77F6 5:05E5'9.12E4 9.76E4 7.5274 1.27E6 1.36E5 3.62E5 7.78E6 2.2 1E5 GI-LLI 1 .260,.ý3.41 E3 3.32E3 7.74E4'6403E3.1 MP.5 1.06E5ý2.85E5 5.34E4 31.50E5 7.,22E35 1.50E5 1I.04E5 2-48E5 6.2 8E3 8.88E3 1.04E4 18 .40E3 2.118E5 4.58E5 1.20E5 8.16E5.1.73 £5 Unit 1 ODCM Revision 28 September 2006 11 51 TABLE 348 DOSE AND DOSE RATE R,.VALUES -GROUND PLANE ALL AGE GROUPS m -mrem/yr pCi/sec NUCLIDE TOTAL BODY SKIN H 3 C. 14, Cr 51 Mn 54 Fe 55 Fe 59 Co 58 Co 60 iZn 65 Sr 89 Sr 90 Zr 95 Nb 95*Mo 99 1131 1133 Cs. 134 Cs 137 Ba 140 La 1,40 Ce 141 Ce 144 Nd 147 4ý65E6 1 .40E9 2,73E8 3,80E8'2.15EI0 7.46,E8.2. 1 6E4, 245E8 ,1.36E8, 3,99E6 1.7M!7 2.39E6 6.8 3E,9.LO03EI0 2.105E7 1.92E7 1.317E7 6,96E7 8.46E6 5.50E6 1.64E9, 3 .20E8 4.45E8 2.53E10 8.57E8ý2.5 t 4 2.85E8 1.6 E8 4.63E6 2,09E7*2.91 E6: 7.97E9 1.20El0 2.315E7 2. 1,8E7 1.54E7 8.07E7 1.01 E7 Unit I ODCM Revision 28 September 2006 I1 52 TABLE 3-9 DOSE AND DOSE RATE R. VALUES -COW MILK -INFANT m 2-mrem/yr AtCi/sec NUCL IDE.H 3".C 14'Cr51 Mn 54.Fe 55 Fe 59 Co,58 Co 60 Zn 65 Sr 89-Sr 90 Zr 905 Nb 95 Mc 99 1131 1133 Cs 134 Cs 137 Ba 140 La 140 Ce 141 Ce 144: Nd 147 BO)NE 3.23E6:8.43E7 1.22F8 3.53E9 6.93E9'8.19E]0 4.21 E5 6.81E8.8.52E6 2.41E10 3;47610 1.2 1E8 2.03EI 2.2864 1.49E6 4.43E2 LIVER 2.38E3 6.89E5 2 51 E7 5.44E7 2.,1 3E8 1.39E7 5.90E.7 1.21EI0 9.39E2 1.64E5!1 .04E8 8.02E8 1.24E7 4.49(T10 4.06E10.1.21E5 7.99 1.3964 6,10E5 4.55E2 T BODY 2.38E3 6.89E5 8,35E4 5.68E6 1.45E7 8.3 6E7.1.39E8 5.58E9 1.99E8 2.09E 10 6.66E2 1.1 7E5 2.03E7 3.53E8 3163E6 4.54 E9 2.88E9:6.22E66 2.06 1.64E3 8.34E4 2.79EI THYROID 2.38E3 6:89E5 5.45E4 KIDNEY 2.38E3.6.89E5 1. 1964 5.56E6 LUNG 2.38E3 6.89E5 1.06E5 2.66E7 6.29E7 2.64E1 I 2.26E9 5.87E9 1,01E3 154E5 1.55E8 ,9,37E8 1.46E7 1.16E10 1.09EI0 2-87E4 4.28E3 2.46E5 1.76E2 GI-LLI 2.38E3 6.89E5 12:,43E6 9.21E6 ,6.91 E6 1.02E8 3.46E7 1.40E8 11021 10 ,1.42E8 1.02E9 4.68E5 3.03E8 3143E7 2.86E7 2.1 0E6 1.22E8 1.27E8 2:97E7 9.339E4 7.18E6 8.54E7 2.89E5 4.74E9 4.41 E9 7A42E14 rmrem/yr per [tCi imr 3.Unit 1 ODCM Revision 28 September 2006 1153 TABLE 3-10'DOSE AND DOSE RATE R 1 VALUES -COW MILK -CHILD mn 2-mrem/vr gCi/see T. BODY THYROID KI NUCLIDE BONE LIVER DNEY LUNG GI-LLI H3" C 14'Cr 51 Mn 54 Fe 55 Fe 59.Co 58 Co 60 Zn. 65;Sr 89 Sr 90 Zr 95 Nb 95 Mo 99 1 .131 1133 ,Cs 134 ,Cs 137 Ba 140'La 140 Ce 141 Ce 144 Nd 147 1.65E6 6.97E7 6.52E7 2..63E9 3.64E9 7..53E1 2-17E3 1..86E5 3.26E8 4-04E6 I .50E10 2. 17EI0 5.87E.7 9.70 i.1.15E4 1.04E6 2.24E2 1.57E3 3.29E5 1.35E7 3.0.7E7 1.06E8 6.94E6 2.89E7 7.00E9 4.7E2 1.03E4 4.07E7 3.28E8 4.99E6 2.45E1 0 2.08E10 5. 14E4 3.39, 5.73E3 3.26E5 1.81E2 1.57E3 3.29E5 5.27E4 3.59E6 I.15E7 5.26E7 2.13E7 8.52E7 4.35E9 t .04E8, 1.91E10 425E2 5.69E4 1,01E7 1.86E8 1.89E6 5.1 8E9 3.07E9 3.43E6.1.14.8,5 1E2 5.55E4 1.40E]1.57E3 3.29E5 1093E4 1,57E3 3.78F-6 1.57E3 3.29E5.5.34E4 2.0.9E7 3.*.06E7 9.27E8 4.4 ! E9.;6.83 E2 1.00E5 8.69E7 5.39E8 8.32E6 7.61E9 6.78E9 1.67E4 2.51 E3[.ROE5-9.94E1 I.57E3 3.29E5 2.80E6 L1 13E7 6-85E6 1.1 0E8 4.05E7 1.60E8 1.23E9 1.4iEg..01 E9 4.98E5 4.42E8 3.37E7 2.92E7 2. 01 F6 1.32E8 1L3OES 2.97E7 9.45E4 7. 15E6 8949E7 2.87E5 2.73E9.2.44E9 3.07E4 mrem/yr per pCi/m,.Unit I ODCM Revision 28 Septenm bet 2006 1154 TABLE 3-Ij DOSE AND DOSE RATEVALUES -COW MILK -TEEN m-mrem/vr ACi/sec NUCLIDE H3 C14 Cr 51 Mn 54 Fe 55 Fe :59 Co 58 Co60 Zn 65 Sr 89, Sr 90 Zr 95 Nb 95 Mo 99 1131 1133 Cs 134 Cs 137 Ba 140 La 140 Ce 141 Ce 144, Nd 147 BONE 6.70E5 2.78E7ý2.81E7 1,34E9 1.476,9 4.45E10 9.34E2 1.86E5 1.34E8 1.66E6 6.49E9 9.02E9 2,43E7 4.05 4.67E3 4.22E5 9.12EI LIVER 9.94E2 1.34E5 9.01E6 I1.97E7 6.57E7 4.55E6 1:86E7 4.65E9 2.95E2 1.03 E5 2.24E7 1.88E8 2.82E6 1.53E10 1.20E]0.2.98E4 1..99 3.12E3 1.74E5 9.9 I1I T BODY 9.94E2 1.34E5 2.58E4 1.79E6 4.596 2.54E7 1.05E7 4.19E7 2.17E9 4.211E7 1. !0E) 0 5.69E4 4.27E6 1.01E8 8.59E5 7.08E9 4.1t8E9 1.57E6 5.30E-1 3,58E2 2.27E4 5.94E0 THYROID 1.34E5 1.44E4 KIDNEY 9.94E2 1.34E5 5.66E3 2.69E6 LUNG:9.94E32 1.35E5 3.69E4 1.25E7 2.07F7 5.49E10 3.93E8 2.97E9 41.33E2 1.003E5 5.112E7 3-24E8 4.94E6 4.85E9 4.0869 1.0) E14 I1.47E3 5.04E5 5.82E1 GI-.LLI 9.94E2 1.34E5 4.34E6 1,85E7 8.52E6 1.55E8 6127E7 1.97E9 1.75E8 1.25E9 6.80E5 4.42E8 4.01E7 3.72E7 2.13E6 1.90E8 1.71E8 3.75E7 1.14E5 8.91E6 1.06E8 3.58E5 1.85E9 1.59E9.2.00E4.mremlyr per ptCi'rn 3.Unit 1 ODCM Revision 28 September 2006 1155 TABLE 3-12 DOSE AND DOSE RATE Ri VALUES -COW MILK. -ADULT Im 2 mrem/yr gCi/sec NJUCLIDE ,H 3.*C 14" Cr 51:Mn 54 Fe 55 F.e 59 Co 58'Co 60 Zn 65 Sr 89 ,Sr 90 Zr 95 Nb95 NM4o 99 1131 I 133 Cs 134 Cq 137 Ba 140 La,140 Ce. 141 Ce 144 Nd 147 ROWE 3.63E5 1.5 7E7 1,61F.7 8.7 1E8 7.99E8 3.15EI0 5-.34E2'7A IE7 9-019E5 1~74E9 4,97E9.1.35E7 2-26.,2.54F,3 2.29E5 4.74E]LIVER 7.63E2 7.26E4 5.4 1E6/,08E7 3.79E7 2.70E6 I.10E7 2.77E9 1.71E2 6.07E4 1.24E7 1,06E8 1 .58E6 8.89E9 6.80E9 1.69E4 1:14 1,72E3 9.58E4 5 .48El T BODY 7.63E2 7.26E4 1.48E4 1 .03E6 2:52E6 1!.45E7 6.05E6 2.42E7 1.25E9 2.29E7 7.74E9 1. 16E2 3.27E4 2.36E6 6.08E7 4.82E5 7.27E9 4.46E9 8.83E5 3.01E-I 1.95E2 1.23E4 3.2.8E0 THYROID 7.63E2 7.26E4 8.85E3 KIDNEY 7.63E2 7.26E4 3.26E3 1. 61 E6 L UNG:7.63E.2 7.26E4 1.96E4 6.Q4E6.I .06E7 3.47EI0 2.32E8 1.85E9 1269E2 6.00E4 2.81 E7 1. 82E8 2.76E6 2.88E9 2.31E9 5.75E3 7.99E2 5.68E4 3.20EI GI-LLI 7.6M,2 7.26E4'3.72E6 1,66E7 6-211E6 1.26E8 5.47E7 2.06E8 1.75E9 1..28E8 9.11 E8 5-43E5 3-69E8 2 .87E7, 1.80E71 1.42E6 1.32E8 2.177E7 8~35 E4 6.58E6 7.74E7 2.63E5 9.55ES 7168ES 9 .69E,3 mremlyr per uCilm 3.Unit I ODCM Revision 28 September 2006 fI 56 TABLE 3-13 DOSE AND DOSE RATE RP VALUES -GOAT MILK -INFANT nim 2 mremlyr jiCi/sec NUCLIDE H 3 C I4 Cr 51 Mn 54 Fe 55 Fe 59 Co 58 Co 60 Zn 65 Sr 89 Sr 90 Zr 95:Nb 95 Mo 99.1131 1133 Cs 134 Cs 137 Ba 140 La 140 Ce 141 Ce 144 Nd 147 BONE 3.23E6 1.08E6 1-59E6 4.24E8 1.48E10 1.72E1l 9.42E4 8.17E8 1.02E7 7.23E10 1' 04E1 I 1 45E7 2A3E0, 2.74E3 5 .79E5 5.32E1 LIVER 633E3 6.89E5 3.01E6 708E5 2.78E6 1.67E6 7.08E6 ,1.45E9 1.13C2 388FA 1.27E7 9.63E88 1..49E7:,35E1 I 1.22E 11 1.45E4 9.59E-I 1;67E3 732E4.5,47E]T. BODY 6.89E5 1.00E4 6.82E5 1.89E5, i .09E6 4.16P6 1.67E7 6.70E8 4.24E8 4:38E10 8'.04EV 2.24F4 2.47E6 4.23E8 4.36E6 1.36E10 8.63E9 7.48F5 21478-I 1.96E2 3.3 580O THYROID 6.33E3 6.89E5 6.56E3 KIDNEY 6-33E3 6.89E5, 1.43E3 6.67E5 7.04E8 3.1,6E 1I 2.71 E9 1.22E2 1.89E7 1.12E9 1.75E7 3.47E10 3.27EIO 3.44E3.5.141E2 2.96E4 2A IlEl LUNG GJ-LL1 6.33E3 6.33E3 6.89E5. 6.89E5 1 .28E4 2.93E5 1. 11 E6 3.46E5 8.98E4 8.2 ]E5 1,33E6 4. 1 6E6-- 1.68E7-- 1.23E9.-- 3.'04E8-- .2.15E9--5.65E4-- 3.27E7-- 4.17E6-- 3.44E7-- 2.52E6 1.42E 10 3..66E8 4.32E10 3.81E8 8.91E3 3.56E6-- 1.13E4 8..62E5-- 1;03E7-, 3.46E4 mremfyr per ,Ci/m 3.Unit 1 ODCM Revision 28 September 2006 1I 57 TABLE 3-14 DOSE AND DOSE RATE Ri, VALUES -GOAT MILK -CHILD m 2-mrem/vr liCi/sec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H ' -P A 1'7T%" Al7rl A I7L-2 A loVl A 17U C 14*Cr 51 Mn 54 Fe 55 Fe 59 Co 58 Co.60 Zn 65 Sr.89 Sr 90 Zr 95 Nb 95 Mo 99 1 131 1 133 Cs 134 Cs 137 Ba 140 La 140 CeIH41 Ce 144 Nd 147 1.65E6 9.06E5 8-52E5 3.'15E8 7.771E9 1.581E'1 2.621E2 5.05E4 3.91E8 4.84E6 4.49E10 6.521E10 7.05E56 1.16E60 1.3$E3 S.253E5.2.68E11 3.29E5 1.62E6 4.81E5 1 38E6 8.35E5 3.47E6 8.40E8 5.7611 1.96E4 4.95E6 3.94E8 5.99E6 7.37E10 6.24E10 6.18E3 4.07E-1 6.88M32 3.91 E4 2.17E1I 3.29E5 6.34E3 4.3 1 E5 IA9E5 6.86135 2.56E6 1.02E7 5.23E38 2.22E8 4.01E10 5.13E1 1.40E4 1.22E6 2.24E8 2.27E6 1.55E10 9.211E9 4.12E5 1.37E-1 1.02E12 6.66E3'1.681E0 3.29E5 3.52E3 3.29E5 9.62E2 4.54E5 3.29E5 6.43E3 2.72135 3..99E5 1.30Ell 1.111E9 5.29E8 8.25EI1 1.,85E4 1,06E7 6.46E8 9.98E6 2.28E10 2.03E10 2.01 E3 3.02E2 2.16E4 1.19131 3.29E5 3.36E5 1.36E6 8.91E4 1.43E6 4.871E6 1.92E7 1.48E8 3.01E8 2.13E9 3.63E7 4.09E6 3,501E7 2.41E36 3.97E8 3.91138 3,.57E6 1.1!3E34 8.59E5 1.02E7 3,44E4 8.19E9 7.32E9 3.68E3.mremn/yr per 3.Unit ' ODCM Revision 28 September 2006[1.58 TABLE 3-15:DOSE AND DOSE RATE R, VALUES -GOAT MILK -TEEN m 2-mrem/vr RCi/sec NUCLIDE BONE LIVER H 3" CG.14*Cr 51 Mn 54 Fe 55 Co 58 Co 60 Zn 65 Sr 89 Sr90 Zr 95 Nh 95 Mo 99 1131 1,133 Cs 134 Cs 137 Ba 140 La 140 Ce 141 Ce 144 Nd 147 6.70E5 3.61E5 3.67E5 1.61 E8 3.14E9 9.36E10 1.13E2 2.23E:4 1.61 E8 1.99E6 1.95E:10 2.71E10 2.92E6 4.86E-I 5.60E2 5:06E4 1.09EI 2.64E3 L .34E5 1.08E6 2.56E5.8.57E5 5.46E5 2.23IE6 5.58E8 3.56E1 1.24E4 2.7216 2.26E:8 3.,38E6 4.58EP1 3.60E10 3.58E3 2.39E-1 3.74E2 2.09E4 1.19EL, T. BODY 2.64E3 1.34E5.3.1 iE3 5.9,7E:4 13 1 E5 1.26E6 5.03E6 2.60E8 8.99E7 2.31F10(2.45E:1 6.82E3 5.19E:5 1L21 E8 1.03E6 2.133EI.o 1.25E] 0 1.88E5 6,36E-2 4.3.0E]2.72E3 7.13E-1 2.,64E3 1.34E5 1.73E3 THYROID KIDNEY LUNG 2.64B3 1.34E5 6-82E2 3.23E5 2.64E3 1.35E5 4.44E3!2.62E5:2.70E5 6.59E10 4.72E8 3.57E8 5.23EI 1.20E4 6.23E6 3.89E8 5.93E6 1.46E10 1.23E10 1.211E3 1.76E2 1.25E4 6,99E0 GI-LLI 2.64E3 1.34E5 5.23E5 2.22E6 1.11 E5:7.53E6.2:91 E7 2.36E8 3.74E8 2.63E9 8.22E4 5.30E7 4.87E6 4.47E7 2.56E6 5,70E8 5. 12E8 4.50E6 1.37E4 1.071E6 1.27E7 4.29E4 5,56E9 4.76E9 2.411E3 e / p .3 mrem/yr per pCiim.Unit I ODCM Revision 28 September 2006 A 59 TABLE 3-16 DOSE AND DOSE RATE Ri VALUES -GOAT MILK -,ADULT m -mrem/yr tCi/sec NUCLJDE"143 C 14 Cr 51 Mn 54 Fe 55 Fe 59 Co 58 Co 60 Zn,65 Sr 89 Sr 90 Zr 9s Nb 95 Mo 99 1131 113-3 Cs 134 Cs. 137 Ba 140 La 140 Ce 141 Ce 144 Nd 147: BONE 3.63E5 2104E5 2.10E5 L05E8 1.70E9 6.62E1I0 6.45E1 1331 F4 8.89E7 1.09E6 1.12E110 1I.49E 10 1 62E6 2.71 E-!1.3.06E2 2.75E4 5..69E0 LIVER 2.031=3 7.26E4 6.50E5 1.41 E5 4.951E5 3.25F5 1.32E6 3.33E8 2.o7E11 7-2913 1.511E6 1.27E38 1.90E6 2.67E10 2-03,3 1.36E-1.071E2 IA5E4 6.57E0 T BODY 2.0303 7.26E.4 1.78E3 1.24E5 3.28E4 JL90E5 7T27E5 2.911E6 1.51 E8 4.89E7 1.63E10 1.40EI 3.92E3 2.871E5 7.29E7 5.79E5 2.18 E10 1.34E 10 1!.06E35 3.6 1E-2 2.34E13 1.48E3 3.9313-1 THYROID 2;03E3 7.26E4 1.06E3 KIDNEY 2.03E3 7.26E4 3.92E2 L93E5 LUNG-2.03E3 7.26F4, 2.36E3 7.85E4 1,38135 4.17E10.2.79E8 2.23E8 3-25E1 7.2103 3A113E6 2.1 SE8 3.3.1E6 8.631ý9 6.93E9 6.911E2 9.60EI 6.82E3 3.84E0 GI-LLI:2.03E3 7.26E4 4.48E5 1.99E6 8.07E4 1.65E-6 6.58E6 2.48E7 2. 10E8 2.73E8 1.9 1E9 6,56E4 4.421E7 3.49E6 3.36E7 1.711E6 4.67IE8 3.95E8 3.'33E6 1.00E4"7.90E5 9.30E6 3.15E4 2.86139 2.30E9 1.166E3.mrem/yr per pCi/rmn.Unit I ODCM Revision 28 September 2006 I'l 60 TABLE 3-17 DOSE AND DOSE RATE RI. VALUES -'COW MEAT -CHILD IiCi/sec ,NUCLIDE BONE LIVER H 3*C 14" Cr51 Mn 54 Fe 55 Fr 59 Co 58 Co 60 Zn. 65 Sr 89 Sr 90 Zr 95 Nb 95 MO 99 1.131 1133 Cs 134 Cs 137 Ba 140 La -140 Ce 141 Ce 144 Nd 147 5.29E5 2,89E8 2.04E8 2.38E8 2.65E8 7.01E9 1.5 1E6 4.10E6 4.15E6 9.38E-2 6.09E8 8.99E8 220E7:2;80E-2 1.1764 1.48E6 5.93E3 2.34E2 1.06E5 5.15E6 i .53E8, 3.30E8 9.41 E6 4.64E7 6.35E8 3,3265:!..59E6 5.42E4 4.18E6 1.166E-1I, 1 00E9 8.60E8 1I93E4 9.78E-3: 5.8203 4.65E5 4.80E3 T BODY 2.34E2 1.06E5 4.55E3 1.37E6 4.74E7 1.65E8 2.88E7 1.37E8 3.95E8 77.57E6 1.78E9 2.95E5.I 14E6 1.34E4 2.37E6 439E72.2.1 1E8 1.27E8 1.28E6 3.30E-3 8.64E2 7.91E4 3.72E2 TABLE 3.17:DOSE AND DOSERATE* Rt. VALUES -COWMEAT., cHILD m2-rorem/vr IIIIH .......... THYROID 2.34E2 1.06E5 2.52E3 1.38E9 2.15E)KIDNEY 2.34E2 1 06E5 6.9ME2 1.44E6 LUNG 2.34E2 1I06E5 4.61E3.866E7 9.58E7 4.00E9 GI-LLI.2.34E,21.1.06E5 4.32E6 2.84E7 3.4-E8:5.49E7 2.57E8 I. 12E8 1.03E7 9.44E7 3.46E8 2.95E9 4.48E4 3.72E5 4.67E-2 5.39E6 5.39E6 Ll11E7 21.173E2 7 .26E6 1.216E8 7,6186 4.75E5 1.50E6 L 16E5 6.86E6 1.93E-1 3.10E8 2.80E3 S6.27E3 2.55E3.2.57E5 2.64E3 1.1168 1.015E.1.15E4"mrem/yr per pCi/ I 3.Unit I ODCM Revision 28 September 2006 1.1 61 TABLE 3-18 DOSE AND DOSE RATE R 4 VALUES -COW MEAT -TEEN m 2-mrem/vr p.Ci/sec NUCLIDE.BONE LIVER H3" C.14" Cr 54, Mn 54 Fe 55 F7 59 Co 58 Co60 Zn 65 Sr 89.Sr 90 Zr 95 Nb 95.Mo 99 1 131 I 133 Cs 134 Cs 137: Ba 140.La 140 Ce 141 Ce 144 Nd 147 2.81E5S]1,50E8 1.15E8:1.59E8 1.40E8 5.42E9 8.50E5 2.37E6ý2.24E6 5.05E-2 3.46E8 4.88E8 1.19E7 1.53E-2 6.19E3 7.87E5 3.16E3 1.94E2 5.62E4 4.50E6 1.07ES 2.69E8 8.05E6 3.90P7 5.52E8, 2.68E5 1!:32E6 390E4 3.13E6 8.57E-2 813E8 6.49E8 1 ;46E4, 7.51E&3 4.14M3 ,3.26E5 3.44E3 T BODY 1.94E2 5.62E4 2.93E3 8.93E5 249E7 ,;86E7 8, 80E27 2.57E8 4.0 1E6 1.34E9 1.94E5 7.24E5 7.43E3 ,1.68E6:2.61E-2 3.77E8 2,26E8 7;68E5 2,00E-3 4.75E2 4.23E4 2.06E2 THYROID 1.94E2 5.62E4 1.62E3 KIDNEY 1.94E2 5.62E4 6.39E2 1.34E6 LUNG 1.94E2 5.62E4 4.16E3.6.77S7 8.47E7 9.15E8 1.20E1.1353E8 3.94E5 1.28E6 8.92E4 5.40E6 1.50E-1 2.58E8 2.21E8 4.95E3 1.95M3 1.94E5 2.02E3 GI-LLi I i94E2 5.62E4 4,90E5 9.24E6 4.62E7 6.36E8 1.1IE8 5.09E8 2.34E8 1.6707 1.52E8 6.19E8 5.63E9 6.98E4 6.20E5 6.48E-2 1I,01E7 9.24E6 1.84E7 4.3 1E2 1.18E7 1.98E8 1.24E7 9.8727 8.58E27 9.81E3'mrem/ryr per pCi/m 3.Unit 1 ODCM Revision 28 September 2006 1162 TABLE 3-19 DOSE ANDDOSE RATE Ri VALUES -COW MEAT -ADULT: m 2-mremlyr liCi/see JNUCLIDE BONE LIVER T BODY THYROID KIDNEY LUNG GI-LLI H3" C 14'Cr 51 Mn 54 Fe.55 Fe 59 Ca 58 Co 60 Zn 65 Sr 89 Sr 90 Zr 95ýNh 95 Mo 99 113 1 1I1133 Cs 134 Cs 137 Eta140 La 1,40 Ce 141 Ce: 144 Nd 147 3.33E5 1 .44E,8 2.26IE8 I1.66E8 8.38E9.1.06E-6 3.04E6 ,2.69E6*6 -04E-2'4.35E8 5,88E8 1.44E7 7.3.883 9,33E5 3.59E3 3.25E2 6.66E4 5.90E6 L28E8 3.39E8.1.04E7 5.03E.7 7.19E8 314065-1.69E6 4.71 E4 3. 85E6 1.05E-1 1.03E0 8.04E8 1.81 E4 9.37E-3 4499E3 3.90E5 4.15E3 3.25E2 6.66E4 3.65E3 1.13E6 2,98E7 1.30E8 2.34E7 1.1 IE8 3.25E8 4.76E6 2.06E9 2.30E5'9.08E5 8.97E3 2.21E6 3.20E-2 8.45E8 5.26E8 9.44E5 2,48E-3 5 .66E2 5.01 E4 2.48E2 3.25E2:6.66E4 2.18E3 3ý.25E2 6.66E4 8,03M2 1.76E6 3.2*5E2 6.66E4 9.46E7.4.8 lEg I.2-'6E9 1,54EI 5:34E5 1,67E6 1.07E5*6.61E6 1.83E-1 3.35E8 2.73ES 6.1 5E3 2.32E3 2.31 E5 2.42E3 3.25E2 6.66E4 9.17E5 1.81E7 7.34E7 1.13E9 2.12E8 9.45E8 4.53E8 2.66E7 2.42E8*1.08E9 1.03E10 1.09E5*1.02E6 9.44E-2 1.81E7 1.56E7 2.97E7 6.BE2 1.I91E7 3.. 16E8 1.99E7 1.1 IE8 9.07E7 L04E4 mrem/yr per PCi/m 3.Unit 1 ODCM Revision 28 September 2006 1163 TABLE 3-20.DOSE AND DOSE RATE R, VALUES -VEGETATION -,CHILD m 2-mrem/yr pCi/sec NUCLIDE BONE LIVER: H 3." C 14" Cr 51 Mn 54 Fe,55 Fe 59: Co 58-Co 60 Zn 65 Sr 89 Sr 90 Zr 95 Nb 95, Mo 99 1131 1133 Cs 134 Cs 137 Ba 140 La.140 ce 1.41 Ce 144 Nd 147 3-50E6 7.63E8'3.97E8 8.12E8 13.59E10 1 .24E112 3.86E36 i.02E6 7.163E7 1.69E6 I .601E0 2.39E 10 2.77E8 3.25E3 6.56E5 1.27E8 7.23E4 4.01E3.7.01E5ý6.65E8 4-05ES8 6,42E8 6.45E7 3.78E8 2.16E9-8:50E5 3.99E.3 7.70E6 7.20E,7 2.09136 2.631E10 2.291310 2.43E5 1.13E3:3.27E5 5.86E4 T BODY 4.0,E3 7.0,11ES 1.17E5 1.77E38 3.25E8 1.97E8 1. 12E9 1.35E39 1.03E9 3.15E13t 7.566E5 2.885 E5 ,1.91E6 ,4.09E7 7.92E5 5.55.E9 3.38E9 1.62E7 3,83E2 4.85E4 6.78E6 4.54,E3 THYROID 4.11E3.7.01 E5 6.49E4 2.381E10 3.89E8 K5ID NE Y 4.0 E3 7010IE5 1.77E34 1.86E8 LUNG 4.0103 7.01E5 1.188E5 2.291E8 L.86E38 1.36E9 GI-LLI 4o0:iE3 7.01 E5ý6.20E6 5.58E8 7.50SE7 6.6918 3.76E8 2.10E9 3.80E8 1 .39E9 1.67E110 8.8613E 7.37E8 6.3 7E6 6.41 E6 8.44E5 1.42E8 1.43E8 L.40138 3.1 6E7 4,08E8 1.04E10*9.28E7 1.22E6 3.75E5 1.65137 1.183E8 3.49E6 8.15139 7.46E9 7.90E4 1.43E5 2.211E7.3.22E4 2.93B9 2.68E9 I A5ES.mrem/yr per PCi/m 3.Unit 1 ODCM Revision 28 September 2006 1I 64 TABLE 3-21 DOSE AND DOSE RATE R, VALUES -VEGETATION -TEEN mz-mremlyr g.Ci/sec NUCLIDE BONE LIVER H 3 C 14" Cr 51 Mn 54 Fe 55 Fe 59 Co,58 Co 60.Zn 65 Sr;89 Sr 90 Zr 95 Nb 95 Mo 99 I 131 1,133 CUs 134 Cs 137 Ba 140 La 140 Ce 141 Ce 144 Nd 147 1.45E6ý3. OES 1.79E8 24.E8, 1.51lE10 7.5 IEII 1 .72E6 4.80E5 1.85E17 9.29E5 7.110E9 I .01E10'1.38E8'14S1E3 2.93E5, 5.27E7 3.66F,4 2.59E3 2.91 E5 4.54E8 2.20E8 4.18S8 4.37E7 2.49E&8 1.47E.9 5.44E.5 2,66E5 5.64E6 5,.39,E7 1.58E6 11.61810 1.35E10 8.8 8E2, S.189ES 2.1 8E7 3 .98E4 T BODY 2.59E3 2.91E5 6R16E4 9;01 E7 5.13E7 1.61E8 1.01ES 5.60E8 6.86E8 4.33E8 1.85E1I/3.74E5 1.46E5 1.08E6 2.89E7 4.80E5, 7.75E9 4.69E9:8.9 1 E6 2.36E2 2. 17E4 2.83E6 2.3863 2.59E3 2.91E5 3,42E4 THYROID KIDNEY LUNG 2.59E3 2.9 1,E5 1.,35E4 1.36E8 2.59E3.2.9 1 E5 9.79E4 1.3 2ES 9.4.1ES GI-LLI 2.59E3 2.9 IE5 1./03E7 9.32E8 9.53E7 9.89E8 6.02E8 3.24E9 6.23E8 1.80E9 2.1 IEIO 1.26E9 1.14E9 1.01E7 1.07E7 1.1 9E6 2.08E8'1.92E8 2*13E8 5.1OE7 5.4OE8 L33ElO 1.44E8 1.57E10 2.20E8 7.99E5 2.58E5 1.29E7 9.28E7 2.76E6 5.31 E9 4.59E9 5.74E4 8.89E4 1.30E7 2.34E4 2-03E9 1.78E9, 1.14E5 mremtyr per 4Ciim 3 Unit I ODCM Revision 28 September 2006 1.1 65 TABLE 3-22 DOSE: AND DOSE RATE RI VALUES.- VEGETATION -ADULT m NUCLIDE H 3 C 14" Cr 51 Mn:54 Fe 55 Fe 59 Co 58 Co 60: Zn 65 Sr.89 Sr 90 Zr 95 Nb 95 Mo 99 ,1131'1133 Cs 134 Cs 137 Ba 140 La 140:Ce 141 Ce .144 Nd 147 BONE 8.97E5 2:00E8 I.26E8 3.17E8 9.96E9 6.05E111 1.1 8E&6 3.55E5 4.04E7 1.00E6 4.67E9 6.36E9 1.29E8 1.98E3 1 .97E5 3"29E7 3.36E4 LIVER 2.26E3 1.79E5 3.13E18 1l.3E813 2.961E&3.08E7 1.67E8 1.01E9 3.77E5 1.98E5 6.14E6 5.78E7 1.11E10 8.70E9 1.61E5'9.97E2 1!.33E5 1.38E7 3.88E4 T. BODY 2.26E3 1 .79E35 4.64E4 5.07E7 3.22E7 1.13E8 6.90E7 3.69E38 4.56E8 2.86E8 2.55E5 1.06E5.1.17E-6 3.31 E7 5.30E5 9.08E9 5:70E9 8.42E6 2.63132:1.51E34 1.771E6 2.3203 2-mrem/vr jtCi/sec THYROID 2.26E3 1.79E5 2.77E4 iý-KIDNEY 2.26E3 1.79E5 1.02E4 9.31 E7 LUNG 2;26E3 1.79E5 6.15E4: 7.69E7 8.27E7 1.90E10 2.56E8 6~.75E8 5.92135 1. 95E5S.1.39E7 9.9117 3.03E6 3.59E9 2.95E9 5.49E4 6.19E4 8.1 6E6 227E4 GI-LLI 2.26E33 11.79E5.'I IA7E7 9.58E8 7.911E7 1.021E9 6.24E8 3.14E9 6.36E8 1-60.E9 1. 751310 1.20E9 1.20E9 1.42E7 1.5307 1.56E6 1.94E8 1.68E8'2.65E8 7.3212E7 5. 09E8 1.86E8 1.19E9: 9:81 ES 9.25E4 mrem/yr per XCi/m 3 Unit 1 ODCM Revision 28 September 2006 1166 TABLE 3-23 PARAMETERS .FOR THE EVALUATIONKOF DOSES TO REAL MEMBERS OF THE:PUBLIC FROM GASEOUS.AND LIQUID EFFLUENTS Pathway Fish Parameters U (kg/yr) -. adult Daipj (1 rei/pC i)Value 21 Reference Reg. Guide 1.109 Table E&5 Reg. Guide 1.109 Table E-1 1 Fish Each Radionuclide Shoreline Shoreline Inhalation U (hr/yr)-adult-teen 67 67 D a lpj P i M (mrem/hr per pCi/m 2)DFAija Each Radionuclide Each Radionuclide Reg. Guide 1.09 Assumed to be same as Adult Reg. Guide 1.109 Table. E-6 Reg. Guide 1.109 Table E-7 Unit I ODCM Revision, 28 September 2006 II 67 .Type of Sample Radioiodine and Particulates (air)Radioiodine and Particulates (air)Radioiodine and Particulates (air)Radioiodine and Particulates (air)Radioiodine and Particulates,(air) Direct Radiation (TLD)Direct Radiation (TLD)Direct Radiation (TLD)Direct Radiation (TLD)Direct Radiation (TLD)Direct Radiation (TLD).Direct Radiation (TLD)Direct Radiation (TLD).Direct Radiation (TLD)Direct Radiation (TLD)Direct Radiation (TLD)Direct Radiation (TLD)Direct Radiation (TLD)TABLE 5.1 NINE MILE POINT NUCLEAR STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGR, SAMPLING !LOCATIONS

• Map Collection Site Location (Env. Program No.)Nine Mile PointRoad North (R- 1)2 Co. Rt. 29 &LakeRoad (R-2)3 4.6 7 8 9 1.0 11 12.13 14 15:16 17 18 Co. Rt. 29 (R-3)Village of LycomingNY (R-4)MontarioPoint Road (R-5)North Shoreline.Area (75)-North Shoreline Area.(76)North Shoreline Area (77)North Shoreline Area (23)JAF East Boundary (78)Rt. 29 (79)Rt. 29 (80)Miner Road (81)Miner Road (82)Lakeview Road (83)Lakeview Road (84)Site Meteorological Tower (7)Energy information Center (18)L"cation 11.811i @ 88'E 1.1 mi,@ 104'ESE 1.5 mi@ 132' SE 1.8 mi @ 143; SE 16.4 mi @ 420 NE 0.1 mi @ 5'N 0:1 mi @ 25* NNE 0.2mi @ 45 NE 0.8..mi @ 760 ENE 1.O mi @ 90_.E 1.1 mi @JI15' ESE 1.4 mi @ 1330 SE 1.6 mi @ 1590 SSE 16 mi @ 18 17.S.1.2 mi @ 200' SSW 11 mi @ 225 SW 0.7 mi @ 2500 WSW 0..4 mi @ 265'W* Map= See Figures 5.1-1 and 5.1-2 Unit I ODCM Revision 28 September 2006 I168

.Tvpe of Sample.Direct Radiation (TLE Direct Radiation (TLD Direct. Radiation (TLE Direct Radiation (TLD ,Direct Radiation (TLD Direct Radiation (TLD Direct Radiation (TLED Direct Radiation (TLC Direct Radiation (TLD* Direct. Radiation (TLE ,Direct Radiation (TLE Direct Radiation (TLD Direct Radiation (TLD Direct Radiation (Tlr Direct Radiation (TLD Direct -Radiation.(TJL Direct Radiation: (:TLh Direct Radiation (TLD Direct Radiation (TLD Surface Water Surface Water TABLE 5.1 NINE MILE POINT NUCLEAR STATION, RADIOLOG ICAL ENVIRONMENTAL MONITORIN G PROGRAM SAMPLING LOCATIONS* Map Collection Site Location (Fnv. Program Nq.)19 North Shoreline (85) 0.2 r 20 Nurth Shiurcitic,(86) 0A I 21 North Shoreline (87) 0.1 n)). 22 Hickory Grove (88) 4.5 n) 23 Leavitt Road (89) 4.1 n) 24. Rt. 104 (90) 4.2 n) 25 Rt. 51A (911) 4.8.n I) 26 Maiden Lane Road (92) -4,4 .n 27 Co. Rt. 53 (93) 4.4 n) 28 Co. Rt. 1 (94) 4.7 n I) 29 Lake Shoreline.(95) 4.1 n I) 30 Phoenix, NY Control ,(49) 19.8 I). 31 S, W. Oswego, Control (14) 12.6 I) 32 Scriba, NY(96) 3.6 IT 33 Alcan Aluminum, Rt. IA (58) 3.1 rr I) 34 Lycoming, NY (97) 1.8 mr 35 New Haven, NY (56) 5_3 n'36 W. Boundary, Bible Camp (15) 0.9 mr 37 Lake Road (98) 1.2 mn 38 0SS Inlet Canal (NA) 7.6 in 39 JAFNPP Inlet Canal (NA) 0.5*nm I I .........Locatiou ni,@'294' WNW hii@ 315 W ni @ 341I NNW i @ 97' E rIi@ l1 CESEt ni @ 135. SE hi@ 156' SSE hi-@ t 83ý S i"@. 205' SSW ni @ 223- SW ni @ 237° WSW mi @ 163' S mi @2260 SW ii @ 199. SSW ii @ 220* SW ai g !1433' SE ii @ 123" ESE ii @ 237' WSW ki@ 101°E ii @ 235' SW ii @ 70' ENE (NA) = Not applicable

• Map = See Figures 5.-1 and 5.1-2..Unit 1 0DCM Revision 28 September 2006 1169 Type of Sample Shoreline Sediment Fish Fish Fish Milk Milk Milk.Milk (CR)Food Product Food Product Food Product Food Product Food Product TABLE 5.1 (Cont'd)NINE MILE POINT NUCLEAR STATION RADIOLOGICAL.

ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS* Map Collection Site Location (Env. Program No.)40 Sunset Bay Shoreline (NA) 1.5 m 4.1 NMP Site Discharge Area (NA) 0.3 m (and/42 NMP Site Discharge Area (NA) 0.6 m 4'3 0swego Harbor Area (NA) 6.2 m 76 Milk-Location

1. ,76 6.3 m 64 Milk Location #55 9.0 m 66 MilkLocation H/4 7-8 m 77 Milk Location .13.9 (Summerville) 48 Produce Location #6*

• 1.9 m (Bergenstock) (NA)49 Produce Location # l* 1.7 m (Cu leton) (NA)5.0 Produce Location #2*
• i9 m (Vitulo) (NA).51 Produce Location 95*4S 1.5 m (C.S. Parkhurst} (NA), 52 Produce Location #3 *
• 1.6 mi ,(C. Narewski) (NA)Location i g E.J.@ 315 NW or)i@55.NE i @ 235' SW i. @ 120&ESE@ 95@ E i @ 1) 3* ESE ni 19i, SSW i @ 1410 SE, i @96 E.ig i @1 14 F.SP.(@ 8.40E* Map (NA)CR= See Figures 5.1-1 and 5.1-2-FoodProduct Samples need not necessarily be collected from alllisted locations.

Collected samples will be of the highest calculated site average. D/Q-Not applicable Control Result (location) Unit I ODCM Revision 28 II 70 September 2006 Type Of:Saniple Food Product Food Product (CR)FoodProduct (CR)Food Product Food Product Food Product Food Product Food.Product (CR)Food Product Food Product Food.Product Food Product Food Product TABLE 5.1 (Cont'd)NINE MILE POINT NUCLEAR STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS* Map Collection Site Location (Env. Proeram No.) Location 53 Produce Location ,4, ,* 2.1 mi @ I I 0 ESE (P. Parkhurst) (NA)54 Produce Location #7.*

• 15.0-mi @ 223' SW (Mc Millen) (NA)55 Produce Location #8** 12.6Thi @ 225- SW ,(Denman) (NA)56 Produce Location #9*
• 1.6 mi @ 171V S (O'Connor) (NA)57 Produce Location #10*
• 2.'2 mi @ 123'ESE (C.. Lawton) (NA).58 Produce Location #11 2.0 mi -" 12 ESE (C. R. Parkhurst) (NA)59 Produce Location #12*
• 1.9 mi@ 115' ESE (Barton) (NA)60 Produce Location i , 13*" 15.6 mi @225" SW (Flack) (NA)61 Produce Location #14*
• 19 m i@ 950 E (Koeneke) (NA)62 Produce Location # 15*
• 1.7 mi @ 136"SE (Whaley).(NA) 63 Produce Location # 16* *- 1.2 mi @.207' SSW (Murray) (NA)67 Produce Location #17*.* 1.76 mi @ 97 E (Battles)68 Produce Location ft18** 1.52mi @ 85 E (Kronenbitter)
• Map (NA)CR= See Figures 5.1-1 and 5.1-2-Food Product Samples need not necessarily be collected from all listed locations.

Collected samples will be of the highest calculated site average DIQ.-Not applicable Control Result (location) Unit I ODCM Revision, 28 1i 71 September 2006 FIGURE 5.1-1 Unit 1 ODCM Revision 28 September,2006 Ii 72 SCALE Of MILES FIGURE 5.1-2 LEGEND NINE MILE POINT t o .................. CountyRa,:+. _:. -' .' ." --.-O FF-S ITE M A P T..&Sao" IJRhayid... .. ......... O FF'(County LiUes ....... : ...Ton+ Li .............. ........... .2 City & VIIo i ...... .. ....I A I o l .. ............ ......... ... +.E r I hO N M E N T A IL S A M P L , .. .... ..5 )LOWA ION tmimi 43-28','4 ot 1 glag., Ooslo4,. N.Y.Lund Afen 9WS 5uare miles 49 52 40 67 53 5 cd37 0 N/&56" SMet Jet... . FIGURE 5.1-2a 0 t.7 20.M 79 /- (e)L A K T AN Niagara Mohawk Power Corporation retains ownershipin certain transmission line and swilhyard facilities within the exclusion area boundary. Access and usage are controlled by Nine Mile Point Nuclear Station, LLC by Agreement. SCALE- MILESI 1?TI2TW~ ~ 1 ~1.SITE BOUNDARIES NIAGARA MOHAWK POWER .CORPORATION NINE MILE, POINT-UNIT 1 Unit I ODCM Revision 28 September 2006 I 74 APPENDIX A LIQUID DOSE FACTOR DERIVATION Unit I ODCM Revision 28 September 2006 S1175 AppendixrA Liquid Effluent Dose Factor Derivation, Ajat Aiat (irern/hr per ýtCi/ml) which embodies the, dose conversion factors, pathway transfer factors (e.g., bioaccumulation factors), pathway Usage factors, and dilution factors for the points of pathway origin takes into account the dose from ingestion of fish'and drinking water and the sediment. The total body and organ dose conversion factors for each radionuclide will be used from Table EE- II of Regulatory

Guide 1 .109. To expedite time, the dose is calculated for a maximum -individual instead of each age group. The maximum individual dose factor is a composite of the highest dose factor A& of each nuclide i age group a, and, organ t, hence Aiat. It should be noted that the fish. ingestion.

pathway is the most significant pathway for dose from liquid effluents. The:water consumption pathway isincluded for consistency with. NUREG 0133.The equation for calculating dose contributions given in section 1.3 requires the use of the composite dose factorAit for each nuclide, i. The dose factor equation.for a fresh water site is: Ait -t(we + -A-t 69.3 U sW e-At -At b )1 Aw .-c )+Uf (BF~i(e I Pf)(DFL) iat + , s. _ eI) ps,,1l-e i b)(DFS), Aiat (D)oA Where: Ajat Is the dose factor for nuclide i, age group a, total body or organ t, for all appropriate pathways, (mrem/lir per p.Ci ml).Ko. Is the Unit conversion factor, 1.14E5-1E6pCi/.tCi x 1E3 ml/kg 8760 hr/yr.U_ = Water consumption (1/yr); from Table E-5 of Reg. Guide I. 109..U= Fish consumption (Kg/yr); from TableE-5 of Reg.. Guide 1. 109.U Sediment Shoreline-Usage (hr/yr); from Table E-5 of Reg. Guide 1.109.(BF). Bioaccumulationfactor for nuclidei,, in fish, (pCi/kg per pCi/1), from Table A- I of Reg. Guide. 1.109.(DFL)at Dose conversion factor for age, nuclide, i, group a, total body or organ t,.(mrem/pCi); from Table E-1 I of Reg, Guide 1.109.(DFS)i Dose conversion factor for nuclidei and total body, from standing on contami nated ground (mem/hr per pCi/n 2 .); from Table E-6 of Reg. Guide 1.109.D, Dilution factor from the nearfield area within one-quarter mile of the release point .to the..potable. water intake for the adult water consumption. This is the Metropolitan Water Board, Onondaga County intake structure located west the City of Oswego;, (unitless). Unit I ODCM Revision 28..11 76 September 2006 Appendix A (Cont'd)1), .Dilution factor from the near field area within one quarter mile of the'lrelease point to the shoreline deposit (taken at.the same point where wetake environmental samples 1.5 miles;unitless). 69.3 'conversion factor :693 x 100,, 100 = K,. (L/kg-hr)

• 40*24 hr/day/.693 in L/M 2-d, :and K, = transfer coefficient from water to sediment in Likg per hour.tw,tpf, Average transit time requiredfor each nuclide to reach the point of expoisure for internal dose, it is the total time elapsed from release of the nuclidesto either ingestion for water (w) and fish (f)or shoreline deposit (s), (hr).tb = Length of time the sediment is exposed to the contaminated water, nominally 15 yrs (approximate midpoint offacility operating life), (hrs).-. decay~constant for nuclidei (hr-).W Shore width factor (unitless)

from Table A-2 of Reg. Guide 1.109.Example Calculation For 1-131 Thyroid Dose Factor for an Adult from a Radwaste liquid effluents release: (DFS)i = 2.80E-9 mrem/hr per pCi/mn 2 (DFL)j., = 1.95E-3 mrem/pCi tp. = 30 hrs. (w = water)BFi = 15 pCi/Kg per pCi/L = 24 hrs, (f = fish)Uf = 21 Kg/yr tb 1.314E5 hrs. (5,48E3 days)D, = 40 unitless U, 730 L/yr D, = 12 unitless Ko = 1.14E5 (.pi/j.CCi)(ml/kr.)

U, = 12 hr/yr (hr/yr)W = 0.3 3.i 3.61 E-3hr"'tp, = 5.5 hrs .(s = Shoreline Sediment)These values will yield.an Ai,, Factor of 6.79E4 mrem-ml per lCi-hr as listed in Table 2-4. It should be noted that only a limited number of nuclides aie listed on Tablcs 2-l to .2-8. These are the most common nuclides encountered in effluents. Ifra nuclide is detected for which afactor is not listed, then it will be calculated and included in a revision to the ODCM.In addition, not all dose factors are used for the dose calculations. A maximum individual is used; which is:a composite ofthe maximum dose-factor of each age group for each organ as reflected in the applicable chemistry procedures. Unit 1 ODCM Revision 28 11 7.7 September 2006 APPENDIX B PLUME SHINE DOSE FACTOR DERIVA TION Unit I ODCM Revision 28 September 2006 II 78 APPENDIX B For elevated releases the plume shine dose factors for gamma air (Bi) and whole body (Vi), are calculated usingythe finite plume model with an elevation above ground equal to the stack height. To calculate the plume shine factor for gamma whole body doses, the gammai air dose factor:is adjusted for the attenuation of tissue, and the ratio of mass absorption coefficients between tissue and air. The equations are as tbllows: GammaAir B) 1 j K.E I., WIhere: K -conversion factor. (see RG V, below for actual value).= mass absorption coefficient (cm 2/g; air for Bi, tissue:for Vi)E Energy of gamma ray per disintegration (Mev)V$ average wind. speed for each stability .class R = downwind distance (site boundary, m)0 = sector width (radians)s subscript for stability class I,, = I function = 1,+ k12 for each stability class.(unitless,see .Regulatory Guide 1.109)k 2 Fraction of the attenuated energy that is actually absorbed in air. (see Regulatory Guide 1.109, see below for equation)WhQle Body-Vi = I.ISFB;ic Where: td tissue depth (g/cm')SF = shielding factor from structures (unitless) 1.11 Ratio ofmass absorption coefficients between tissue and:air.Where all other parameters are defined above.'K = conversion factor = [3.7 E1O dis,] 1.6 E-6 erg]Sic Mev =0.46 L1293 ]q 123g L100 erq g-radpa Where: jA mass attenuation coefficient (cm 2/g; air for B 3 , tissue for Vi)= defined'above Unit I ODCM Rcvision 28 I 79 September 2006 APPENDIX B (Cont'd)There are seven stability classes, A thru F. The percentage.oftheyear that each stability class occurs is taken from the U-2 FSAR. From this data, a plume: shine dose factor is calculated for each stability class and each nuclide, multiplied by its respective fraction and then summed.The wind speeds corresponding to each stability class are, also, taken from the;U-2 FSARI To confirm the accuracy of these values, an average of the 12 month wind speeds for 1985, 1986, 1987 and 1988 was compared to-the average of the FSAR values. The average wind speed of the actual data is equal to 6.78 m/s, which compared favorably to the FSAR average wind speed equal to 6.77 m/s.The average gamma energies were calculated using a weighted average of all gamma energies emitted from the nuclide. These energies were taken from the handbook, "Radioactive Decay Data Tables", David C. Kocher.The mass absorption (p,) and attenuation (pt) coefficients were calculated by multiplying the mass absorption (4/p.) and mass attenuation (pip) coefficients given in the Radiation Health Handbook by the air density equal to 1.293 E-3 g/co or the tissue density of I g/cc where applicable. The tissue depth is 5g/cm 2 ,for the whole body.The downwind distance is the site boundary.Unit I ODCM Revision 28 September 2006 1180 APPENDIX B(Cont'd)SAMPLE CALCULATION Ex. Kr-89-DATA E=F STABTLTTY CLASS ONLY -Gamma Air 2.22MeV k = = .871 K = 0.46 2.943 E-3m l'a VF = 5.55 m/sec 5.5064E-3m ' R = 644m 0.39*19m .vertical plume spread taken from "Introduction to Nuclear Engineering'!, John R. LaMarsh-I-Function Ue= 0.06 1- 0.33 12 = 0.45 I = I4 -k1 2 = 0.33 + (0.871) (0.45) =0.72 B, = 0.46 C i-sec) (Mev/eras](2.943E-3m-')(222Mev)(.72)(Tw/) (g/m 3) (erQ) (5.55 m/s) (.39) (644m)(g-rad)= 1.55(-6) rad/s (3600 s~hr) (24 h/d) (365 d&v) (1E3mrad/rad' Cl/s (lEuCi)Ci= 2.76(-2) m .radlr'= 1.1(.7) [2.76(-2) mrad/v.pCi/sec 1-(.0253 cm 2/g)(Sg/cm 2)Vi*[e I 1.89(-2) mradfvr j.Ci/sec NOTE: The. above calculation is for the F stability class only. For Table 3-2 and procedure values, a*weighted fraction of eachstability class was used to determine the Bj and Vi values.Unit I ODCM Revision 28 II 81 September 2006 APPENDIX C ORGAN DOSE PARAMETERS FOR IODINE 131 and 133, PARTICULATES AND TRITIUM Unit I ODCM Revision 28 September 2006 1182 APPENDIX C ORGAN DOSE PARAMETERS FOR IODINE- 131 AND -133, PARTICULATES AND TRITIUM This appendix contains the methodology ývhich was used to calculate the organ dosefactors forI- 1-31, 1-133, particulates,.and. tritium. The dose factor, Ri, was calculated using the methodology outlined in NIJREG-0133.: The radioiodine and particulate.ODCM Partl1 (Control DLCO 3.6.1:5) is applicable to the location in the unrestricted area where the combination of existing pathways and receptor age groups indicates the maximum potential exposure occurs, i .e, the critical receptor. Washout was calculated and determined to be negligible. Ri values have been calculated for the adult, teen, child and infant age groups for a ll pathways. However, for dose compliancecalculations, a maximum individual is. assumed that is a composite of highest dose.factor of each age group for each organ and pathway.. The methodology used to calculate these values follows: C.i Inhalation Pathway R,(I)-K'(BR).(DFA)ij. where: R 1 (1)k'(BR).(DPA)ija dosefactor for each identified radionuclide i of the organ of interest (units =mrem/yr per ý,Ci/m 3);= a. constant of unit conversion, 1E6 pCi/iCi= Breathing rate of the receptor of age group a, (units = m 3/yr);The inhalation dose factor for nuclide i, organ j and age group a,:and organ t (units = mrem/pCi). The breathing rates:(1BR)a for the various age groups, as given in Table E-5 of Regulatory Guide 1 109 Revision i1,are tabulated below.Age Group (a)Breathing Rate (m 3/vr)Infant Child Teen Adult 1400:3700 8000ý8000 Inhalation dose factors (DFA),t 2 for the various age groups. are given-in Tables E-7 through E-1.0 of Regulatory Guide 1.109 Revision 1, Unit I ODCM Revision 28 11,83 September 2006 APPENDIX C (Cont'd)C.2 Ground Plane Pathway R&(G) -KK"(SF)(DFG), (1-e-%it Where: R,(G) = Dose factor for the ground plane pathway for each identified radionuclide i*for the organ of interest(units = m 2 mrem/yr per ýtCi/sec)K' A constant of unit conversion, I1E6 pCi/pCi K"= A constant of unit conversion, 8760 hr/year= The radiological decay constant for radionuclide

i. (units sec 1)t = The exposure time, sec, 4.73E8 sec (15 years)(DFG)1 = The ground plane dose conversion factor for. radionuclide i; (units = rnrem/hr per pCi/m 2)SF The shielding factor (dimensionless)

A shielding factor of 0.7 is discussed in Table E- 15 of Regulatory.Guide 1.109 Revision 1. A tabulation of DFGj values is presented in Table E-6 of Regulatory:Guide.1J09-Revision 1...Unit ] ODCM Revision 28 II 84 September 2006 APPENDIX C (Cont'd)C.3 Grass-(Cow or Goat)-Milk Pathway R,(C) W1 U.. F-r)QEf S + 0_ -f4 5 (NX,+ X.) ;j Where: R.(C) Dose factor for the cow milkor goat milk pathway, for each identified radionuclidei for the organ of interest, (units = m2-mrem/yr per pC i/sec)K' A constant ofunit conversion, 1E6pCi/fiCi Qf The cow's or goat's feed consumption rate, (units = Kg/day-wet weight)U, The receptor's milk consumption rate for age group a. (units = litersiyr) YO The agricultural productivity .by unit area of pasture feed grass, :(units = kg/m2)YS The agricultural productivity by unit area of stored feed, (units= kg/m2)FM, 'The stable.element transfer coefficients, (units.= pCi/liter per pCi/day)r Fractibn of deposited activity retained on cow's feed.grass (DFL)i, = The ingestion dose factorfor nuclide i, age group a, and total body or organ t (units = mremfpCi)The radiological decay constant for radionuclide i, (units=sec -1)=. The decay constant for remOval of activity on leaf and plant. surfaces by weatheringequal to 5 .73E-7 sec -1 (corresponding to a 14 day hal.f-life)= The transport time from pasture to cow or goat, t.on ilk, to receptor,. (units = sec)th = The transport time. from pasture, to harvest, to. cow or goat, to milk, to receptor (units = sec)Unit I ODCM Revision 28.1I 85 September 2006 APPENDIX C (Cont'd)Fraction of theyear that the cowor goat is on pasture (dimensionless) fs= Fraction of the.cow feed that is pasture grass while the cow is on pasture (dimensionless) Milk cattle and goats are. considered to be fed from two potential sources, pasture grass and stored feeds.Following the development in Regulatory Guide 1. 109 Revision I, the value of f. is considered unity in lieu of site ,specific information, The value of:fp is 0.5 based on 6 month grazing period. This value for f, was obtained from the environmental group.Table C-1 contains the appropriate values and their source in .Regulatory Guide 1.109 Revision,]. The concentration of tritiumn inwimilk is based on the airborne concentration rather than t!he deposition. Therefore, the .RT(C) is based on X/Q: RPT(C) = K'K" FiQfU.(DFL)at 0.75(0.5/1) Where: RT(C)-Dose factor for the cow or goat milk pathway for tritium forthe organ of interest, (units =mrem/yr per Atci/m 3).= A constant of unit conversion, 1 E3 g/kg H= Absolute humidity of the atmosphere, (units =g/m 3)0.75= The fraction of total feed thatis water 0.5= The ratio of the specific activity of the feed grass water-to the atmospheric water Other values are given previously.. A site specific value of H equal to 6.14 g/m 3.is used. This value was obtained, from the, environmental group using actual site data.Unit 1 OODCM Revision 28 1 86 September 2006 APPENDIX C (Cont'd)CA4 Grass-Cow-Meat Pathwav Ri(C) = K' Q..) IX)(DF)i,,t (xi + ?~[f~f + (1-Lf 5) (e YP Ys-Xith )] e-t Ri(M) Dose factor for the meat ingestion pathway for radionuclide i foir any organ of-interest, (units = m1 2 mrem/yr per /aCi/sec)Ff The stable element transfer coefficients, (units pCi/kg per pCi/day)U:1 = The receptor's meat consumption rate for agegroup a, (units = kg/year)t4 = The transportltime from harvest, to cow, to receptor, (units, sec)tf= Tlie transport timie from'pasture, to cow, to receptor, (bnits =sec)All othertterms remain the same as defined for the milk pathway. Table C-2 contains the values which were usedin ca.cultting Ri(M).-The concentration of tritiwn in meat is based onairborne concentration rather than deposition. Therefore, the RT(M) is based on X/Q.RT(M) =K'IC"FQrUap(DFI,)iai [0.75(0.5/H)] Where: Rr(M) = Dose factor for the meat ingestion pathway .for tritium for any organ of interest, (units =mrem/yr perpCi/im 3)All other terms are defined above.C.5 Vegetation Pathway The integratedconcentration, in vegetation consumed by man follows theexpression developed for milk. Man is consideredto consume two types of vegetation (fresh and stored) that differ only in the time period between harvest and consumption, therefore: R4(V) =K *r (DFL)t ~Uý.F~e Y(1+ )L-?qtL+ Us ,,F 9e I Unit 1 ODCM Revision 28 September 2006 Il 87 APPENDIX C (Cont'd)Where: Rj(V) Dose. factor for vegetable pathway for radionuclide i for the organ of interest,.(units m 2-mrem/yr per ptCi/see)K' = A constant of unit conversion, 1E6 pCi/gCi UL. The consumption rate of fresh lca6, vegetation by the receptor in age group a, (units = kg/yr)S1. The con.umption rate of stored vegetation by the receptor in age group a (units = kg/yr)FL = The fraction of the annual intake of fresh leafy vegetation grown locally F. = The fraction of the. annual'intake ofstored vegetation grown locally tL = The average time between harvest. of leafy vegetation and its consumption, (units=,see)th = The average time between harvest of stored vegetation and its consumption, (units = see)YV : = The.vegetation areal P density, (units kg/m 2)All other factors have been dcfincd previously. Table C-3 presents the appropriate parameter values and their source in Regulatory Guide 1. 109 Revision .In lieu of site-speci fic:data, values for FL and Fg of, 1.0 and 0.76, respectiYcly, were used in the calculation. These values were obtainedfrom Table E-15 of Regulatory Guide 1.109 Revision 1.The concentration of tritium in vegetation is based on the airborneconcentration ratherthan the deposition. Therefore, theR-r(V) is based. on X/Q: R1(V) = K'K`" [ULafL + U', fe](DFL)a, 0.75(0.5/H) Where: RT(V) dose factor for the vegetable pathway for tritium for any organ of interest, (units = mrem/yr per AtCi/m 3).All otherterms are defined in preceeding sections.Unit] ODCM Revision 28 HI 88 Septem ber 2006 TABLE C-1 Parameters for Grass-(Cow or Goat)-MilkPathways Parameter Qf (kg/day)r (DFL)ija (mrem/pCi) F, (pCi/liter per pCi/day)Y (kg/mr 2)Y, (kg/mr 2)th (seconds)tf (seconds)U,,P (liters/yr) Value 50 (cow)6 :(goat)1.0 (radioiodines) 0.2 (particulates) Each: radionuclide Each stable element 2.0 0.7 7.78 x 10 6 (90 days)1.73 x 105 (2 days)330 infant 330 child 400 teen 3 10 adult Rcfcrcnce (Reg. Guide 1.109 Rev. 1)Table E-'3 Table E-3 Table E-4 5 Table E- 5 Tables E-1 I to E-14 Table E-.I (cow)Table E-2 (goat)Table E-1,5 Table E-15 Table E-15 Table E-15 Table E-5 Table E-5 Table E-5 Table E-5'Unit I ODCM Revision 28 September 2006 I 89 Parameter r Ff (pCi/Kg per pCi/day)U,, (Kg/yr)(DFL)ija (mrem/pCi) Yp (kg/m 2)yV (kg/m 2)th (seconds)tf (seconds)Qf (kg1day)TABLE. C-2 Parameters for the Grass-Cow-Meat Pathway Reference Value (Reg. Guide 1-109 Rev. I)1.0 (radioiodines). Table E-I5 0.2 (particulates) Table E- 15.Each stable element Table E-I 0 infant Table E-5 41 child Table E-5 65 teen Table E-5 110 adult Table E-5 Each radionuclide Tables E- 1 Ito E- 14 0.7 TableE-15 2.0 Table E-15 7.78E6 (90 days) Table E-1,5 1.73E6 (20 days) Table E- 1.5 50 Table E-3 Unit 1 ODCM Revision 28 September 2006 11 90 TABLE C-3 Parameters for the Vegetable Pathway Parameter r (dimensionless)(DFL)i;ia (mrem/pCi) UL)a (kg/yr) -:infant-child-teen-adult US)a (kg/yr) -'infant-child-teen-adult Value 1.0 (radioiodines) 0.2 (particulates) Each radionuclide 0 26 42 64 0 520 630 520 8.6E4 (1 day)5.18E6 (60 days)2.0 Reference (Rez. Guide 1.109 Rev. 1)Table E-1.Table E -I Tables E-11 to E-14*Table.E-5 Table E5-Table E-5 Table E&5 Table E-,5 Table E-5 Table E-5 Table E-5 Table E- 15 Table E-. 15 Table E- 15 t L (seconds)th (seconds) 2)Unit I ODCM ReVision.28 September 2006 1.1 91 APPENDIX D DIAGRAMS OF RADIOACTIVE LIQUID AND GASEOUS EFFLUENT TREATMENT SYSTEMS AND MONITORING SYSTEMS Unit I ODCM Revision 28 September 2006 II 92 I -ii T'~I~~iiZI7 L~1iif V L1I~IZZI1~ ~nTTIT1-- TA I I i iI I I.I-I 1i1 I I i I I ~ I 1.W.4 VALVE SymsoLS 4,14- ATET KIMAT 1-4TA-- mON DAMPE maxUTOV 3 VE N TILA I I QAM~P EA S LINE SYM.,L, ATATOLSVTLV DESIGN*ATO I In'PCIam* NOTE CONTRO 4A EO GAT ION I I I r4STRLMNTSMBL WI--, TW. .CONTROL VALVEOPLERATIQj ON ~ u: MOM:L PIP~NB SYMBOL!PT WI'A'-.FLOW DEVICES~A WAr 81;ýt6 44 A 14~7J¶E~C 6-PU WAPT-A V AAZTGTA ASTAt~ SAnTA 1145.141. APTIATA.TT1STT~ 5,505 (TArT TA TWTW AWWm PUMPS AND FANS-{-I-3-.4 WIEWS TO WIPTA Ar_ALVE POSIjIONS EXPANSION JOINTS FIG.UR-M ..ILD". PIPIN(NINE OFFS1 ONRlVALVE OPERATION OMIT MAIMS SLM.E D-O.0--TATZSU am1 MIARAETI INSTRUMENT AND'MENT SYMBOLS MILE POINT UNIT 1.~TE DOSE CALCT MANUAL FSIR 14 A- 13901 ,FIGURE I -t REVM3 1185) Ii II"a WWIt I~UT ILCO R 1 IMI NINEMML POINTCT WI Ii i-- --------------------------------------u~ J L j I XI-4 ,MAIN CONDENSER AIR REMOVAL AND OFF GAS SYSTEM ho..V.3t" 0UWAM FIGURE D-2 STr PACKING, EXHAUSTER, AND RECOMBINER NINE MILE POINT: UNIT I OFFSITE DOSE CALC. MANUAL m.ý REACTOR BUILDING VENTILATION SYSTEM Vl-;3 ,=%6 VD.K P~Tp, rc. W"" mr -a~. ~t*. MMA~C tNt tn~t FIGURE D-3 REACTOR BUILDliNG VENT SYSTEM NINE MILE POINT UNIT 1 OFFSITE DOSE CALC. MANUAL WASTE DISPOSAL BUILOIN,3 VENTILATION <SYSTEM ,it ----- --- ------ -- ------------- PIl---------------*- -L- -*"-wr ------- ---...S~~~ Awl-3W*100I NaZIV.~ l ...".......u FIGURE D-4 WASTE DISPOSAL BUILDING VENT SYSTIEM NINE MILE POINT UNIT I OFFSITE DOSE CALC. MANUAL STACK -PLAN AND ELEVATION R.A.P , ** .aPMC I5OIUW~fl~ PR~r Pý.ATTI: I M ad tL 377 -A*PLA N-r-. =s'-W FIGURE 1)D-5 NW-I STACK NIAGARA MOHAWK POWER CORPORATION NTNE MILE POINT UNIT I OFFSITE DOSE CALC. MANUAL Pm, i E vAr i co iq LOOKINO NO D-5 OFF GAS BUILDING VENTILATION SYSTEM 5-92 C F #4 Timm amDro 17 IL.. .... ..........I FIGURE D46 OFFGAS BUILDING VENT SYSTEM NINE MILE POINT UNIT I.OIFFSLT E DOSE CALC. MANUAL.I This Page and Figure Deleted.D-7 Stack Sample and Sample Return NIAGARA MOHAWK POWER, CORPORATION NINE MILE POINT -UNIT 1 OFFSFTE DOSE CALC. MANUAL D-8 This Page and Figure Deleted.D-9 This Page and Figure Deleted.D-1. )o!NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT -UNIT 1 OIFFSITE DOSE CAMC MANUAL___________________________ I OFFSITE DOSE CALC. MANUAL}}